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All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Evaluation of Thermolytic Hydrogen Generation Rate Models at High-Temperature/High-Hydroxide Regimes

This report describes the results of testing performed to extend the applicable ranges of temperature and hydroxide concentration for use within the Glycolate and Global Total Organic Carbon (TOC) Hydrogen Generation Rate (HGR) expressions. Seven experimental conditions (six simulants of the 242-25H Evaporator system chosen as a D-optimal set of experiments and a single test conducted at an elevated boiling point of 170 °C) were investigated in the presence of sodium glycolate and Xiameter TM AFE-1010. Glycolate was employed to study the extension of the Glycolate Thermolytic HGR expression while Xiameter TM AFE-1010 was employed to study the extension of the Global TOC Thermolytic HGR expression. The following conclusions were derived from this testing: The Glycolate Thermolytic HGR expression may be confidently used to predict thermolytic HGRs from glycolate at temperatures as high as 170 °C and hydroxide concentrations as high as 23 M.; The hydroxide and temperature-dependence predicted by the Global TOC Thermolytic HGR expression has been confirmed at temperatures as high as 170 °C and hydroxide concentrations as high as 23 M, suggesting that the Global TOC Thermolytic HGR expression may be used at these ranges.; Methane was observed from tests with Xiameter TM AFE-1010 at production rates higher than those observed for hydrogen. These rates were observed at temperatures higher than 100 °C.; Preliminary models suggest that increasing hydroxide/temperature causes an increase in Methane Generation Rate (MGR) from Xiameter TM AFE-1010. The following recommendations are based on this testing: The existing equations for thermolytic HGR from glycolate and non-glycolate organics should be used at Concentration, Storage, and Transfer Facilities (CSTF) storage and evaporation conditions, including temperatures and hydroxide concentrations exhibited in the 242-25H Evaporator.; Further investigation should be made into the influence of methylsilanes on CSTF flammability. This investigation should include: determination of the types of methylsilanes historically added to the CSTF, determination of methane formation rates from each type of methylsilane, and determination of the extent of degradation of methylsilanes in CSTF waste.; Characterization techniques should be developed by Savannah River National Laboratory (SRNL) to assist in the speciation of methylsilane-containing waste in the CSTF.; Additional testing with radioactive waste should be performed to determine the MGRs possible in radioactive waste and better inform model predictions made from testing with simulants.

08 HYDROGEN↗
DOE OSTI2016

Microstructural Characteristics of U-Zr Ingot Slug and Atomized Powder for SFR metal Fuel

Rod-type fuel slugs and spherical powder of U- 10wt.%Zr alloy was also fabricated and characterized in microstructure by injection casting, gravity casting and centrifugal atomization process. U grains became finer from ∼30μm in conventionally ingot rod to ∼2μm in atomized powder. Laminar structure became so fine from ∼0.2μm in conventionally cast ingot rod to 0.05∼0.1μm in atomized powder. Metallic fuels such as U-Zr and U-Pu-Zr alloys have been considered as a nuclear fuel for a sodium-cooled fast reactor (SFR), related to the closed fuel cycle for managing minor actinides and reducing a high radioactivity levels since the 1980's. Metallic fuel slugs of U-Pu-Zr alloys for SFR have conventionally been fabricated by an injection casting method. U-Pu-Zr alloy melts are injected and cast into quartz tube molds during injection casting. Metallic fuel slugs are collected by fracturing quartz tubes from the molds. Demolded quartz tubes are treated with radioactive wastes. Recently, management of minor actinides (MA) became an important issue because direct disposal of the long-lived MA can be a long-term burden for a tentative repository up to several hundreds of thousand years. The metallic fuel for SFR should be remotely fabricated under a radiation shielded environment such as a hot cell, because it contains highly radioactive MA such as Np, Am and Cm. The addition of MA in the fuel composition introduces new challenges in the manufacturing of a metallic fuel slug. Am has a high vapor pressure under a vacuum atmosphere at above a melting temperature of U-Zr or U-Pu-Zr. Used graphite crucibles and quartz molds cannot be recycled easily because of difficult decontamination of radioactive residues on their surfaces. It is necessary to minimize the vaporization of Am and volume of radioactive wastes. Atomized metallic fuel does not need these quartz molds during fabrication, as uranium alloys are prepared as a particle form. It is possible to reduce the formation of radioactive wastes by applying the fabrication process of a particulate metallic fuel. Metallic fuel having a fine microstructure, which originates from a rapid solidification effect during atomization process, results in a higher FG release rate during irradiation. In the case of applying a new forming method such as vibrocompaction in Zr sheath, the sheath can act as a reliable FCCI (fuel/cladding chemical interaction) barrier and a re-constraint for external elongation. In addition, atomized fuel powder, dispersed in a Na liquid metal having a high thermal conductivity, may have a lower temperature distribution during irradiation. In order to develop an efficient fabrication process of metallic fuel for SFR, U-Zr ingot rods were fabricated in a lab-scale with various casting methods. Rod-type U-Zr fuel slugs were fabricated by gravity casting and injection casting. Spherical powder of U-Zr alloy was also fabricated by centrifugal atomization. And then, the microstructural characteristics were investigated by using electron microscopes and energy-dispersive electron X-ray spectroscopy. Atomized U-10Zr powder generally showed spherical morphology of below 200 μm in particle size. U grains became finer from ∼30 μm in conventionally ingot rod to ∼2 μm in atomized powder. Laminar structure became so fine from ∼0.2 μm in conventionally cast ingot rod to 0.05∼0.1 μm in atomized powder. Atomized particulate metallic fuel for SFR has a fine microstructure, resulting in a higher fission gas release rate during irradiation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nevada National Security Site Waste Acceptance Criteria (NNSS/WAC)

This document establishes the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Field Office (NNSA/NFO), Nevada National Security Site Waste Acceptance Criteria (NNSSWAC). The NNSSWAC provides the requirements, terms, and conditions under which the Nevada National Security Site (NNSS) will accept the following: DOE low-level radioactive waste (LLW); DOE mixed low-level waste (MLLW); DOE classified waste/matter; U.S. Department of Defense (DOD) classified waste/matter The LLW and MLLW listed above may also be classified waste. Non-radioactive classified waste is accepted for disposal and shall meet the waste acceptance criteria for radioactive waste as specified in this document. Classified matter may be shipped to the NNSS for permanent burial as classified waste in accordance with the provisions in Section 3.1.18 of this document.

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

Fault Stability Perturbation by Thermal Pressurization and Stress Transfer Around a Deep Geological Repository in a Clay Formation

The increase of temperature of a low-permeability, fluid-saturated media may trigger significant thermal pressurization, where the expansion of pore fluid cannot be accommodated by the thermal expansion of the pore space. With the aid of a coupled thermohydromechanical numerical simulator, we investigate the possible impact of thermal pressurization during the life of a deep geological repository (DGR) for high-level radioactive waste, characterized by the emplacement of radioactive material-filled canisters in a series of parallel tunnels, excavated in a low-permeability clay formation. We represent the fault as a planar structure embedded in a thermoporoelastic material and shear activation evaluated by a strain-softening Mohr-Coulomb failure criterion. The results show that stress changes caused by temperature and thermal pressurization of a rock mass around the emplacement tunnels may trigger a slip event on a fault plane in proximity of the geological disposal site: Rupture nucleates at depth, hundreds of meters below the DGR. Stress transfer plays a key role, while a direct hydraulic connection between the repository and the nucleation zone is not necessary in order to trigger rupture. A low stress ratio may favor the occurrence of slip up to a distance of 600 m of the fault from the outermost tunnel. Furthermore, these results highlight the need of investigating hydromechanical properties and local stress conditions at depth to characterize the geomechanical response of weak planes located in the surroundings of a high-level radioactive waste repository and to provide sufficient knowledge for the safe development of the DGR site.

58 GEOSCIENCES↗
DOE OSTI2019

Port Hope Area Initiative Project Update - 19146

The Port Hope Area Initiative (PHAI) is a federal environmental clean-up program. Its mandate is the remediation and local, long-term, safe management of approximately 1.7 million cubic metres of historic low-level radioactive waste in the adjacent communities of Port Hope and Port Granby in Southern Ontario, Canada. The PHAI is being undertaken by Canadian Nuclear Laboratories (CNL), on behalf of Atomic Energy of Canada Limited, a federal Crown corporation. With funding of $1.28 billion CAD it is currently one of the largest environmental remediation projects in Canada. The safety of workers, the public and the environment while delivering the PHAI on time and within budget is CNL's number one priority. The PHAI is being carried out as two projects - the Port Hope Project and the Port Granby Project. The Port Hope Project is the larger of the two and includes the cleanup of approximately 1.2 million cubic metres of historic waste from various locations within the community. The Port Granby Project will relocate approximately 450,000 cubic metres of historic low-level radioactive waste and marginally contaminated soils, located at an existing waste management facility on the shoreline of Lake Ontario, to a new engineered aboveground mound being built about a kilometre north of the current site. This paper will provide an update on both projects with an emphasis on successes to date and challenges yet be faced. (authors)

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

Industry overview of TRANSCEND - 20271

The United Kingdom has a legacy of nuclear facilities to decommission and radioactive waste to manage and dispose of. The TRANSCEND programme is an independent research programme with significant participation from industrial organizations that are involved in that decommissioning programme and radioactive waste management. The participation has led to a programme that is both academically robust enough to attract significant funding from the EPSRC, the universities themselves and relevant enough to be supported financially by the industrial partners who see the needs and drivers. The programme offers benefits in terms of the development of people and in enabling and promoting innovation in both hardware and software tools. Industrial engagement is fundamental to being able to capitalise on that innovation. (authors)

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

Development of understanding of disposability of high-heat-generating wastes in the UK - 15359

The inventory of radioactive wastes planned for geological disposal in the UK is diverse. This inventory includes a range of high-heat-generating wastes and potentially some other nuclear materials (some spent fuel, uranium and plutonium) that are subject to government policy decisions and therefore may be declared as wastes for geological disposal in the future. To ensure that safe disposal solutions can be developed, it will be necessary to understand the influence of this heat on engineered barrier systems for the range of generic disposal concepts being considered in the UK. Furthermore, waste producers need to develop packaging solutions for these materials that take account of any thermal constraints. For these reasons, RWM has established the High-heat-generating Wastes Project (Project Ankhiale), a dedicated project to enhance understanding of the factors affecting geological disposal of high-heat-generating wastes with a view to supporting future decision making and concept selection. The project has been established around a multi-disciplinary Integrated Project Team (IPT), comprising members of RWM, Amec Foster Wheeler and external experts, to provide a comprehensive approach across a range of technical disciplines. Phase 2 of the project, which comprises a series of specific technical activities, commenced in March 2013 and, since then, progress has been made across the following five broad categories: - Compilation of the inventory of UK high-heat-generating wastes;- Waste package performance work; - Development of a better understanding of thermal constraints for different buffer materials; - Exploration of the range of disposal parameters using a thermal dimensioning modelling tool; and - Concept engineering design work. (authors)

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

In Situ Disposal to Optimize Decommissioning Waste Management - 17455

Cavendish Fluor Partnership (CFP) is teaming with the UK government's Nuclear Decommissioning Authority (NDA) to explore opportunities for onsite disposal of structures and wastes from decommissioning of twelve Magnox nuclear sites. CFP is piloting implementation of draft guidance issued by the Environmental Agency (EA), Scottish Environmental Protection Agency (SEPA) and Natural Resources Wales (NRW) designed to optimise management of wastes generated in closing nuclear sites positioned across Scotland, Wales and England. Where it can be demonstrated to be the best overall solution for people and the environment, on-site / in situ disposal as an option for optimising waste management could save millions of pounds sterling in the overall costs for site closure as opposed to previous plans that required retrieval, packaging, transportation and offsite disposal of all radioactive wastes generated during site decommissioning. The UK government's Department for Business, Energy and Industrial Strategy (BEIS), NDA, and UK regulators are consulting with stakeholders on a proposal to amend legislation and enable better, more proportionate regulation of nuclear sites in the final stages of decommissioning and clean-up. Practical application of existing UK legislation typically means removing most, if not all, radioactive waste and residual contamination from the nuclear licensed site for disposal or management elsewhere, irrespective of any wider health and safety, social, environmental or economic considerations. Proposed changes to legislation would give greater weight to the environment agencies' guidance and provide for a more sustainable approach where optimization could include on-site/in situ disposal solutions and early opportunities to reuse land with appropriate regulatory oversight. Recent characterization activities at two reactor sites have identified areas of sub-grade low level contamination in structures and soil around used fuel ponds. CFP, a partnership of UK-based Cavendish Nuclear Limited and US-based Fluor Corporation, has initiated preliminary site modelling that indicates much of this material, including significant volumes of sub-grade structures would be conducive for in situ disposal under the new proposed guidance. Based on Fluor experience with similar decommissioning projects at US Department of Energy sites, CFP will look to leverage in situ disposal to optimise decommissioning waste management and consequently optimise overall facility decommissioning strategies. (authors)

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

Development of an Execution Strategy Analysis (ESA) Capability and Tool for the Management of Spent Nuclear Fuel (SNF) - 16575

An Execution Strategy Analysis (ESA) capability and tool are being developed to evaluate alternative execution strategies for future deployment of a consolidated Interim Storage Facility using a consent-based siting process per the Administration's Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (Strategy). Application of an ESA approach not only leverages but also goes beyond traditional project analysis tools. The ESA tool allows for ongoing performance assessment of the evolving project execution plan that takes into account significant assumptions, risks, and uncertainties throughout the project lifecycle. The ESA process and tool will be used to support the development of plans, budgets, and alternative execution/implementation strategies for meeting the goals of the Strategy. This paper describes the development of the ESA capability and tool and identifies the value of applying such a capability in implementing a long-term strategy for managing spent nuclear fuel and high-level radioactive waste. (authors)

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

Development of a Conditioned System-Level Groundwater Model to Evaluate Long-Term Groundwater Impacts within a Performance Assessment - 20115

A preliminary performance assessment (PA) of single-shell tank Waste Management Area (WMA) A-AX located at the U.S. Department of Energy (DOE)'s Hanford Site in southeastern Washington is being conducted to satisfy the requirements of DOE Order 435.1 [1] as it relates to closure of the single-shell radioactive waste tanks in the WMA A-AX tank farms. A PA assesses the fate, transport, and impacts of radionuclides within a low-level radioactive waste disposal facility in its assumed closure configuration and the subsequent potential doses to humans over a 1,000-year compliance period and a 10,000-year performance evaluation period. The WMA A-AX preliminary PA evaluation is structured around the complementary use of process-level and system-level models to calculate the facility performance against established DOE Order 435.1 [1] performance objectives. Process-level models are those that represent a detailed phenomenological representation of processes of concern in the PA. Process models typically only represent one or a few of the components of the PA, such as groundwater flow and transport, and must be integrated with other modeling elements to perform PA calculations. System-level models are those that are abstracted from the process models, retaining the essential features of the process model, while allowing integration of all aspects of the PA in a single modeling framework. System-level models are often characterized by coarser numerical discretization, lower dimensionality, or other similar simplifications compared to the process-level model. Traditionally, a three-dimensional (3-D) process-level model is utilized primarily to evaluate the long-term impact on groundwater and the potential doses to individuals who consume contaminated groundwater. System-level models are also utilized to evaluate the groundwater pathway in PAs. These models typically have reduced dimensionality (1-D) and are conditioned utilizing flow fields (Darcy fluxes) and moisture content distributions that are abstracted from the process level models. The abstraction approach assures that the flow field in both models is consistent for a specific set of input parameters for flow, differing only in the discretization and dimensionality of the two models. The preliminary WMA A-AX PA incorporates a detailed representation of the geological system and hydraulic properties within the 3-D model STOMP{sup C} numerical code so that the effects of relevant features and processes on water flow and radionuclide transport in the subsurface can be evaluated. The complementary system-level model is developed utilizing the GoldSim{sup C} code to implement a simplified, 1-D equivalent model to represent the groundwater pathway. Contaminant transport through the vadose zone and unconfined aquifer for Tc-99 and I-129 were evaluated in each of the models. Adjustments to the saturated portion of the GoldSim{sup C} 1-D model were required to mimic the dispersion effect captured with the 3-D STOMP{sup C} model. Once this conditioning was conducted, highly similar results for the transport of Tc-99 and I-129 were achieved at the point of calculation, located 100 m downgradient from the WMA A-AX fenceline. These radionuclides represent elements that are regarded as primary dose drivers in the PA groundwater pathway analysis. The high degree of conformance between the two models suggests that the use of the equivalent 1-D system model is suitable for evaluating the full suite of radionuclides that will be released from the tank sources within WMA A-AX over the 1,000-year compliance period and over the 10,000-year evaluation period. (authors)

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

Innovative Data Management System in Support of Remedial Actions for Legacy Radioactive Contamination - 17280

The Port Hope Area Initiative (PHAI) is a project mandated and funded by the Government of Canada to remove contamination from properties with legacy low-level radioactive waste (LLRW) contamination arising from a crown-owned radium and, later, uranium refinery in the Town of Port Hope, Ontario. From 1976 to 1981, a community clean-up project was undertaken, with the specific purpose to decrease the level of historic LLRW contamination within Port Hope. Since then, residual LLRW has been managed by the Low Level Radioactive Waste Management Office (LLRWMO) (and its predecessors). The PHAI and the Port Hope Project (PHP) were initiated in 2001 with a formal agreement between affected municipalities and the federal government to develop a safe, long-term management solution for the historic LLRW remaining within Port Hope. An important activity of the PHP is the survey of about 4,800 'small scale sites' (SSS) throughout the community to confirm the presence or absence of LLRW. This involves surveys of properties based on the measurement of exterior gamma, contaminated objects and surfaces, soil concentrations and indoor radon and classification to Port Hope project specific criteria. A very large amount of information is involved in the project including; scheduling individual field visits to the properties, capture of field data and comments, laboratory sample tracking, QA/QC, report generation and internal and client project reporting at regular schedules. Field data collection of exterior gamma radiation surveys includes direct (real time) and interactive loading of data into the Information Management (IM) system to support initial QA/QC and reporting back to the field on radiation levels. Screening soil measurements including X-ray fluorescence (XRF) and gamma radiation are collected from all 20 cm increments in each borehole in addition to down-hole gamma radiation measurements and logging of soil characteristics which are used in documentation and interpretation of the property condition. The IM system outlined in this paper was developed to respond to the requirements for radiological surveys in Port Hope. The IM system continues to evolve as a part of continuing review and improvement. The IM system incorporates Web-enabled field tools and cloud hosted data and applications, provides increased field efficiency, timeliness in reporting, real time QA/QC analysis, and consistency in tracking project advances and changes made to the data for decision-making and reporting on an ongoing basis. Examples of specific applications are provided. (authors)

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

The underground research facilities (URF) network - illustrating the IAEA's concept and use of professional networks - 15382

Since 2001, the IAEA has championed the concept and use of professional networks (Communities of Practice) to advance best practices and to provide a platform for cooperation in radioactive waste management, decommissioning and environmental remediation across the globe. The Underground Research Facilities (URF) Network was the first in a series of networks implemented under the auspices of the IAEA to share experiences, elicit best practices, and transfer knowledge on topics relevant to the development and implementation of geological disposal of long lived and high level radioactive waste - i.e. intermediate level waste (ILW), high-level waste (HLW) and/or spent nuclear fuel (SNF). Special emphasis is placed on the role and use of URFs to support such geological disposal developments. In addition to the direct benefits provided to Member States through its programme of work, the URF Network also delivers less tangible, yet equally valuable benefits through the informal exchange of information, the development of peer-to-peer professional networks, and seeding of multi-lateral cooperation between Member States. Network members represent a range of nuclear programme sizes and advancements in their respective repository programmes. After a decade of fruitful cooperation, delivering more than 30 training courses on topics relevant to geological disposal, the URF Network is gradually reassessing its mode of operation and focus of work. Emphasizing selected strategic goals - focused in part on more systematic training of URF Network members human resources and in part on select key technical issues - would benefit all network members, while remaining cautious in the commitment of scarce resources. (authors)

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

Application of SIAL{sup R} Solidification Technology for Processing of Problematic Waste Streams - 18267

SIAL{sup R} is a proven, safe, cost-effective and licensed waste solidification technology that has been used successfully for 20 years in Slovakia and the Czech Republic for processing of various problematic radioactive waste streams. This abstract presents and summarises the most interesting performance records of the SIAL{sup R} geopolymer solidification technology, the applied mobile devices as well as the characteristics of final products. Radioactive waste solidification into matrix SIAL{sup R} is based on polycondensation of inorganic aluminosilicate components with chemical composition similar to average cement. The main components of matrix SIAL{sup R} are natural inorganic raw materials containing SiO{sub 2}, Al{sub 2}O{sub 3}, CaO, MgO, Na{sub 2}O, K{sub 2}O, Fe{sub 2}O{sub 3}, and TiO{sub 2}. The treatment/solidification process is modified for individual radioactive streams with different compositions. To determine the most appropriate composition, the SIAL{sup R} application technology is laboratory tested on real samples of each individual waste stream identified for solidification. Then an appropriate mixture of aluminosilicate composition and other inorganic compounds is defined for a given solidification application. When determining the most appropriate mixture, other requirements are taken into consideration: maximum dose rate on the drum with solidified product, mass, volume, alpha radionuclide content in the final product, need for a waste's dewatering before solidification, dry matter content in the solidification product as well as other specific waste acceptance criteria (i.e., compressive strength, leachability, diffusion coefficients, radiation stability, biodegradability of matrix, distribution and partition coefficients, influence of frost and water presence to the product stability including long term physical stability in water, dustiness, flammability, time dependence of compression strength and water reduction, product homogeneity and also radiolytic G-value for hydrogen release, GH{sub 2}). The Amec Foster Wheeler baseline approach to implementing the SIAL{sup R} waste treatment process is structured on three steps - waste characterization, waste retrieval and pre-treatment and at last the solidification of pre-treated waste itself. The equipment used for the solidification process is relatively small, lightweight and designed for the specific application on site, directly 'in situ' or local to the tanks storing the sludge's/resins/other waste streams. This mitigates the problems normally experienced by transferring sticky, tight waste through long pipes. The equipment is normally 'tailor-made' to take into account the real space conditions and other existing facility requirements. At Bohunice and Dukovany sites, the facilities for encapsulation were designed to occupy existing areas of the plants in close proximity to the tanks where the waste was being stored. This allowed for the removal of the waste using Remote Operated Vehicles and for the waste to be immediately treated on site without the need for transportation thus saving both, time and resource. To date, SIAL{sup R} has been used successfully to immobilize approximately 2 x 10{sup 6} kg of waste which includes sludge, resin, crystalline borates and contaminated organic waste from Bohunice and Mochovce, Slovakia and approximately 300 m{sup 3} of spent ion exchange resins and sludge from tanks on site at the Dukovany NPP (Nuclear Power Plant) in the Czech Republic. The technology now has a track record of over 20 years which includes on-going research and development. The International missions WANO (the World Association of Nuclear Operators) and OSART (the Operational Safety Review Team) evaluated the SIAL{sup R} matrix technology at Dukovany NPP as an example of good practice. (authors)

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

Transport and Treatment of Chapelcross Gas Ducts for Clearance and Recycling - 19662

In the summer of 2018, Cyclife, the EDF group's specialist radioactive waste management and decommissioning company, transported sixteen gas ducts and associated equipment from the shutdown Chapelcross nuclear power plant to treatment facilities in the UK and Sweden. The client requested a full removal, treatment and disposal service for approximately 820 te of low level metallic waste comprising Upper Gas Duct Assemblies (including bellows), Upper Gas Duct elbows, Filter Pots and Convoluted pipework. The complex scope involved heavy lifting, transport by road and sea, processing and recycling on a large scale, thereby necessitating involvement of a wide range of suppliers at different tiers all managed by Cyclife. The purpose of the entire project was to safely handle and treat all of the components for clearance and recycling into the public domain for the manufacture of new steel products. Most of the treatment took place in Cyclife's nuclear licensed metal treatment facility in Sweden. Cyclife has a long and unique experience of arranging safe transport of large contaminated components on road and sea. The metal treatment facility in Sweden can, without special arrangements, receive components up to 30 meters long with a weight of up to 400 tonnes for decontamination, size reduction and melting. The actual treatment of the components consisted of the following steps: a) Separation of a gas duct from the gantry, b) size reduction, radiological verification measurements and clearance of the gantry, c) stepwise segmentation of the gas duct and elbow including removal of asbestos, d) decontamination, e) melting, f) radiological assessment and clearance of the metal ingots, g) management of the secondary waste for return to UK. The result of the project, upon completed treatment of all the material, demonstrates a high recycling rate of the material. 5-10% of the original mass of material will have to disposed as radioactive waste. This number include all residues, untreatable material (asbestos, mineral wool, filters etc.) and two melt batches with a too high activity content for clearance. The paper takes the reader through the entire project from the initial planning to the clearance of the metal and the return of the secondary waste to UK for disposal. It will also present and discuss pros and cons for alternative management approaches in several perspectives. (authors)

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

Opportunities and limits of Decay storage based on Dismantled Materials in Nuclear Power Plants - 17650

We investigate the potential of decay storage of low and intermediate radioactive waste for release procedures. We present the Safetec software tool HL3 and its application to data from the German nuclear power plant at the site in Stade. In general, during the dismantling process of nuclear power plants, waste of slightly contaminated and/or activated systems is produced in large amounts. These are in particular demolition rubble from concrete structures and ingots from the melting process of metals. If, at the time of disposal, the nuclear inventory is near or slightly above federal clearance limits, the material cannot be released as conventional waste. However, due to the decline of the nuclear inventory over time, a significant amount of that waste could qualify for the release procedure in the future. Although this concept of decay storage is well known, specific predictions of the mass flow to be released is non-trivial. The mass flow over time depends on the composition of known and assumed nuclides (nuclide vector), as well as on the underlying threshold values for release. To handle the waste stream as a function of time, the nuclide composition and specific threshold values, Safetec developed a bespoke software tool named HL3. Its primary function is to visualize and to test/analyze various nuclide vectors and its decompositions over time. For the application within German law, threshold values according to the different disposal routes can be considered. Using the batch interface in the software tool HL3, large amounts of data can be automatically analyzed in a traceable and quick way. To test both, the tool and the benefit for the concept of decay storage, we have applied our code HL3 to material from the dismantling of the pressurized-water reactor at the site in Stade, Germany. We set the focus on concrete debris and metal ingots, which in principle are best suited for decay storage over years. The assignment and the packaging of that waste from Stade were based on application documents for decommissioning of the Year 2000 and according to the state of art conditioning technique at that time. Based on the data of that waste and the appropriate disposal routes, we calculate and discuss the amount of waste releasable after decay storage. As for the situation in Germany, we find that only a negligible amount would be releasable, mainly due to other reasons. These are the deterioration of detectability of radionuclides with time, and other regulatory conditions. However, on an international scale, a different assessment can be drawn. Our results thus motivate to reconsider the option of interim decay storage to reduce radioactive waste in large amounts. (authors)

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

Developments in the Regulatory Framework of Interim Storage of Spent Nuclear Fuel (SNF) in Germany - 17374

According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license, granted by the competent authority in this field, which has been the Federal Office of Radiation Protection (BfS) until July 2016 and currently is the Federal Office for the Safety of Nuclear Waste Management (BfE). A first license was granted in 1983 for storage of spent nuclear fuel (SNF) and vitrified high level waste (HAW) at the interim storage facility in Gorleben. This license was reviewed in 1995. Dry Interim storage at the NPP sites in its current form started in 2002 in the interim storage facility next to the NPP Lingen. Since that time at each site of an operational NPP storage facilities were licensed and erected. Thus additionally to the three central storage facilities (Ahaus, Gorleben and Rubenow) there are now 12 storage facilities onsite the NPPs. All of these facilities use dry storage in metallic dual purpose casks (DPCs). The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years after closing the transport and storage cask. The first license to expire will be the storage license for the interim storage facility in Gorleben in the year 2034. The German 'Programme for the safe and responsible management of spent fuel and radioactive waste' expects to take the disposal facility into operation in the year 2050. Thus necessary preparations for interim storage till that time have to be taken on the side of the regulator. In the German Atomic Energy Act precautions have already been taken and a consulting of the German parliament (Bundestag) is mandatory if inevitable events would require an extension of currently licensed interim storage period. In the year 2016, the regulatory body in Germany was given a new structure. The BfS was formerly competent authority in the field of radiation protection since 1989. It had competencies for ionizing and non-ionizing radiation protection (e.g. mobile phone radiation, high voltage power lines and the effects of electric fields), licensing storage of spent nuclear fuel and as the operator of facilities for final disposal. Now there is a new structure with two Federal Offices (BfS and BfE) and a federal company for final disposal (BGE). The BfE is responsible for regulation in the field of nuclear disposal since 30.07.2016. So the package design approval for DPCs and the storage licenses for SNF are now granted by this office. The search for a repository site will be coordinated there as well. The site selection, construction and operation of the repository will be in the responsibility of BGE. The parts of the former BfS dealing with recovery of the waste from Asse (a former salt mine used for disposal research) and the construction of the repository Konrad (for low and medium level radioactive waste) will be transferred into that company. It is planned to incorporate the German company for construction and management of waste disposal facilities (DBE) which is owned by the German utilities as well. The BfS will focus on the scientific foundations and further advancement of radiation protection. (authors)

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

Evaluation of Low-Level Waste Disposal Receipt Data for Los Alamos National Laboratory Technical Area 54, Area G Disposal Facility – Fiscal Years 2015-2017

The Los Alamos National Laboratory (LANL or the Laboratory) generates radioactive waste as a result of various activities. Operational or institutional waste is generated from a wide variety of research and development activities including nuclear weapons development, energy production, and medical research. Environmental restoration (ER), and decontamination and decommissioning (D&D) waste is generated as contaminated sites and facilities at LANL undergo cleanup or remediation. The majority of this waste is low-level radioactive waste (LLW) and is disposed of at the Technical Area 54 (TA -54), Area G disposal facility.

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DOE OSTI2017

Routing Options for Transporting Commercial Spent Nuclear Fuel in the Continental United States - 17504

The U.S. Department of Energy (DOE) has a goal to develop solutions for the long-term, sustainable management of the nation's spent nuclear fuel and high-level radioactive waste. The DOE is planning for an integrated waste management system to transport, store, and dispose of spent nuclear fuel and high-level radioactive waste from commercial electricity generation, as well national defense activities. The DOE Office of Nuclear Energy (DOE-NE), Office of Spent Fuel and Waste Disposition's (SFWD) Integrated Waste Management (IWM) program is applying integrated waste management system analysis, system engineering, and decision analysis principles to inform future decisions regarding potential future nuclear waste management system architectures. Integrated waste management system architecture analyses are being conducted to support the future deployment of a comprehensive system for managing nuclear waste that considers all aspects of the back-end of the nuclear fuel cycle (transportation, storage, and disposal). One aspect of a potential future nuclear waste management system is transportation of spent nuclear fuel from the fleet of commercial nuclear reactors to destinations where waste management facilities, such as an interim storage facility, would be located. Example rail transportation routes from commercial nuclear reactor sites to various locations in the continental U.S. - chosen as hypothetical locations for future waste management system facilities - were evaluated in 2014 using DOE's GIS-based Stakeholder Tool for Assessing Radioactive Transportation (START). The objective of that evaluation was to identify and parameterize generic transportation routes for use in integrated waste management system analyses, the associated uncertainty in these selections, and the uncertainty associated with different transportation route selection criteria. The analysis presented herein considers rail transportation routes from commercial nuclear reactor sites to hypothetical destinations within the continental U.S. (authors)

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