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Complete β-decay patterns of 142 Cs, 142 Ba, and 142 La determined using total absorption spectroscopy

Background: The β decays of fission products produced in nuclear fuel are important for nuclear energy applications and fundamental science of reactor antineutrinos. In particular, nuclear reactor safety is related to the decay modes of radioactive neutron-rich nuclei, primarily via the emission of γ rays, neutrons, and electrons. Additionally nuclear reactors are the most powerful man-made source of antineutrinos emitted during the β decay of fission products. These antineutrinos are used to inspect fundamental properties of leptons as well as informing reactor operation. However, the majority of data on complex decays of fission products collected in the evaluated nuclear data repositories like Evaluated Nuclear Structure Data File (ENSDF) and Evaluated Nuclear Data Files (ENDF) are based on low-efficiency and often incomplete measurements resulting in questionable reference reactor antineutrino flux predictions, see the analysis by [Nichols, J. Nucl. Sci. Technol. 52, 17 (2015)]. Various assessments like the one done under the auspices of the [Yoshida et al., Assessment of Fission Product Decay Data for Decay Heat Calculations: A report by the Working Party on International Evaluation Co-operation of the Nuclear Energy Agency Nuclear Science Committee (Nuclear Energy Agency, Organization for Economic Co-operation and Development, Paris, France, 2007), Vol. 25], as well as by [Sonzogni, Johnson, and McCutchan, Phys. Rev. C 91, 011301(R) (2015)] and [Dwyer and Langford, Phys. Rev. Lett. 114, 012502 (2015)], list the A = 142 isobars with high cumulative fission yield among the important nuclei where data for reactor decay heat and/or antineutrino production should be verified and/or improved. Purpose: Here, our goal is to improve the quality of β -decay measurements and evaluate the impact of modified decay schemes on reactor decay heat and antineutrino energy spectra, for fission products along the A = 142 isobaric chain. This work is an in depth follow-up on [Rasco et al., Phys. Rev. Lett. 117, 092501 (2016)]. which presented briefly the impact of the corrected decay scheme of 142 Cs . Here, we extend the data to full isobaric decay chain including the daughter nuclei, 142 Ba and 142 La, and present more details on the 142 Cs results. Method: The decays of neutron-rich isobars of mass A = 142 produced by means of proton-induced fission of 238 U were measured using the Modular Total Absorption Spectrometer (MTAS) array on-line at the mass separator and Tandem accelerator at Oak Ridge National Laboratory. Results: The β -decay schemes for 142 Cs and 142 La were modified with respect to the nuclear data repositories. A small β-delayed neutron branching ratio for 142 Cs emitter was remeasured as $0.10^{+5}_{–3}% %. Improved precision on the measured half-lives is reported. Small corrections to the low-energy decay of 142 Ba are made. The β-decay patterns for 142 La and 142 Cs are presented. The decay heat release and cross section for the detection of reactor antineutrinos are deduced and compared to earlier results. Conclusions: The β-feeding pattern for 142 Cs having decay energy value $Q_β$ of over 7 MeV was substantially modified with respect to the current ENSDF entry. Smaller changes were encountered for 142 La, but since this A = 142 isobar also has a large cumulative yield in fission, the changes influence both decay heat and the antineutrino spectra. The previously known β intensities for 142 Ba decay ($Q_β$ value of 2.2 MeV) were verified and slightly modified. Overall, increased decay heat values and lower flux of antineutrinos interacting with matter are presented.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

How to make deposition of images a reality

An analysis is performed of the technical and financial challenges to be overcome if deposition of primary experimental data is to become routine. The IUCr Diffraction Data Deposition Working Group is investigating the rationale and policies for routine deposition of diffraction images (and other primary experimental data sets). An information-management framework is described that should inform policy directions, and some of the technical and other issues that need to be addressed in an effort to achieve such a goal are analysed. In the near future, routine data deposition could be encouraged at one of the growing number of institutional repositories that accept data sets or at a generic data-publishing web repository service. To realise all of the potential benefits of depositing diffraction data, specialized archives would be preferable. Funding such an initiative will be challenging.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Continuous mutual improvement of macromolecular structure models in the PDB and of X-ray crystallographic software: the dual role of deposited experimental data

Macromolecular structures deposited in the PDB can and should be continually reinterpreted and improved on the basis of their accompanying experimental X-ray data, exploiting the steady progress in methods and software that the deposition of such data into the PDB on a massive scale has made possible. Accurate crystal structures of macromolecules are of high importance in the biological and biomedical fields. Models of crystal structures in the Protein Data Bank (PDB) are in general of very high quality as deposited. However, methods for obtaining the best model of a macromolecular structure from a given set of experimental X-ray data continue to progress at a rapid pace, making it possible to improve most PDB entries after their deposition by re-analyzing the original deposited data with more recent software. This possibility represents a very significant departure from the situation that prevailed when the PDB was created, when it was envisioned as a cumulative repository of static contents. A radical paradigm shift for the PDB is therefore proposed, away from the static archive model towards a much more dynamic body of continuously improving results in symbiosis with continuously improving methods and software. These simultaneous improvements in methods and final results are made possible by the current deposition of processed crystallographic data (structure-factor amplitudes) and will be supported further by the deposition of raw data (diffraction images). It is argued that it is both desirable and feasible to carry out small-scale and large-scale efforts to make this paradigm shift a reality. Small-scale efforts would focus on optimizing structures that are of interest to specific investigators. Large-scale efforts would undertake a systematic re-optimization of all of the structures in the PDB, or alternatively the redetermination of groups of structures that are either related to or focused on specific questions. All of the resulting structures should be made generally available, along with the precursor entries, with various views of the structures being made available depending on the types of questions that users are interested in answering.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

Raw diffraction data preservation and reuse: Overview, update on practicalities and metadata requirements

A topical review is presented of the rapidly developing interest in and storage options for the preservation and reuse of raw data within the scientific domain of the IUCr and its Commissions, each of which operates within a great diversity of instrumentation. A résumé is included of the case for raw diffraction data deposition. An overall context is set by highlighting the initiatives of science policy makers towards an `Open Science' model within which crystallographers will increasingly work in the future; this will bring new funding opportunities but also new codes of procedure within open science frameworks. Skills education and training for crystallographers will need to be expanded. Overall, there are now the means and the organization for the preservation of raw crystallographic diffraction data via different types of archive, such as at universities, discipline-specific repositories (Integrated Resource for Reproducibility in Macromolecular Crystallography, Structural Biology Data Grid), general public data repositories (Zenodo, ResearchGate) and centralized neutron and X-ray facilities. Formulation of improved metadata descriptors for the raw data types of each of the IUCr Commissions is in progress; some detailed examples are provided. Lastly, a number of specific case studies are presented, including an example research thread that provides complete open access to raw data.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

Partial defect verification of spent fuel assemblies by PDET: Principle and field testing in Interim Spent fuel Storage Facility (CLAB) in Sweden

Here, the need for the development of a credible method and instrument for partial defect verification of spent fuel has been emphasized over a few decades in the safeguards communities as the diverted spent fuel pins can be the source of nuclear terrorism or devices. The need is increasingly more important and even urgent as many countries have started to transfer spent fuel to so called "difficult-to-access" areas such as dry storage casks, reprocessing or geological repositories. Partial defect verification is required by IAEA before spent fuel is placed into "difficult-to-access" areas. Earlier, Lawrence Livermore National Laboratory (LLNL) has reported the successful development of a new, credible partial defect verification method for pressurized water reactor (PWR) spent fuel assemblies without use of operator data, and further reported the validation experiments using commercial spent fuel assemblies with some missing fuel pins. The method was found to be robust as the method is relatively invariant to the characteristic variations of spent fuel assemblies such as initial fuel enrichment, cooling time, and burn-up. Since then, the PDET system has been designed and prototyped for 17×17 PWR spent fuel assemblies, complete with data acquisition software and acquisition electronics. In this paper, a summary description of the PDET development followed by results of the first successful field testing using the integrated PDET system and actual spent fuel assemblies performed in a commercial spent fuel storage site, known as Central Interim Spent fuel Storage Facility (CLAB) in Sweden will be presented. In addition to partial defect detection initial studies have determined that the tool can be used to verify the operator declared average burnup of the assembly as well as intra-assembly bunrup levels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Automated Metadata Extraction: Challenges and Opportunities

Proper application of the FAIR data principles is what separates a vibrant data ecosystem, in which research data are frequently shared and reused, from a lifeless data graveyard. Automated metadata extraction systems have been proposed as a means of bolstering the findability, interoperability, and reusabil- ity of data repositories with little or no human intervention. These extraction systems mine metadata by crawling a repository and applying lightweight extractors that, for various types of file (e.g., image, CSV file), extract or synthesize relevant attributes. In practice, however, the automated creation of generally useful metadata is fraught with challenges. Data consumers may have different perspectives as to what metadata representations are useful, the standards for recording metadata tend to change over time, and the software model for processing updates can introduce unnecessary human and computational effort. Thus, generalizing extraction for a broad audience of data consumers is a difficult and relatively unsolved problem.In this work, we explore these challenges faced by extraction systems in the context of constructing our own extraction system for science data. We first define the metadata extraction problem and provide context to the issues faced in generalizing metadata. Additionally, we identify potential research directions to help alleviate many of these challenges for all automated extraction systems. Ultimately, this work represents a first step in designing ubiquitous metadata extraction systems that can maximize the value of research data while minimizing the human efforts required in doing so.

Skluzacek, Tyler↗

Modeling Interfacial Glass-Water Reactions: Recent Advances and Current Limitations

Describing the reactions that occur at the glass-water interface and control the development of the altered layer constitutes one of the main scientific challenges impeding existing models from providing accurate radionuclide release estimates. Radionuclide release estimates are a critical component of the safety basis for geologic repositories. The altered layer (i.e., amorphous hydrated surface layer and crystalline reaction products) represents a complex region, both physically and chemically, sandwiched between two distinct boundaries pristine glass surface at the inner most interface and aqueous solution at the outer most interface. Computational models, spanning different length and time-scales, are currently being developed to improve our understanding of this complex and dynamic process with the goal of accurately describing the pore-scale changes that occur as the system evolves. These modeling approaches include geochemical simulations [i.e., classical reaction path simulations and glass reactivity in allowance for alteration layer (GRAAL) simulations], Monte Carlo simulations, and Molecular Dynamics methods. Finally, in this manuscript, we discuss the advances and limitations of each modeling approach placed in the context of the glass-water reaction and how collectively these approaches provide insights into the mechanisms that control the formation and evolution of altered layers.

36 MATERIALS SCIENCE↗

Development and Durability Testing of a Low‐Temperature Sintering Bi–Si–Zn Oxide Glass Composite Material (GCM) 129 I Waste Form

The capture and safe storage of radiological iodine ( 129 I) from nuclear fuel reprocessing is of concern due to its long half‐life and potential mobility in the environment. The development of durable waste forms in which to store captured iodine requires materials that are both compatible with the iodine capture phases and durable to repository environments. To that end, Sandia is developing a low‐temperature sintering Bi–Si–Zn oxide glass composite material (GCM) waste form and herein presents results of durability testing. Furthermore, durability studies were extended to both the occluded iodine capture material Ag‐Zeolite (Mordenite, MOR) as well as the GCM, synthesized with compositional variations including: amount of Ag flake added, AgI‐MOR particle size in the GCM, and mass loading of I within the AgI‐MOR. Product consistency test (PCT‐B), chemical durability (MCC‐1) tests, and single‐pass flow‐through (SPFT) tests, were performed on both the individual components of the GCM and the completed GCMs. Durability tests indicate low GCM dissolution rates (<10 −3 g/m 2 ·d) across wide variable ranges including: pH, AgI‐MOR loading, I loading, and AgI‐MOR particle size. Results indicate that the Bi–Si–Zn oxide glass matrix sharply limits the release of iodine from the otherwise relatively fast degrading AgI‐MOR getter material. Furthermore, the formation of an amorphous AgI phase during sintering of the GCM results in the limitation of iodine release during waste form degradation. Durability of GCM and release rates approximate those of established nuclear waste glasses, or analogues such as basaltic glass. This suggests that the Bi–Si–Zn GCM is a viable candidate as a repository iodine waste form.

Mowry, Curtis D.↗

'Strategy is a commodity, implementation is an art' - 2 years of implementation of the UK national LLW strategy

The Low Level Waste Repository (LLWR) is the primary facility for disposal of Low Level Waste (LLW) in the United Kingdom (UK), serving the UK nuclear industry and a diverse range of other sectors. Management of LLW in the UK historically was dominated by disposal to the LLWR. The value of the LLWR as a national asset was recognised by the 2007 UK Governmental Policy on management of solid LLW. At this time, analysis of the projected future demand for disposal at LLWR against facility capacity was undertaken identifying a credible risk that the capacity of LLWR would be insufficient to meet future demand if existing waste management practices were perpetuated. To mitigate this risk a National Strategy for the management of LLW in the UK was developed by the Nuclear Decommissioning Authority (NDA), partnered with LLW Repository Ltd. (the organisation established in 2008 to manage the LLWR on behalf of NDA). This strategy was published in 2010 and identified three mechanisms for protection of the capacity of LLWR - application of the Waste Hierarchy by waste producers; optimised use of existing assets for LLW management; and opening of new waste treatment and disposal routes to enable diversion of waste away from the LLWR. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Reuse of low specific activity material as part of LLWR design optimisation

A final cap will be emplaced over the disposed waste as part of the closure engineering for the UK's Low Level Waste Repository (LLWR). Additional profiling material will be required above the waste to obtain the required land-form. Consideration has been given to the potential opportunity to reuse Low Specific Activity Material (LSAM, defined as up to 200 Bq g{sup -1}) imported from other sites as a component of the necessary profiling material for the final repository cap. Justification of such a strategy would ultimately require a demonstration that the solution is optimal with respect to other options for the long-term management of such materials. The proposal is currently at the initial evaluation stage and seeks to establish how LSAM reuse within the cap could be achieved within the framework of an optimised safety case for the LLWR, should such a management approach be pursued. The key considerations include the following: The LSAM must provide the same engineering function as the remainder of the profiling material. The cap design must ensure efficient leachate collection, drainage and control for Low Level Waste (LLW) (and, by extension, LSAM) during the Period of Authorisation. In the longer term the engineering design must passively direct any accumulating waters preferentially away from surface water systems. An initial design has been developed that would allow the placement of around 220,000 m{sup 3} of LSAM. The potential impact of the proposal has been assessed against the current Environmental Safety Case. (authors)

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Update on the review of the LLWR 2011 environmental safety case

The UK Low Level Waste Repository has submitted a fully revised Environmental Safety Case (ESC) to the Environment Agency for the continued operation of the site. The Environment Agency is reviewing the submission. As part of the review of the ESC, we have been engaging with the Environment Agency to answer questions and provide further clarification where required. Once the review is complete, LLWR will apply for a revised permit for the continued operation of the site. We are required by our current Permit to operate the site in accordance with the assumptions of the ESC. We have developed a process for the implementation and maintenance of the ESC as a 'live' safety case under formal change control, and the development of waste acceptance arrangements identified as necessary to ensure that the repository is operated in a safe and optimised way, consistent with the assumptions and results of the ESC. Engagement with waste consignors has been essential in the development of revised waste acceptance criteria. Additional work has also been carried out in the development of an Article 37 submission, presenting the ESC to local stakeholders and developing of waste emplacement strategies. (authors)

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HLRW management during MR reactor decommissioning in NRC 'Kurchatov Institute'

A program of decommissioning of MR research reactor in the Kurchatov institute started in 2008. The decommissioning work presumed a preliminary stage, which included: removal of spent fuel from near reactor storage; removal of spent fuel assemble of metal liquid loop channel from a core; identification, sorting and disposal of radioactive objects from gateway of the reactor; identification, sorting and disposal of radioactive objects from cells of HLRW storage of the Kurchatov institute for radwaste creating form the decommissioning of MR. All these works were performed by a remote controlled means with use of a remote identification methods of high radioactive objects. A distribution of activity along high radiated objects was measured by a collimated radiometer installed on the robot Brokk-90, a gamma image of the object was registered by gamma-visor. Spectrum of gamma radiation was measured by a gamma locator and semiconductor detector system. For identification of a presence of uranium isotopes in the HLRW a technique, based on the registration of characteristic radiation of U, was developed. For fragmentation of high radiated objects was used a cold cutting technique and dust suppression system was applied for reduction of volume activity of aerosols in air. The management of HLRW was performed by remote controlled robots Brokk-180 and Brokk-330. They executed sorting, cutting and parking of high radiated part of contaminated equipment. The use of these techniques allowed to reduce individual and collective doses of personal performed the decommissioning. The average individual dose of the personnel was 1,9 mSv/year in 2011, and the collective dose is estimated by 0,0605 man x Sv/year. Use of the remote control machines enables reducing the number of working personal (20 men) and doses. X-ray spectrometric methods enable determination of a presence of the U in high radiated objects and special cans and separation of them for further spent fuel inspection. The sorting of radwaste enabled shipping of the LLRW and ILRW to special repositories and keeping of the HLRW for decay in the Kurchatov institute repository. (authors)

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The role of large scale demonstration experiments in supporting the implementation of a high level waste programme

Large scale demonstration experiments in underground research laboratories (both onsite and off-site) are currently undertaken by most high level radioactive waste management organisations. The decision to plan and implement prototype experiments, which might have a life of several decades, has both important strategic and budgetary consequences for the organisation. Careful definition of experimental objectives based on the design and safety requirements is critical. The implementation requires the involvement of many parties and needs flexible but consequent management as, for example, additional goals for the experiments, identified in the course of the implementation, might jeopardise initial primary goals. The outcomes of an international workshop in which European and Japanese implementers (SKB, Posiva, Andra, ONDRAF, NUMO and Nagra) but also certain research organisations (JAEA, RWMC) participated identified which experiments are likely to be needed depending on the progress in implementing a disposal programme. Already earlier in a programme, large scale demonstrations are generally performed aiming at reducing uncertainties identified during the safety case development such as thermo-hydraulic-mechanical process validation in the engineered barrier system and target host rock. Also feasibility testing of underground construction in a potential host rock at relevant depth might be required. Later in a programme, i.e., closer to the license application, large scale experiments aim largely at demonstrating engineering feasibility and performance confirmation of complete repository components. Ultimately, before licensing repository operation, 1:1 scale commissioning testing will be required. Factors contributing to the successful completion of large scale demonstration experiments in terms of planning, defining the objectives, optimising results and main lessons learned over the last 30 years are being discussed. The need for international coordination in defining the objectives of new large scale demonstration experiments is addressed. The paper is expected to provide guidance to implementing organisations (especially those in their early stages of the programme), considering participating in and/or or conducting on their own large scale experiments in the near future. (authors)

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Stakeholder opinions on the use of the added value approach in siting radioactive waste management facilities

In some countries nuclear waste facility siting programs include social and economic benefits, compensation, local empowerment and motivation measures and other incentives for the potential host community. This can generally be referred to as an 'added value approach'. Demonstration of the safety of a repository is seen as a precondition of an added value approach. Recently much focus has been placed on studying and developing public participation approaches but less on the use of such incentive and community benefit packages, although they are becoming a more common element in many site selection strategies for nuclear waste management facilities. The primary objective of this paper is to report on an ongoing study of stakeholders' opinions of the use of an added value approach in siting a radioactive waste facility in the Czech Republic, Poland and Slovenia. The paper argues that an added value approach should adapt to the interests and needs of stakeholders during different stages of a siting process. The main question posed in the study is as follows: What are the measures which should be included in 'added value approach' according to the stakeholders? The research data consists of stakeholders' responses to a survey focusing on the use of added value (community benefits) and incentives in siting nuclear waste management facilities. The survey involved use of a questionnaire developed as part of the EU-funded IPPA* project in three countries: the Czech Republic, Poland and Slovenia. (* Implementing Public Participation Approaches in Radioactive Waste Disposal, FP7 Contract Number: 269849). The target audiences for the questionnaires were the stakeholders represented in the national stakeholder groups established to discuss site selection for a nuclear waste repository in their country. A total of 105 questionnaires were sent to the stakeholders between November 2011 and January 2012. 44 questionnaires were returned, resulting in a total response rate of 41% (10/29 in the Czech Republic, 11/14 in Poland and in 23/64 in Slovenia). (authors)

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D and D knowledge management information tool - a web based system developed to share D and D knowledge worldwide

Deactivation and decommissioning (D and D) work is a high risk and technically challenging enterprise within the U.S. Department of Energy complex. During the past three decades, the DOE's Office of Environmental Management has been in charge of carrying out one of the largest environmental restoration efforts in the world: the cleanup of the Manhattan Project legacy. In today's corporate world, worker experiences and knowledge that have developed over time represent a valuable corporate asset. The ever-dynamic workplace, coupled with an aging workforce, presents corporations with the ongoing challenge of preserving work-related experiences and knowledge for cross-generational knowledge transfer to the future workforce [5]. To prevent the D and D knowledge base and expertise from being lost over time, the DOE and the Applied Research Center at Florida International University (FIU) have developed the web-based Knowledge Management Information Tool (KM-IT) to capture and maintain this valuable information in a universally available and easily accessible and usable system. The D and D KM-IT was developed in collaboration with DOE Headquarters (HQ), the Energy Facility Contractors Group (EFCOG), and the ALARA [as low as reasonably achievable] Centers at Savannah River Sites to preserve the D and D information generated and collected by the D and D community. This is an open secured system that can be accessed from https://www.dndkm.org over the web and through mobile devices at https://m.dndkm.org. This knowledge system serves as a centralized repository and provides a common interface for D and D-related activities. It also improves efficiency by reducing the need to rediscover knowledge and promotes the reuse of existing knowledge. It is a community-driven system that facilitates the gathering, analyzing, storing, and sharing of knowledge and information within the D and D community. It assists the DOE D and D community in identifying potential solutions to their problem areas by using the vast resources and knowledge base available throughout the global D and D community. The D and D KM-IT offers a mechanism to the global D and D community for searching relevant D and D information and is focused on providing a single point of access into the collective knowledge base of the D and D community within and outside of the DOE. Collecting information from subject matter specialists, it builds a knowledge repository for future reference archiving Lessons Learned, Best Practices, ALARA reports, and other relevant documents and maintains a secured collaboration platform for the global D and D community to share knowledge. With the dynamic nature and evolution of the D and D knowledge base due to multiple factors such as changes in the workforce, new technologies and methodologies, economics, and regulations, the D and D KM-IT is being developed in a phased and modular fashion. (authors)

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Characterization of radioactive wastes with respect to harmful materials

In addendum 4 to the license of the German KONRAD repository, which considers mainly radiological aspects, a water law permit was issued in order to prevent the pollution of the near-surface groundwater. The water law permit stipulates limitations for 10 radionuclides and 2 groups of radionuclides as well as mass limitations for 94 substances and materials relevant for water protection issues. Two collateral clauses, i.e. additional requirements imposed by the licensing authority, include demands on the monitoring, registering and balancing of non-radioactive harmful substances and materials /1/. In order to fulfill the requirements of the water law permit the German Federal Office for Radiation Protection (BfS) being the operator of the KONRAD repository has developed a concept, which ensures the compliance with all requirements of the water law permit and which provides standardized easy manageable guidance for the waste producers to describe their wastes. On 15 March 2011 the competent water authority, the 'Niedersaechsischer Landesbetrieb fuer Wasserwirtschaft, Kuesten- und Naturschutz' (NLWKN) issued the approval for this concept. Being the most essential part of this concept the procedural method and the developed description of nonradioactive waste package constituents by use of standardized lists of materials and containers is addressed and presented in this paper. The waste producer has to describe his waste package in a standardized way on the base of the lists of materials and containers. For each material in the list a comprehensive description is given comprising the composition, scope of application, quality control measures, thresholds and other data. Each entry in the list has to be approved by NLWKN. The scope of the lists is defined by the waste producers' needs. Using some particular materials as examples, the approval procedure for including materials in the list is described. The procedure of describing the material composition has to be considered in the KONRAD waste acceptance requirements. The respective part of these requirements will be introduced. In order to clarify the procedure of describing waste packages by use of the standardized lists of materials and containers some examples of typical waste package descriptions will be presented. (authors)

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Overview of the United States department of energy's used fuel disposition research and development campaign

The United States Department of Energy (US DOE) is conducting research and development (R and D) activities within the Used Fuel Disposition Campaign (UFDC) to support storage, transportation, and disposal of used nuclear fuel (UNF) and wastes generated by existing and future nuclear fuel cycles. R and D activities are ongoing at nine national laboratories, and are divided into two major topical areas: (1) storage and transportation research, and (2) disposal research. Storage R and D focuses on closing technical gaps related to extended storage of UNF. For example, uncertainties remain regarding high-burnup nuclear fuel cladding performance following possible hydride reorientation and creep deformation, and also regarding long-term canister integrity. Transportation R and D focuses on ensuring transportability of UNF following extended storage, addressing data gaps regarding nuclear fuel integrity, retrievability, and demonstration of subcriticality. Disposal R and D focuses on identifying multiple viable geologic disposal options and addressing technical challenges for generic disposal concepts in various host media (e.g., mined repositories in salt, clay/shale, and granitic rocks, and deep borehole disposal in crystalline rock). R and D will transition to site-specific challenges as national policy advances. R and D goals at this stage are to increase confidence in the robustness of generic disposal concepts, to reduce generic sources of uncertainty that may impact the viability of disposal concepts, and to develop science and engineering tools that will support the selection, characterization, and ultimately licensing of a repository. The US DOE has also initiated activities that can be conducted within the constraints of the Nuclear Waste Policy Act to facilitate the development of an interim storage facility and supporting transportation infrastructure. (authors)

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Radionuclide inventory of vitrified waste after spent nuclear fuel reprocessing at La Hague

The primary aim of the present work was to determine the inventories of the radionuclides and stable elements in vitrified high-level waste produced at La Hague and delivered to Germany, which are of importance for long-term safety assessment of final repositories for radioactive wastes. For a subset of these radionuclides and stable elements, the inventories were determined - either by direct measurements or by involving established correlations - and reported by AREVA. This allowed verification of the validity of application of a model approach utilizing the data of burnup and activation calculations and auxiliary information on the reprocessing and vitrification process operated at La Hague. Having proved that such a model approach can be applied for prediction of inventories of actinides, fission and activation products in vitrified waste, the present work estimated the minimum, average and maximum inventories of the radionuclides, which are of importance for long-term safety assessment of final repositories for radioactive waste but were not reported by AREVA for delivered CSD-V canisters. The average and maximum inventories in individual CSD-V canisters predicted in the present approach were compared to the inventories predicted by Nagra for canisters with vitrified waste delivered from La Hague to Switzerland [1]. This comparison revealed a number of differences between these inventories despite the fact that the canisters delivered to Switzerland were produced in essentially the same way and from the common reprocessing waste stock as CSD-V canisters delivered to Germany. Therefore, a further work is required in order to identify the reason for the discrepancy in the present estimation versus the Nagra estimation [1]. Such a work should also address the recommendation by the international peer review of the Safety Report of the Project Opalinus Clay to obtain estimates of the inventories of long-lived mobile radionuclides (such as {sup 14}C, {sup 36}Cl, {sup 79}Se, and {sup 129}I), which contribute most to the dose estimates in the radiological safety assessments, if possible, in agreement with other countries with similar waste streams in order for a coordinated set of data to be generated [2]. Since vitrified waste from reprocessing of spent nuclear fuel at La Hague was delivered to several countries - Belgium, France, Germany, Japan, Netherlands, and Switzerland - an international effort can be recommended. (authors)

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