bluuebunny/crossref_metadata_2025_split huggingface dataset:
Mizunami Underground Research Laboratory in Japan: Pre-Excavation Site Characterization and Influence on Design Concepts
Authors: McCrank Glen F. , Amano Kenji , Koide Kaoru , Matsui Hiroya , Mikake Shinichiro , Nakano Katsushi , Ota Kunio , Saegusa Hiromitsu , Sugihara Kozo , Takeda Seietsu , Takeuchi Shinji 
Published: April 2020

Abstract
The Mizunami Underground Research Laboratory (MIU) will be constructed by the Japan Nuclear Cycle Development Institute (JNC) in central Japan. The project is planned in three overlapping phases consisting of I) Surface-based Investigation II) Construction and III) Operations Phases. This paper addresses the methods used to investigate the geological environment, the integration of the site knowledge into conceptual models and the application of the knowledge in designing the facility; some aspects of the future experimental programme are discussed.
Radwaste Processing and Volume Reduction in America: A Commercial Providers Prospective
Authors: Greenman William , Rae Glen 
Published: April 2020

Abstract
In the early 1980’s, the United States realized the need for developing and implementing volume reduction programs as a result of rapidly increasing disposal prices. To better serve the nation, in 1984 Duratek established a fixed base volume reduction facility near Oak Ridge, TN to provide radwaste volume reduction and enhanced waste forms. Over the years, this facility has been expanded to provide many different treatment schemes with technologies that include two of the nations only radwaste incinerators, the largest ultra compactor and the largest metal melter. The original Oak Ridge site has also geographically expanded to three other volume reduction sites (two in Tennessee and one in South Carolina). These technologies provide extreme VR’s for DAW and other solid waste as well as a beneficial and controlled recycling outlet for metals resulting in essentially “0” waste for disposal. In addition, technologies such as waste segmentation and sorting, material decontamination, liquid waste processing (with benefits of 80–90% reduction in radwaste generation compared to media based systems), survey for release, and other beneficial metal recycling completed the primary technologies and reduced waste volumes further.
This paper will provide information on the techniques developed and utilized over the years to reach ultimate volume reduction and assist in reducing the nations radwaste disposal volume by well over 50%.
Application of Polyhexamethyleneguanidine Type Biocides at Cementing the Radioactive Waste
Authors: Varlakov Andrey P. , Gorbunova Olga A. , Barinov Aleksandr S. , Iljin Vadim A. , Efimov Konstantin M. , Gembitsky Petr A. 
Published: April 2020

Abstract
In order to prevent biological corrosion of cement compound containing radioactive waste, it is proposed to use biocidal additives of polyhexamethyleneguanidines (PHMG), which have a wide range of biocidal activity.
It has been shown that inclusion of biocidal additives of polyhexamethyleneguanidines in quantities 0,25–2% wt. into the grout used for the solidification of radioactive waste (RW) or for recovering the integrity of old RW repositories provides the necessary bacteriostatic and biocidal protection of cement compound and improves all the regulated properties — compression strength, Cs-137 leaching rate, frost-resistance, radiation resistance and long term water resistance.
Comparative Assessment of Low-Level Radioactive Waste Life-Cycle Disposal Costs of U.S. Commercial Facilities
Authors: Bentz E. J. , Bentz C. B. , O’Hora T. D. 
Published: April 2020

Abstract
This paper provides a comparative assessment of low-level radioactive waste (LLW) life-cycle costs for U.S. commercial disposal facilities. This assessment includes both currently operational facilities and planned commercial facilities.
After identifying the individual facility’s operational period, current or planned capacity, and historical disposal volumes (where applicable), the paper describes the respective facilities’ waste acceptance criteria, anticipated waste characteristics, and disposal technologies employed.
A brief identification of key components of cost categories that constitute life-cycle cost for the disposal facilities is provided, as well as an identification of factors that affect life-cycle cost. A more specific comparison of certain life-cycle cost components for the disposal facilities is provided, with regard to U.S. LLW disposal volumes and characteristics. Similarities and differences in total life-cycle cost and life-cycle category-specific costs among the U.S. facilities are presented and discussed. The data presented reveals that:
• No new LLW commercial disposal facilities have been sited in the U.S. since 1988, and that siting of LLW disposal facilities in the U.S. has become increasingly difficult and contentious, necessitating long lead times and significant up-front costs — without any certainty of success.
• Overall, life-cycle costs for LLW disposal at U.S. commercial facilities have increased significantly over time, reflecting increased regulatory compliance requirements, state-imposed access fees and taxes, local community hosting incentive costs, and cost escalation inherent in delays in establishing facilities or modifying existing licensed facilities.
• Life-cycle costs are also significantly affected by the nature of the engineered isolation technology employed, reflecting the geologic characteristics of the siting location and the activity levels of the wastes accepted.
• Since many of the newly-planned facilities anticipate receiving lower total volumes with an increasingly greater percentage of higher activity wastes (than historical volumes disposed) and are to be sited in more ecologically sensitive geologic regions, they will require more comprehensive — and hence more expensive — engineered isolation technologies.
As a result, currently planned facilities are anticipated to experience significantly higher total life-cycle costs than existing operational facilities.
Experimental Study on Radioactive Contamination of Spent Reactor Graphite From the SGCE (Tomsk-7) Plutonium-Production Reactors
Authors: Bushuev Anatoly , Verzilov Yury , Zubarev Victor , Aleeva Tatiana , Petrova Ekaterina , Proshin Igor , Kachanovsky Alexander , Silin Boris , Dmitriev Alexander , Zakharova Elena , Ushakov Sergey , Nikolaev Andrey , Baranov Igor , Kabanov Yury , Kolobova Ella 
Published: April 2020

Abstract
The paper is devoted to investigations of spent reactor graphite contamination. The study was based on the sampling from the graphite stacks and sleeves of three shut down plutonium-production reactors in Siberian Group of Chemical Enterprises (SGCE), Seversk (Tomsk-7). Several hundreds graphite samples were analysed by means of α-, β-, γ- and X-ray spectrometry. Contents of 3H, 14C, fission products, actinides, products of impurities neutron activation were determined. The inventories of above mentioned radionuclides were estimated. Correlation between content of various groups of radionuclides was studied. Graphite sleeves contamination was studied depending on the period of their irradiation.
Comparison of obtained experimental data for the three SGCE reactors with calculated predictions showed substantial difference. The conclusion was made on applicability of the data obtained for estimations of spent graphite contamination in other reactor types including RBMK reactors.
The main part of the studies was performed in the frame of the ISTC Projects No. 561 and No. 1409 fulfilment.
New Treatment Facility for Optimized Conditioning of Activated Core Components
Authors: Gestermann G. , Finkbeiner R. 
Published: April 2020

Abstract
In the course of decommissioning of one German BWR power station, a number of activated core components have to be prepared for interim storage in a shielded building at the NPP site. For optimum use of the storage capacity, a maximum volume reduction is to be achieved.
The components to be treated are 774 water channels and 130 control elements. The resulting material has to be accommodated in special design stainless steel drums which initially will be positioned in the fuel pool awaiting subsequent treatment. For accommodation of the material in the drums, the core components have to be cut by an underwater shear, and in order to achieve a maximum loading of the drum, after cutting the material is compacted by an underwater in-drum compactor. An integrated facility consisting basically of the underwater shear and the in-drum compactor has especially been designed and built for that given task. The paper will describe that equipment and the work to be performed.
Treatment and Conditioning of Low- and Intermediate Level Radioactive Waste at the Austrian Research Centers Seibersdorf (ARCS)
Authors: Neubauer Josef 
Published: April 2020

Abstract
The Austrian Research Centers Seibersdorf (ARCS) serve as the country’s centralised facilitiy for the treatment, conditioning and interim storage of low- and intermediate level radioactive waste (Radwaste). All Radwaste originating in Austria is transferred to ARCS and comes from one or several sources: The former operation of a research reactor, the application of radioactive materials in medicine, R&D, industry, IAEA Laboratories at Seibersdorf, and technically enhanced natural radioactive materials, such as by-product at industrial metallurgical processes and other processes, is transferred to ARCS for treatment, conditioning and interim storage.
Dependent on the waste category, the relevant treatment technology is applied.
In total more then 6900 m3 of low and intermediate level radioactive solid wastes originating from Austria were treated in the period 1976 to 2000.
The aim of treatment and conditioning is the safe enclosure of the Radwaste by the use of barriers to surround and isolate the waste and transform it to an insoluble form. At the same time it is important to reduce the volume of the raw waste by applying appropriate treatment technologies, in order to save space in the storage facilities and thereby achieving a cost savings.
Systematical Approach to the Radioactive Waste Management in Slovak Republic
Authors: Breza Milan 
Published: April 2020

Abstract
The aim of this paper is to inform of the world’s specialists about the systematical approach to the radioactive waste management in Slovak Republic (Slovakia).
In Slovakia are in operation six preassurized light water reactors at NPPs V-1, V-2 in Jaslovske Bohunice and reactors 2×440 MWe in NPP Mochovce at present. The installed elektro-capacity produced from these reactors is 2640 MWe total.
The first experimental NPP in the former Czechoslovakia was NPP A-1. It had used natural uranium as a fuel and heavy water as a moderator and CO2 as a cooling medium. The installed preformance of this NPP A-1 was 150 MWe and was in operation since 1972 to 1977. After two technological incidents former Czechoslovak government have decided to take out of this NPP A-1 from operation.
According to former Soviet projects concerning for costructions of the above mentioned NPP in Jaslovske Bohunice Slovakia, to these project didn’t belong the projects for constructions of the technological units for the radioactive waste treatment, expect of the evaporating plant for the contaminated water concetration and storage tanks for the evaporated concentrates storage and storage possibilities for the contaminated solid waste storage. A comparable complicated situation was at NPP A-1, because all type of the radioactive waste streams content gama, beta and alfa radioanuclides as well. The second technological incident at NPP A-1 (with the some portion of the cooling medium CO2 leaking form the reactor vessel out) caused contamination for the NPP A-1 operating spaces by alfa radionuclides and alfa radionuclides penetrated in all waste streams at storage tanks, pipelines, vessels, etc. as well.
During the beginning of the 80’s were realized some states R and D tasks that solved the systematical approach to the radioactive waste treatment in former Czechoslovakia. The main aim of the first etap was solved the all problems connected for the radioactive waste, treatment that are formed during NPPs of the VVER-440 type operation. It was necessary to secure the production of the electro energie. The most seriously priority had the tasks, that solved of the evaporator concentrates and solid waste treatment, especially burnable, pretratment. Simultaneuosly were solved the problems connected with the proposals, designings and realisations of the disposals for the long term storage of the treated RAW in the Czech and Slovak regions. In this presentation I have tried to deliver main information, that give of the survey for RAW treatment and their long term safety storage at disposal system in Mochovec - Slovakia.
Current Status of the New Spent Fuel Dry Storage Facility in Romania
Authors: Andrei Veronica , Glodeanu Florin , Rotaru Ioan , Daian Ioana 
Published: April 2020

Abstract
The Cernavoda Nuclear Power Plant (NPP), in commercial operation since 1996, produces more than 10% of the electricity produced in Romania. Recently, the Romanian Government declared its commitment for completion of a second reactor of the CANDU design, under construction on the Cernavoda site.
The annual spent fuel arising from a CANDU reactor is about 100tU. The current policy for spent fuel management as practiced by the plant owner is to store it in the reactor bay for minimum six years and in a dry storage facility for a minimum of 50 years. For geological disposal of spent fuel, the “wait and see” strategy is considered the best approach, as Romania has a relative low scale nuclear program and wants to benefit by the international progress in this field.
The construction of a new spent fuel dry storage facility located in the vicinity of the nuclear power reactor site represents a main priority for the next three years. The site of this facility will accommodate two nuclear units’ inventories of spent fuel for the entire planned lifetime. An international public-limited tender was organized to select the supplier of the dry storage technology in early 2001. The tenderer was asked to propose a proven and licensed technology capable of storing CANDU spent fuel according to specified design parameters and safety and environmental requirements.
Design, construction, operation or licensing legal specific requirements for such a facility is generally not established and other already existing national requirements are applicable to a limited degree. Taking into account the different approaches and iterative processes required for Romanian authorities to regulate the nuclear activities for different fields, this paper considers the realistic path forward.
The current status and main aspects of the development and licensing of the new nuclear facility in Romania is presented in this paper.
Safe Management of Spent Fuel and Radioactive Waste in Bulgaria: Precondition for Future Development of Bulgarian Nuclear Program
Authors: Necheva Christina T. , Jordanova Olesja V. 
Published: April 2020

Abstract
Bulgaria has to solve complicated problems connected with the safe management of spent fuel and radioactive waste with regard to the future of its nuclear program. The efforts at national level are concentrated on the safe management of the low and intermediate level waste and on the safe temporary storage of spent fuel. For Bulgaria as a country - producer of nuclear energy on the basis of imported fresh fuel, an option of major interest is to deal with spent fuel final solution in co-operative way with the supplier of the fresh fuel.
Development of a New Method to Reduce the Measurement Uncertainties in Gamma-Ray Spectroscopy of Bituminized Radioactive Waste Drums
Authors: Perot Bertrand , Artaud Jean-Luc , Chabalier Bernard , Bonifay Pierre , Bernard Sébastien , Misraki Josette , Fulleringer Daniel , Chany Philippe , Dogny Stéphane 
Published: April 2020

Abstract
The determination by gamma spectroscopy of the activity of radionuclides emitting low energy radiation (less than 200 keV) in bituminized waste drums can be affected by significant uncertainty if the chemical composition of the matrix is not well-known. Indeed, some elements with high atomic numbers (Z) can significantly modify the absorption of these radiations. The CEA has therefore studied, using numerical simulation, the feasibility of a method which would take into account this effect, based on the analysis of the gamma spectra continuum due to Compton scattering (1). This article describes an experimental validation of the simulations and of the method.
Operational Results of Radionuclide Assay System at the Kori Nuclear Power Plant: Status and Results
Authors: Park Jin Beak , Lee Kun Jai , Kang Duck Won , Kim Dae Hwan 
Published: April 2020

Abstract
In 1996, the fully automated radionuclide assay system, which is used for the non-destructive analysis and for the evaluation of radioactive waste drums, was installed at Kori nuclear power plant site and started to operate. The system is composed of gamma scanning hardware facility, equipment for separation and detection of non-gamma radionuclide, scaling factor prediction program, and control facility for radionuclide assay system. In this study, the operational status and results of radionuclide assay system are investigated from the experience of several years of operation. The results of the radionuclide assay system are compared with the results of dose to curie conversion program (DOSE).
High-Level-Waste and Spent Fuel Storage in Switzerland
Authors: Sach Udo , Schreck Goswin , Ritter Max , Wenger Jean-Pierre 
Published: April 2020

Abstract
At present, Switzerland has no final repository for radioactive wastes. Very early, the Swiss nuclear power plant operators were aware of the necessity to expand interim storage capacity for spent fuel elements and operational wastes. Already in 1991, Nordostschweizerische Kraftwerke AG (NOK) therefore started building a reactor-site interim storage facility (ZWIBEZ) at its Beznau power plant site.
Moreover, as early as in 1990, “ZWILAG Zwischenlager Würenlingen AG”, a company established by the nuclear power plant operators had initiated the licensing procedure for a central interim storage facility in Switzerland. This central interim storage facility is designed for the storage of all categories of radioactive wastes and includes a conditioning facility for low-level and medium-level waste. Eleven years later, in July 2001, the first transport and storage cask loaded with irradiated fuel elements was stored in this facility.
For both of the stores the concept of dry interim storage in suitable storage casks in a storage hall was chosen for the storage of irradiated fuel elements and vitrified high-level wastes from reprocessing. Cooling is established through natural circulation. Leaktightness of the casks is continuously monitored by means of a cask monitoring system.
The other wastes arising from nuclear power plant operation and reprocessing are stored in a ventilated storage hall which provides shielding and — depending on the radioactive inventory — protection against external impact. The conditioned radioactive wastes, packaged in drums, are placed into open storage containers with identical base and having the same sling points as ISO containers. These containers are stacked up in free-standing stacks up to a height of 16 m. The storage concept varies, depending on the radioactive inventory; for the ZWIBEZ reactor-site interim store, a storage hall for low-level waste has been built without partition walls, whereas the store for the medium and high-level waste in the central interim store ZWILAG has been designed with partition walls dividing the hall into several storage shafts which are closed by shielding slabs.
By including a hot cell into the ZWILAG facility, the purpose of this facility has been expanded beyond interim storage of radioactive waste to cover also the visual inspection of fuel elements and vitrified waste canisters as well as the reloading of fuel elements and canisters from smaller transport casks into combined transport and storage casks. Furthermore, the hot cell enables inspection and/or repair work to be performed in the cask lid area of loaded transport and storage casks, the replacement of the lid seals of storage casks and the conditioning of medium-level waste.
Fundamental Principles of and Reactions to the June 2001 Enviromental Radiation Protection Standards for the Yucca Mountain Site
Authors: Monroe Scott D. , Schultheisz Dan 
Published: April 2020

Abstract
On June 13, 2001, the United States Environmental Protection Agency (USEPA) issued final public health and environmental radiation protection standards for the proposed spent nuclear fuel and high-level radioactive waste repository at Yucca Mountain, Nevada. The Energy Policy Act of 1992 (EnPA) had directed USEPA to develop these standards for the Yucca Mountain site.
Oxidation of Organic Radioactive Waste by Electrochemical Mediation
Authors: Cantrel E. , Fonteyne A. , Impens N. , Rahier A. 
Published: April 2020

Abstract
Liquid as well as solid organic radioactive waste can be processed by means of combustion. However, this method presents several well-known drawbacks including the corrosion of the ovens, the production of radioactive ashes and radioactive or toxic volatile products. The electrochemical mediation process is an excellent alternative to combustion, especially when dealing with hazardous materials such as explosives, pesticides, drugs and nuclear organic waste. Nevertheless, using the silver(II) species as electrogenerated mediator, requires to work in concentrated nitric acid media, which results, via its electrolysis, in a continuous NOx emission. The classical approach, using absorption columns and scrubbers to trap NOx, impairs the attractiveness of this process. Therefore the SCK•CEN has developed and patented an original method to suppress in-situ any formation of NOx, conferring the process mobility and compactness.
Standards and Specifications for Packaging Waste in the UK
Authors: Barlow S. V. , Palmer J. D. 
Published: April 2020

Abstract
The role of Nirex is to provide the United Kingdom with safe and environmentally sound options for the long-term management of radioactive waste generated by the UK’s commercial, medical, research and defence activities. This includes all intermediate level waste and some low level waste (ILW and LLW). One of the key objectives of Nirex over the past 10 years or so, has been to ensure that when waste is packaged, it is in a form suitable for its future safe management including storage, transport, handling and potential disposal.
Being responsible for development of future long-term waste management facilities, Nirex is uniquely placed to define standards and performance specifications for waste packages that include wasteform and container design, quality assurance and data recording requirements. In addition to the specifications, Nirex also provides detailed advice on the suitability of specific packaging proposals and plant designs against the foreseen requirements for future transport, handling, storage and potential disposal, as defined by the Nirex phased disposal concept. Where packaging proposals meet these requirements, Nirex is prepared to endorse the proposed approach through the issue of a ‘Letter of Comfort’.
This approach has enabled the commencement of waste packaging operations with a high degree of confidence that the waste product will meet future waste management requirements, including potential disposal requirements. This paper provides a summary of the standards and specifications developed by Nirex for waste packages, and of the assessment process applied by Nirex in providing advice and endorsement of specific packaging proposals.