bluuebunny/crossref_metadata_2025_split huggingface dataset:
Functional Requirements and Performance Assessment: Application to the Case of Spent Fuel Disposal in Clay
Authors: Marivoet Jan , Sillen Xavier , De Preter Peter 
Published: April 2020

Abstract
Geological repository systems for the disposal of radioactive waste are based on a multi-barrier design. Individual barriers contribute in different ways to the overall long-term performance of the repository system, and furthermore, the contribution of each barrier can considerably change with time. In a systematic analysis of the functional requirements for achieving long-term safety a number of basic safety functions can be defined: physical confinement, retardation / slow release, dispersion / dilution and limited accessibility.
In the case of the geological disposal of spent fuel in a clay formation a series of barriers are designed or chosen to contribute to the realisation of the basic safety functions. The physical confinement is realised by the watertight, high-integrity container, which prevents contact between groundwater and the confined radionuclides. In first instance the retardation / slow release function is realised by the slow dissolution of the waste matrix and by the limited solubility of many elements in the near field. However, the natural clay barrier provides the main contribution to this safety function. The migration of radionuclides through the Boom Clay is mainly due to molecular diffusion, which is an extremely slow process. Furthermore, many elements are strongly sorbed by the clay minerals what makes their migration even much slower. The dispersion / dilution function mainly occurs in the aquifer and the rivers draining the aquifer in the surroundings of the disposal system. Various performance indicators are used to quantify the contributions of each safety function and to explain the functioning of the repository system.
Simulation Model of Regional and Urban Development Impacts: An Approach for Transparent and Communicative LILRW Disposal Siting Process
Authors: Golobic Mojca 
Published: April 2020

Abstract
In most of the failed nuclear objects sitting processes the main barriers were not the technical or safety questions, but the problem of communication and reaching the public acceptance about decision. The paper presents one of the possible approaches that could contribute to the more community sensitive, transparent, communicative and finally successful siting process. The computer-supported approach for assessment of LILRW disposal impacts on regional and urban development is one of several modules loosely connected to a spatial simulation model in GIS. The results are shown as cartographic representation of impacts on different land uses and a possible future land use pattern. This model will be used for evaluation of alternative sites from regional and urban development impact aspects and for negotiating compensation schemes for local communities. One of the very important findings of this work is that the modeling outcomes depend more on value assumptions, according to which the input parameters are chosen, than on accuracy of computations. Therefore any conclusion based on the model outcomes should be verified in a wide panel of experts, interest groups and also wider public.
Managing the Environmental Impacts of Low Activity Wastes From the South African Gold Mining Industry
Authors: Wymer Denis G. , Botha Johan C. 
Published: April 2020

Abstract
Uranium mineralization is associated with the gold deposits in South Africa that have been mined for more than a century. Investigations of the radiological impacts on the environment reveal that the various radioactive wastes — mostly of low activity — associated with gold and by-product uranium production do not warrant the application of anything more than basic control measures. Non-radiological pollution of the water environment is a growing problem, however, aggravated by the closure and flooding of mines. Control measures to address this problem should, as a spin-off, limit the release of radionuclides, thus helping to control future radiological pollution.
Testing Permeable Reactive Barrier Media for Remediation of Uranium Plumes in Groundwater
Authors: Mallants Dirk , Moors Hugo , Wang Lian , Maes Norbert , Vandenhove Hildegarde , Diels Ludo , Bastiaens Leen , Vos Johan 
Published: April 2020

Abstract
In-situ treatment of contaminated groundwater by means of permeable reactive barriers (PRBs) is becoming a cost-effective remediation technique. Various reactive materials that might be used in PRBs were tested in their ability to remove uranium from groundwater. Materials tested include ferric oxyhydroxides, coarse- and fine-grained zero-valent iron, aluminium-iron oxides, and zeolites. Batch tests were used to evaluate the removal efficiency of these materials. To analyse the effect of groundwater composition on the interaction between dissolved uranium and reactive materials, two types of groundwater were used, mainly differing in carbonate content and pH. Considering an equilibration time of 48 hours and initial uranium concentrations between 2.4 and 24 mg/1, finegrained zero-valent iron proved to be most effective with a uranium removal efficiency of more than 96% for carbon-rich groundwater and 99% for carbon-poor groundwater. Intermediate efficiency was observed for coarsegrained zero-valent iron and aluminium-iron oxides. Less than 10% of the dissolved uranium was adsorbed on the iron oxyhydroxides. Zeolites did not remove any uranium from solution. Results further indicated a positive correlation between dissolved inorganic carbon content and dissolved uranium at equilibrium. Because it can be easily obtained at a fairly low price, zero-valent iron is a promising material for use in PRBs.
Investigation of the Degradation of Bituminized Products by the Stored
Authors: Yoshiyuki Ito , Toshiyuki Kondo , Yasumi Murayama , Noboru Kogawa 
Published: April 2020

Abstract
Bituminized products, that are soldified the low radioactive liquid wastes from Tokai Reprocessing Plant, may degrade by the progress of the stored years. Moreover, under disposal condition, it is considered that the bituminized products contact groundwater.
Therefore, it was investigated the physical properties and the gamma radionuclides of the stored products. The physical property of bituminized products did not change in about 10 years storage. And also, the leaching test was carried out, it was confirmed that the leaching of the salts was not easy to release from the bituminized product by the progress of the stored years.
Clearance/Exemption Levels for Slightly Radioactive Materials Arising From Nuclear Decommissioning and Non-Nuclear Industries
Authors: Pescatore Claudio , Teunckens Lucien , Menon Shankar 
Published: April 2020

Abstract
The management of the relatively large volumes of slightly radioactively contaminated material, arising from the decommissioning of nuclear facilities, represents a substantial fraction of the cost of such projects. The recycling of a relevant fraction of this material (or its reuse or disposal) without radiological restrictions, was identified by a Task Group of the OECD/NEA Co-operative Programme on Decommissioning, as a significant means of reducing such costs. The lack of internationally accepted “clearance levels” of radioactivity, at which the material could be utilised without radiological restrictions, seriously limits recycling as a waste management option.
The emergence of the NORM/TENORM issue is of great significance for the discussion of clearance regulations. TENORM arisings occur in huge quantities, two to three orders of magnitude larger than those used in European studies on recycling in the nuclear industry, and the activity levels are generally the same as in very low to low-level nuclear waste. The regulation of TENORM is in its early stages. Their occurrence in a large number of industries, as well as their activity levels and quantities, has not been generally appreciated, even by regulatory authorities, until fairly recently.
National and international bodies have suggested or are in the process of suggesting regulations for TENORM. The most important development is the publication of the European Commission Directive of May 1996 (ratified in May 2000) laying down basic safety standards for protection against ionising radiation, arising both in the nuclear and non-nuclear industries. The International Atomic Energy Agency has also started looking into this area in connection with the revision of its Safety Series 89 document. Significant to note is that both these bodies suggest release criteria into the general economy that are more relaxed for the radioactive materials from non-nuclear industries than for similarly contaminated material from nuclear industries. This issue is being taken up by several other bodies as well. This paper reviews the current debate and underlines the need for consistency in developing regulations and criteria for exemption and clearance of all radioactive materials regardless of their origin.
Political, Economic and Environmental Factors Impacting the Realization of the Plutonium Utilization Strategy in Russia
Authors: Chebeskov Alexander N. , Oussanov Vladimir I. , Iougai Serguei V. , Zavadsky Mikhail I. , Decressin Albert , Vanden Bemden Emile 
Published: April 2020

Abstract
The paper presents some results of system analysis of the plutonium utilization alternatives carried out by Russian specialists with an assistance of their European colleagues in the frames of ISTC#369 and #1443 Projects, financially supported by European Commission states. The multi-attribute utility (MAU) model was chose as the methodology for the analysis. For the time being the work is under continuation, but results of the first stage of the study related to conceptual features of the problem can be summed up and discussed.
The study made has shown that effective application of a quantitative model to the problem of plutonium utilization is possible provided that political and technical objectives are strictly defined. A realistic base for the scenarios under consideration is general requirement of a principal importance. The principal result of system consideration is the conclusion that transition from the practice of plutonium accumulation and storing to the strategy of plutonium utilization in nuclear reactors is justified under the real Russian conditions. In an economic sense it is justified by the reason that the expenses of reactor conversion to MOX fuel utilization are comparable with those of long-term storing of plutonium, taking into account modern, and especially future, safety requirements. On the non-proliferation criterion, conversion to MOX fuel utilization is also justified, so as in this case the main source of the proliferation risk — the plutonium store, where plutonium is being stored in the forms being the most attractive for the non-authorized usage — is eliminated. At last, conversion to utilization of stored plutonium results in the decrease of radiation risk for staff and population, as well as in the decrease of impact on the environment due to reduction of some stages of the fuel cycle.
Ranking of the plutonium utilization alternatives on a base of account economic, non-proliferation and ecological factors has led to reasonable, explicable results. These results were used in “The concept of the Russian Federation on_disposition of weapons plutonium being released in the course of nuclear disarmament”, which was approved as an official document of Ministry of Atomic Energy of Russian Federation. The main approaches of this strategy were changed significantly when comparing with the previous concept of plutonium utilization. One may expect more detailed recommendations to decision makers in course of further elaboration of scenarios and models at the following stages of the study.
Operation Experience of HALW Storage Process at the Tokai Reprocessing Plant
Authors: Nakamura Yoshinobu , Suda Shizuka , Ishiyama Koich , Watahiki Masaru , Mutoh Hideyo 
Published: April 2020

Abstract
Volume of nitric acid solution, which contain the most of fission products (FPs), is concentrated to 0.5–2m3 in an HALW evaporator by reprocessing a spent fuel of ltU. The HALW is stored in HALW storage tanks temporarily till it is transferred to the Tokai Vitrification Facility (TVF). During the storage, the HALW of 1–8m3/y vaporized per storage tank. A shift coefficient of radioactive nuclides from the HALW to off-gas was 4E−11. Through the operation experience, knowledge obtained about storage management of highly active liquid waste (HALW) is reported.
High Pressure Abrasive Water Jet Cutting As a Dismantling Tool
Authors: Denissen L. , Massaut V. , Klein M. , Dadoumont J. , Davain H. 
Published: April 2020

Abstract
Since 1989, SCK•CEN is dismantling the BR3 (Belgian Reactor no3). During these 11 years of decommissioning, a lot of experience is gained in the field of remotely controlled cutting techniques for high active parts and decontamination techniques of highly contaminated parts.
Mechanical cutting techniques (circular saw and band saw) were used for cutting the reactor Pressure Vessel (RPV) and its internals. For some large components in the reactor building, like the steam generator, the pressurizer and the Neutron Shield Tank, a new tool, the High Pressure Abrasive Water Jet Cutting will be used to cut activated and contaminated pieces with material thicknesses ranging from 20 to 170 mm. Deliverance of the cutting equipment is foreseen in the summer of 2001. Cold tests and training sessions for the operators will then take place and by the beginning of 2002 the cutting equipment will be installed in the reactor pool inside the controlled area to start the actual work.
Windscale Advanced Gas-Cooled Reactor Decommissioning: Hot Gas Manifold Dismantling Strategy and Tooling
Authors: Dixon Colin , Bloor Roger C. , Chester Stephen , Benest Terence 
Published: April 2020

Abstract
The Windscale Advanced Gas-Cooled Reactor (WAGR) is the UK’s demonstration project for power reactor decommissioning. As part of the SAFER programme, funded by the DTI, UKAEA have contracted BNFL Magnox to implement the work. The reactor is being dismantled on a ‘top-down’ basis and a major current task is to dismantle the hot gas manifold (hotbox).
AEA Technology was contracted to supply equipment to size reduce and remove the hotbox. The strategy makes use of manual, semi-remote and remote techniques. Semi-remote tools were built and deployed and then set to work with the assistance of personnel entering the radiation area. The equipment was then operated remotely from a control room located in an adjacent building.
Reversibility and Retrievability Principles in the Czech Disposal Concept
Authors: Woller Frantisek , Nachmilner Lumir 
Published: April 2020

Abstract
Reversibility and retrievability are topics that have recently been intensively discussed both within different national programs and at several international waste management events. While reversibility is a technical, iterative tool allowing optimization of the process of deep repository development, retrievability is considered to be a more political issue, a statute for which is not included in Czech legislation. However, both principles have been more or less incorporated into geological repository development and have also been considered as options for existing sub-surface repositories for institutional wastes. This paper summarizes Czech approaches to these issues.
Retrievability, Reversibility and Monitoring of Geological Disposal in France
Authors: Guillaume S. , Mouroux Bernard , Hoorelbeke J.-M. 
Published: April 2020

Abstract
The reversibility for repositories is frequently quoted in the media, and opponents to nuclear energy or to waste burial often justify their opposition by the impossibility of reversibility in deep geological formations, as an echo to the supposed absence of reversibility in everything associated with nuclear energy in general.
Reversibility responds to various motivations: in case of error or of unpredicted events, it must be possible to intervene; inversely, it must also be possible to recycle certain materials many years after disposal; due to a potential lack of confidence, it must finally be possible to progress in a stepwise approach in order to ensure that every generation may orient the process.
That approach leads to a wide definition of reversibility no longer restricted to demonstrating that it is possible to remove a package technically (“retrievability”), but rather encompassing the possibility to modify the process at each step throughout the lifetime of the repository.
Confronted with this definition for reversibility, the preliminary design options of repositories selected for the Meuse/Haute-Marne site in France appear at this stage to be sufficiently flexible to allow further progress in the feasibility study of a reversible repository.
Four main areas of study need to be emphasised:
• It is necessary to further the knowledge concerning the evolution of a repository: What phenomena control the evolution of the repository? What characteristics of the different components do they affect?
• It is necessary to define reversibility levels and their specific characteristics while modulating them, if necessary, according to the different waste types;
• It is necessary to define a monitoring programme in close relationship with the phenomena modelling ensuring the representation of the repository evolution over time and the analysis of the safety conditions during each phase;
• Finally is it necessary to examine the effects of the different states of the repository on the implementation of technologies: for example, what characteristics or what state of the components may complicate package retrieval, in other words make the proposed technological means inadequate to intervene at each reversibility level?
Leaching of Iodide Ion From BiPbO2I Under Reducing Conditions
Authors: Amaya T. , Mukunoki A. , Shibuya M. , Kodama Hiroshi 
Published: April 2020

Abstract
Radioactive wastes containing Iodine-129 are to be disposed of in Japan in an underground facility together with TRU wastes. Iodine-129 has a long half-life (1.6 × 107 y) and it is strongly adsorbed in the thyroid gland when it intrudes into the human body. The main chemical formulae of iodine in an alkaline solution are I− and IO3−, and these anion species are absorbed only to a very small extent on silicate minerals. Iodine-129, therefore, is one of the key nuclides to be studied in the geological disposal of radioactive wastes. Recently, a new inorganic ion-exchanger, BiPbO2NO3, has been developed which reacts with iodide ions in a solution by forming BiPbO2I (BPI). The leach resistance of BPI encapsulated in cement (BPIC) was studied in a solution under geological conditions.
The leaching experiment was carried out in an inert glove box in which the concentrations of oxygen and carbon dioxide were maintained at less than 1 ppm. Pure water was degassed 12 hours prior to use. Two kinds of solution were prepared: one was low salinity solution (RW), and the other was high salinity solution (SW). Leachants were prepared by adding a reductant (N2H4) to each solution and pH was adjusted to a fixed value. BPIC was mixed with the leachant in a plastic container. The container was shaken continuously at ambient temperature for six months. The concentrations of iodide ions, bismuth ions and lead ions in the leachant were analyzed periodically using ICP-AES.
Limited numbers of iodide ions (2%–4%) were released from BPIC in the initial period of leaching, following which no additional release of iodide ion was observed for six months. No significant difference was observed in the X-ray diffraction patterns of BPI in BPIC before and after the experiment. These results indicate that iodine is fixed tightly in BPIC. A mechanism of the leaching resistance is discussed.
Methods of Radical Regeneration of Spent Solvent Containing TBP and Diluent
Authors: Zilberman Boris Ya. , Makarychev-Mikhailov Mikhail N. , Saprykin Vladimir F. , Sakulin Stanislav V. , Dulepov Yury N. , Glushko Vladimir V. 
Published: April 2020

Abstract
Methods of steam-supported distillation of spent solvent containing 30% tributyl phosphate (TBP) in various diluents have been developed. Single-stage and two-stage processes were tested. Single-stage distillation can be realized either with the use of superheated steam under vacuum or with saturated steam at atmospheric pressure, ensuring transportation of all the solvent to distillate, independently of the used diluent. Two-stage process allows separating TBP and diluent, both of them being purified, when the diluent boiling temperature is 50–70°C below TBP boiling point. At the first stage of the process, pure diluent is obtained through distillation at atmospheric pressure with the use of saturated steam. Then TBP is distilled off from wet bottoms of the first stage by continuous-flow vacuum distillation at the circulating bottoms.
Sodium Carbonatation Process Development in a View of Treatment of the Primary Circuit of Liquid Metal Fast Reactor (LMFR) in Decommissioning Phases
Authors: Rodriguez G. , Gastaldi O. 
Published: April 2020

Abstract
As soon as the decision of the final shutdown of SUPERPHENIX was taken, studies linked to the final shutdown and the decommissioning of this reactor were taken into considerations. Amongst all this studies, one concerns the final treatment of the remaining sodium of the primary vessel of the reactor (estimated to a maximum of six tons after a fully draining of the sodium).
The process envisaged is an in situ treatment called the carbonatation process.
This papers describes the principle of this process and presents CEA experimentations done to precise the kinetic of this process as a function of different parameters (temperature, percentage of carbon dioxide, percentage of water, etc).
Packaging and Conditioning of Radioactive Oil Wastes From Cernavoda NPP Operation
Authors: Deneanu Nicoleta T. , Popescu Ion V. , Dulama Mirela F. , Baboescu Elena I. 
Published: April 2020

Abstract
Radioactive oil wastes produced at Cernavoda NPP consist of lubricating oils from primary heat transport pumps, hydraulic fluids from fuelling machines and turbine oils. These are normally LLW containing only relatively small quantities of beta/gamma-emitting radionuclides. The main contamination pathway is the contact with the reactor primary coolant. The waste contains varying amounts of tritium with activity below 108Bq/1.
The objective of solidification is to convert radioactive waste into a stable monolithic form, which minimizes the probability to release tritium in the environment during interim storage, transportation and final disposal. The solidified waste that leaves the Waste Management facility from INR Pitesti should be of such chemical, mechanical, thermal and radiolytic stability as to assure its integrity over the time required for the decay of the contained radionuclides to an acceptable level. Laboratory studies concerning the cementation of radioactive oil waste have been undertaken at the Institute for Nuclear Research Pitesti.
A quality assurance program should accompany the production of waste forms. The goal of all tests should be to obtain a license for a certain process from a competent authority.
Criteria that might be applied in the selection of treatment process and of waste storage and/or disposal are the complexity of volume reduction, the complexity of the technology and equipment, the level of process development, and the regulatory requirements.