bluuebunny/crossref_metadata_2025_split huggingface dataset:
HUP Immobilization of Uranium and Plutonium From Surrogate Liquid Radioactive Solutions Using Porous Molded Glass Crystalline Blocks (Gubka)
Authors: Aloy A. S. , Sapozhnikova N. V. , Anshits A. G. , Sharonova O. M. , Tretyakov A. A. , Knecht D. A. , Tranter T. J. , Macheret J. 
Published: April 2020

Abstract
This paper describes test results for immobilization of surrogate radioactive actinide solutions. A glass crystalline porous material, Gubka, or “sponge” in Russian, was tested to immobilize surrogates of two types of actinide solutions, which exist in the DOE complex. Solutions using uranium as a surrogate for plutonium were stabilized in the porous Gubka block by successive saturation and drying cycles. After calcining, fully dense glass-ceramic tablets were formed in a hot uniaxial press. The tablets included 2–3 different glass phases with no uranium content and a uranium phase. Detailed electron micrographs and X-ray diffraction results are presented in the paper.
New or Modified Old Conceptual Approach to HLLLW Management
Authors: Ninkovic Marko M. , Raicevic Jagos J. 
Published: April 2020

Abstract
One of the greatest challenges in the use of nuclear energy is the high radioactive long-lived waste which is generated during production. It must be dealt with safely and effectively. While technical solutions exist, including deep geological repositories, progress in the disposal of radioactive waste has been influenced, and in many cases delayed, by public perceptions about the safety of the technology. One of the primary reasons for this is the long life of many of radionuclides, actinides and fission products, with half-lives on the order of a hundred thousand to a million years. Problems of perceptions could be reduced significantly, according to our and many others author’s opinion, if there were a way to burn or destroy the most toxic long-lived radioactive wastes. As there are no industrial methods for waste destroying today, in this paper it was suggested a new hybrid approach: instead of final waste disposal, long-termed but yet temporal storage only, striving towards final destruction once a appropriate conditions are maintained. This new or modified old approach could affect current HLLLW management and related activities in: changes of processing technology; prolonging the time period of waste storage at temporal depositories; increasing the investment into research regarding the methods and technologies for destructions of these materials, and slowing down the investments into the very expensive final disposal repositories. It’s authors opinion that such conceptual approach would contribute the reviving interest in nuclear energy, all over the world and especially in small and developing countries.
Recent Developments in D&D of Nuclear Facilities in Spain
Authors: Santiago Juan L. , Vidaechea Sergio 
Published: April 2020

Abstract
Spain occupies a leading position at international level in the field of installation decommissioning. Decommissioning projects have already been performed in relation to uranium mills, the rehabilitation of disused uranium mines is currently in the final phase and the dismantling of the Vandellós-I Nuclear Power Plant is now under way.
On the basis of this experience, this paper describes the key issues in decommissioning technology and presents the approaches adopted by ENRESA to tackle the decommissioning strategy in Spain. In particular practical dismantling and decontamination methods are described, and material and radioactive waste management are discussed.
Superphénix LMFR Decommissioning Sodium Retention Treatment by Mechanical Drilling of Internal Structures
Authors: Baque F. , Haudiquet C. , Major P. 
Published: April 2020

Abstract
For the decommissioning of the French 1200 MWe Liquid Metal Fast Reactor SUPERPHENIX, twenty-five processes were selected to pierce liquid sodium trapping zones after draining of the main vessel. Four processes were chosen and closely studied: mechanical drilling, laser shooting, high pressure liquid cutting, replacement of liquid sodium by another heavier liquid.
Mechanical drilling was finally selected: some in-air tests at room temperature were performed with conventional machines, then some in-sodium tests at 180°C, using a prototype machine. The drilling conditions and tools, the chip recovering sequence and the efficiency of the scenario (four successive drilling operations in three piled plates) were validated: the liquid sodium was drained off.
Decommissioning the Georgia Tech Research Reactor
Authors: Eby Robert , Lundberg Lark , Marske Steve , Hertel Nolan , Ice Rod 
Published: April 2020

Abstract
The Georgia Tech Research Reactor (GTRR) is a 5-megawatt (MW) heavy-water-cooled nuclear reactor located on the Georgia Institute of Technology (Georgia Tech) campus in downtown Atlanta, Georgia. On July 1, 1997, Georgia Tech administration notified the U.S. Nuclear Regulatory Commission (NRC) of their intent to decommission the GTRR. In the summer of 1999, the NRC issued a license amendment to decommission the GTRR in accordance with NRC’s Regulatory Guide 1.86. In the spring of 1999, Georgia Tech and the State of Georgia contracted CH2M HILL to serve as the Executive Engineer to manage the decommissioning project. Later in the summer of 1999, the IT Corporation was selected as the Decommissioning Contractor. The Decommissioning Contractor began the dismantlement process at the Georgia Tech site in November, 1999. By February, 2000, reactor support systems such as the primary and secondary cooling water systems, and the bismuth cooling system were removed and packaged for off-site disposal. Reactor internals were removed in April, 2000. Removal of the bioshield occurred from May through November, 2000. Throughout January, 2001, various concrete structures, including the Spent Fuel Storage Hole, were decontaminated. Dismantlement and decontamination activities were completed by April, 2001. The Final Survey Report to the NRC is planned to be submitted to the NRC December, 2001, 2001. Final license termination by the NRC is anticipated in the spring of 2002.
The Decontamination and Decommissioning of a 2 000 m3 Open Storage Pond for Liquid Waste and Sludges
Authors: Walthéry Robert , Teunckens Lucien , Lewandowski Patrick , Millen Danny , Baumann Sven 
Published: April 2020

Abstract
In the beginning of April, 1998, Belgoprocess started the large scale decommissioning of shutdown facilities at the former waste treatment department of the Belgian Nuclear Research Centre (SCK•CEN).
In the second half of 2000, the decommissioning of a former open concrete storage pond was started. It has two parts, one with a volume of 200 m3, the other has a volume of 2 000 m3. In the past, the tanks were used as open cooling reservoir for the scrubber water of an incinerator for low level beta-gamma wastes. As a result, it contained some 126 Mg of sludges that were treated in a bituminisation facility.
After emptying and rinsing of the tanks, samples were taken from the inner (brick) and the outer (concrete) wall to define the contamination levels in the material. It was found that the first layer of about one centimetre of the brick wall showed some radium contamination. Samples taken from the outer wall showed no contamination, which does not mean that the entire outer wall should be free of contamination. It was seen that the outer wall had some cracks, and there was no guarantee that a layer of bitumen in between the two walls was still in good condition. Also the top layer of the concrete bottom proved to be radium contaminated.
The outer walls of the storage pond were supported by a thick layer of soil, which proved to contain some spots of contamination. The contaminated soil was segregated and evacuated, the remaining part was removed to a temporary storage area, waiting to be monitored for final unconditional release.
The paper gives an overview of the decommissioning activities relating to the open concrete storage pond at the former waste treatment department of the Belgian Nuclear Research Centre (SCK•CEN).
ATENA: Project of Storage and Disposal Plant for Radioactive Sodium Wastes
Authors: Rodriguez G. , Thomine G. , Serpantié JP. , Demoisy Y. 
Published: April 2020

Abstract
CEA, in the framework of its research and development activities in support of Liquid Metal Fast Reactors (LMFR), has developed experimental or demonstration plants, now at end of life, which are potential or existing sources of wastes containing sodium.
The sodium waste management of CEA forecasts the disposal of the sodium contained in these wastes by water and soda processes, with existing means for the non radioactive sodium, and with a specific plant, to be built, for radioactive sodium.
The project ATENA (Atelier de Traitement et d’Entreposage du Na Actif, Workshop for Processing and Storage of radioactive Na) answers to this need of central processing plant for radioactive sodium wastes generated by CEA.
It concerns only the wastes:
• containing radioactive sodium,
• transportable,
• which could not be processed by Phénix and Superphénix facilities for treatment of large primary or secondary sodium volumes, presently in project.
FRAMATOME ANP direction NOVATOME, French LMFR engineering, is chosen as project manager for ATENA.
ATENA will be a new plant with:
• a storage module, able to be a central storage for wastes held at present by several of CEA’s plants,
• a process module: disposal of sodium with several processes,
• a packaging module for wastes resulting from these treatments: packaging of soda and metallic pieces in solid ultimate wastes.
The three chosen disposal processes are existing and are validated. They apply to several wastes to be processed according to the activity of the contained sodium, to the capability of the sodium to be liquified or separated from its container:
• continuous and controlled injection of liquid sodium in soda (NOAH process),
• treatment of solid sodium aggregated with metallic pieces, by controlled and continuous spray of water (Enceinte de Lavage en Actif, Active Washing Enclosure),
• treatment of solid and very radioactive sodium, aggregated or not with metallic pieces, by non-controlled immersion in water (Autoclave).
We present the general concepts of ATENA plant, the process options and safety principles as chosen after the basic design phase.
Management of Mixed Hazardous and Radioactive Waste in the United States
Authors: Greenman William , Cole Kimberly 
Published: April 2020

Abstract
In the United States, mixed-waste is typically defined as waste that contains both radioactive constituents and non-radioactive constituents that pose a threat to human health or the environment (hazardous waste). Prior to 1986 the U.S. Nuclear Regulatory Commission (NRC) had sole regulatory authority over mixed-waste because of its radioactive constituents. In 1986, however, the U.S. Environmental Protections Agency (EPA) was granted regulatory authority over the hazardous constituents in mixed-waste; and, a system of dual regulation was created. Dual regulation of mixed-waste by the EPA and the NRC has caused significant problems for the regulated community. The burden of dual regulation has contributed to the slow development of treatment technologies, and to the overall lack of treatment capacity available to U.S generators of mixed-waste. This paper reviews the requirements that the EPA and the NRC mandate with regard to mixed-waste generation, treatment and disposal; and it explores technical impacts of those requirements as they relate to generators, treatment facilities and the public.
Some Results From Geotechnical Research on Bentonite
Authors: Pacovsky Jaroslav , Travnicek Radek , Vasicek Radek 
Published: April 2020

Abstract
A solution of safe isolation of radioactive waste represents a multidisciplinary problem. The waste isolation is to be provided by a multi-barrier system of an underground repository based on a system of several barriers. The system will consist of a container with high level radioactive wastes itself, an engineering barrier — currently intended to be based on ‘bentonites’ — and natural barrier formed with a suitable rock environment. Each of those barriers alone should prevent spreading of hazardous radionuclides from the container to the biosphere.
In its main part, this paper deals with laboratory research of some most important geotechnical requirements for the engineering barrier, assumedly based on bentonites materials. The group of tested properties contains:
• Hydrophysical properties - permeability, swelling ability, Atterberg’s consistency limits
• Physico-technical properties - thermal properties
• Rheology
Creation of mathematical and physical models forms an independent part of the complex research project for materials of the geotechnical part of the engineering barrier. In the conclusion a short consideration on a prepared physical model of container has been made.
Experimental Research of 14C Migration in Loess and Clay: Distribution Coefficients and Retardation Factors
Authors: Arsene Carmen , Negoiu Dumitru , Andrei Veronica 
Published: April 2020

Abstract
One of the priorities of Romanian program for radioactive waste management is the establishing of the emplacement medium of the low and intermediate radioactive waste repository.
Since 1994, the Saligny site, the potential location for low and intermediate repository in Romania, has been under physical, chemical, hydrological, mineralogical and radionuclide migration investigations.
The understanding and quantification of radionuclide transport in the geological medium is of primary importance with respect to safety and environmental impact assessment for prospective waste management site.
Because of its radiochemical characteristics, 14C, contained in low and medium radioactive wastes generated from operation of Cernavoda NPP Unit 1, require a special attention as concerns their disposal.
Laboratory experiments were conducted to determine the transport parameters of 14C that characterise its migration in the geological formations of the Saligny site.
Special Waste Generated by the Belgian Nuclear Research Centre
Authors: Van Alsenoy Veerle , Demeulemeester Yves , Noynaert Luc , Klein Michel , Lardot Nicolas 
Published: April 2020

Abstract
At research centres, test programs generate very particular waste forms, which cannot be treated using the known regular scenario’s and procedures, and therefore special studies need to evaluate safe handling of these special waste forms. This paper describes how a joint effort between the Belgian Agency for Radioactive Waste and Enriched Fissile Materials, the Belgian Waste Treatment and Storage Facility Belgoprocess, and the Belgian Nuclear Research Centre worked out solutions for the 1st beryllium moderator of the BR2 and the highly activated internal parts of the BR3. The current position regarding activated graphite and aluminium is also summarised.
The Safety Case for a Deep Geological Repository: More Than Just Presenting Numbers
Authors: Schneider Jürg W. , Johnson Lawrence H. , Zuidema Piet , Smith Paul A. , Gribi Peter 
Published: April 2020

Abstract
A convincing safety case is a prerequisite for positive decisions in the stepwise process of planning and implementing a deep geological repository. In addition to evaluations of dose or risk, a safety case must provide arguments that the system and its safety functions are well understood, that the methods, models and data used in the evaluations are applicable, and have been correctly applied, and that the full spectrum of possibilities for the characteristics of the repository system and its evolution has been taken into account in defining the cases to be evaluated. This paper takes examples from the Swiss waste management programme and from natural analogue studies to illustrate how the safety functions for a repository are taken into account in a safety assessment and how independent evidence can be used to support the statement that these safety functions will operate as expected.
Lessons Learned at the Waste Isolation Pilot Plant: Share, Listen, and Learn to Earn Stakeholder Acceptance
Authors: Eriksson Leif G. 
Published: April 2020

Abstract
On March 26, 1999, the United States (U.S.) Department of Energy (DOE) opened the nation’s first deep geological disposal system (repository) for long-lived radioactive wastes/materials (LLRMs) at the Waste Isolation Pilot Plant (WIPP) site, New Mexico, United States of America (USA). The opening of WIPP embodies gradually achieved acceptance, both local and global, on scientific, institutional, regulatory, political, and public levels. In the opinion of the author, five significant determinants for the successful siting, certification, and acceptance of WIPP, were the existence of:
• A willing and supportive host community;
• A strong, independent regulator;
• A regulatory framework widely perceived to (over)protect public health and the environment;
• A structurally simple, old, stable, host-rock with excellent radionuclide containment and isolation characteristics; and
• An open siting, site characterization, repository development, certification and recertification process with regularly scheduled opportunities for information exchanges with affected and interested parties, including a) prompt responses to non-DOE concerns and b) transparency/traceability of external-input into, and the logic behind, the DOE’s decision-making process.
The nation’s and the world’s next deep geological repository for LLRMs is currently scheduled to open in 2010. As follows, in addition to providing a national solution to safe disposal of LLRMs, the opening and continued safe operation of WIPP provides an international role model that effectively dispels the global myth that LLRMs cannot be safely disposed in a deep geological repository.
Mitigation of Uranium in Effluents by Zero Valent Iron: The Role of Iron Corrosion Products
Authors: Noubactep Chicgoua , Volke Peter , Merkel Broder , Meinrath Günther 
Published: April 2020

Abstract
This study was conducted to investigate the influence of iron corrosion products on the removal precipitation of uranium (VI) from aqueous solutions by zero valent iron (ZVI). Batch experiments were conducted with solutions containing 20 mg /L U(VI) (0.084 mM) at room temperature with washed and untreated scrap iron.
The results show significant decrease of the immobilization efficiency of ZVI for uranium in the presence of both atmospheric corrosion products and those generated in situ. Among the relevant groundwater components, those who promote the iron corrosion e.g. SO42−, Cl− produce more corrosion products and have a stronger negative effect on the immobilization process. Carbonate ions also have a negative effect on the immobilization efficiency since they form stable complexes with uranium and reduce the adsorption on the iron surface which is the first step to immobilization. Humic substances didn’t show any negative effect but rather a slight enhancement of the immobilization rate. These results show that washing ZVI materials with hydrochloric acid before testing will overestimate their efficiency in practical use.
Stabilisation of Soft Tailings: Practice and Experience
Authors: Jakubick Alex T. , McKenna Gord 
Published: April 2020

Abstract
Environmentally acceptable reclamation solutions for the decommissioned tailings ponds can be achieved by integrating the tailings deposits into the landscape either by turning the tailings pond/deposit into a lake and maintaining a permanent water cover over the deposit or by stabilisation of the deposits as a “dry” landform.
Conceptually, the dry landscape reclamation involves three remedial steps: 1) Placement of an interim cover on the tailings surface to provide the consolidation load and create a stable working platform. 2) Building of a surface contour providing suitable run off conditions for the surface water. 3) Capping the surface with a final cover to control infiltration into the tailings.
For reliable predictions of consolidation, proper sequencing of remedial steps/measures and cost efficient reclamation of soft tailings it is necessary to use Non Linear Finite Strain (NLFS) codes. The advantage of the NLFS program system Consol2D is that it allows the reconstruction of the history of the tailings discharge, the evaluation of the material parameters, the calculation of settlement in inhomogeneous deposits, the 3D quantification of the settlement trough and the evaluation of the drainage effect of the vertical drains.
NDA in the Frame of Waste Characterization and the Practice at Belgoprocess
Authors: Botte John , Hendrickx Jean-Paul , Luycx Paul 
Published: April 2020

Abstract
Governments and the nuclear sector realize more and more the importance of waste characterization in the end cycle.
Several methods are available for radiological characterization. Each method has his advantages and disadvantages in regard to the definition of the operational range based on the waste characteristics and the knowledge of the waste.
The decision on what technique(s) to be used will be based on the required precision, accuracy, detection limit and isotopes to be declared for a certain purpose and on the related costs.
Recently Belgoprocess, the Belgian waste treatment facility, has performed large efforts and investments to assure a good waste characterization.
Fabrication of MOX Storage Rods for Direct Final Disposal As Method for Immobilisation of Surplus Plutonium
Authors: Küppers Christian , Sailer Michael 
Published: April 2020

Abstract
The German strategy of handling spent fuel was focussed on reprocessing since the 1960s although the German breeder projects are phased out since the late 1980s. Therefore, considerable plutonium was separated from German spent fuel and not reused. Some materials are not storable without periodical treatment and must be processed. The question arises whether a product can be fabricated that is also suitable for final disposal. Fabrication of “MOX storage rods” is considered to be the best treatment, especially under German conditions, for this purpose. This treatment can also reduce costs significantly, probably up to one third. The storage rod option is based on techniques of which Europe has large-scale technical experience and only existing facilities are needed. Relevant amounts of plutonium could be processed in a relatively short time, this means several tons per year, starting before 2005.
Development of Cutting Techniques Applied on Sodium Wastes Application to Liquid Metal Fast Reactors (LMFRS) Decommissioning
Authors: Rodriguez G. , Goedert N. , Cournille M. 
Published: April 2020

Abstract
In the frame of the general clearing of CEA centres of sodium waste, a new Based Nuclear Installation (BNI) is being studied to collect all the active sodium wastes and to treat them in the same building. This BNI is called ATENA. (Atelier de Traitement et Entreposage des déchets Na)
In designing ATENA, CEA identified that there was a lack in the knowledge of techniques for the cutting of metallic sodium wastes. Therefore it was decided to test several cutting technologies on different sodium wastes to evaluate their performances in a specific situation.
These experiments started with a general review of all the cutting techniques used for decommissioning purposes to see if they could be compatible with sodium safety standards.
Several cutting techniques were dismissed and finally eight cutting techniques were selected to be tested with sodium. Four of them were considered to be the best. This paper describes these sodium-cutting trials and explains the reasons of the final selection.