bluuebunny/crossref_metadata_2025_split huggingface dataset:
Using Probabilistic Criteria in an Assessment of the Potential Radiological Consequences of the Decommissioning of a Nuclear Research Reactor
Authors: Wallner Christian , Rall Anna-Maria , Thummerer Severin 
Published: February 2014

In order to assess the risk of radiological consequences of incidents and accidents in nuclear facilities it is important to contemplate their frequency of occurrence. It has to be shown that incidents and accidents occur sufficiently seldom according to their radiological consequences i. e. the occurrence frequency of radiological doses has to be limited. This is even demanded by the German radiation protection ordinance (StrlSchV), which says that in nuclear facilities other than nuclear power plants (NPP) in operation and for decommissioning, the occurrence frequency of incidents and accidents shall be contemplated in order to prove the design of safety measures and safety installations.
Based on the ideas of the ICRP64, we developed a risk based assessment concept for nuclear facilities, which fulfils the requirements of the German regulations concerning dose limits in normal operation and design basis accidents. The general use of the concept is dedicated to nuclear facilities other than nuclear power plants (NPP) in operation and for decommissioning, where the regulation of risk assessment is less sophisticated. The concept specifies occurrence frequency limits for radiation exposure dose ranges, i. e. the occurrence frequency of incidents and accidents has to be limited according to their radiological effects.
To apply this concept, scenarios of incidents and accidents are grouped in exposition classes according to their resulting potential effective dose to members of the general public. The occurrence frequencies of the incidents and accidents are summarized in each exposition class whereas the sum must not exceed the frequency limits mentioned above.
In the following we introduce the application of this concept in the assessment of the potential radiological consequences of the decommissioning of a nuclear research reactor. We carried out this assessment for the licensing process of the decommissioning on behalf of German authorities.
Public Involvement on Closure of Asse II Radioactive Waste Repository in Germany
Authors: Kallenbach-Herbert Beate 
Published: February 2014

From 1967 to 1978, about 125,800 barrels of low- and intermediate level waste were disposed of — nominally for research purposes — in the former “Asse” salt mine which had before been used for the production of potash for many years. Since 1988 an inflow of brine is being observed which will cause dangers of flooding and of a collapse due to salt weakening and dissolution if it should increase.
Since several years the closure of the Asse repository is planned with the objective to prevent the flooding and collapse of the mine and the release of radioactive substances to the biosphere.
The first concept that was presented by the former operator, however, seemed completely unacceptable to regional representatives from politics and NGOs. Their activities against these plans made the project a top issue on the political agenda from the federal to the local level.
The paper traces the main reasons which lead to the severe safety problems in the past as well as relevant changes in the governance system today. A focus is put on the process for public involvement in which the Citizens’ Advisory Group “A2B” forms the core measure. Its structure and framework, experience and results, expectations from inside and outside perspectives are presented.
Furthermore the question is tackled how far this process can serve as an example for a participatory approach in a siting process for a geological repository for high active waste which can be expected to be highly contested in the affected regions.
Applicability of Iron Phosphate Glass Medium for Loading NaCl Originated From Seawater Used for Cooling the Stricken Power Reactors
Authors: Kobayashi Hidekazu , Amamoto Ippei , Yokozawa Takuma , Yamashita Teruo , Nagai Takayuki , Kitamura Naoto , Takebe Hiromichi , Mitamura Naoki , Tsuzuki Tatsuya 
Published: February 2014

As the part of investigation for immobilization of the sludge as one of the radioactive wastes arising from the treatment of contaminated water at Fukushima Dai-ichi nuclear power plant, applicability of vitrification method has been evaluated as a candidate technique.
The aim of this study is to evaluate the influence of NaCl as one of the main constituents of sludge, on glass formation and glass properties. Two kinds of iron phosphate glass (IPG) media in the xFe2O3-(100-x)P2O5, with x = 30 and 35 (mol%) were chosen and the glass formation, structure and properties including density, coefficient of thermal expansion, glass transition temperature, onset crystallization temperature and chemical durability of NaCl-loaded IPG were studied. The results are summarized as follows.
Sodium chloride, NaCl could be loaded into IPG medium as Na2O and Cl contents and their loading ratio could be up to 19 and 15 mol%, respectively. Majority of Cl content of raw material NaCl was thought to be volatilized during glass melting. Loading NaCl into IPG induces to depolymerize glass network of phosphate chains, leads to decrease both glass transition and onset crystallization temperatures, and to increase coefficient of thermal expansion. NaCl-loaded IPG indicated good chemical durability in case of using 35Fe2O3-65P2O5 medium.
Calculation of the Nuclear Material Inventory in a Sealed Vault by 3D Radiation Mapping
Authors: Adsley Ian , Tur Yevgeniy , Klepikov Alexander , Wells David 
Published: February 2014

The paper relates to the determination of the amount of nuclear material contained in a closed, concrete lined vault at the Aktau fast breeder reactor in Kazakhstan. This material had been disposed into the vault after examination in an experimental hot cell directly above the vault. In order to comply with IAEA Safeguards requirements it was necessary to determine the total quantities of nuclear materials — enriched uranium and plutonium — that were held with Kazakhstan. Although it was possible to determine the inventory of all of the accessible nuclear material — the quantity remaining in the vault was unknown.
As part of the Global Threat Reduction Programme the UK Government funded a project to determine the inventory of these nuclear materials in this vault. This involved drilling three penetrations through the concrete lined roof of the vault; this enabled the placement of lights and a camera into the vault through two penetrations; while the third penetration enabled a lightweight manipulator arm to be introduced into the vault. This was used to provide a detailed 3D mapping of the dose rate within the vault and it also enabled the collection of samples for radionuclide analysis.
The deconvolution of the 3D dose rate profile within the vault enabled the determination of the gamma emitting source distribution on the floor and walls of the vault. The samples were analysed to determine the fingerprint of those radionuclides producing the gamma dose — namely 137Cs and 60Co — to the nuclear materials. The combination of the dose rate source terms on the surfaces of the vault and the fingerprint then enabled the quantities of nuclear materials to be determined.
The project was a major success and enabled the Kazakhstan Government to comply with IAEA Safeguards requirements. It also enabled the UK DECC Ministry to develop a technology of national (and international) use. Finally the technology was well received by IAEA Safeguards as an acceptable methodology for future studies.
From Passive Storage to Daily Waste Retrieval; Changing the Working Culture at Sellafield’s Pile Fuel Storage Pond
Authors: Carlisle Derek 
Published: February 2014

The Pile Fuel Storage Pond (PFSP) was built in 1948/50 to treat materials from the Windscale Piles. Multiple operational regimes over the intervening 60 years have resulted in a complex inventory of spent nuclear fuels, solid and liquid intermediate level wastes. A coordinated programme of work, designed to retrieve and safely dispose of the pond contents, has been implemented to enable the decommissioning of the facility.
The long period of passive storage operations which preceded the implementation of the programme meant that the operator was faced with a dual challenge of providing new technical capability and changing a working culture that was inappropriate for the dynamic environment required to successfully deliver the programme.
It was recognised that the nature of the programme meant that implementing a standard manufacturing approach to operations would not be appropriate. In order to create a dynamic retrievals focussed working culture, the operator has vigorously embraced change programmes aimed at improving a number of working practices including encouraging innovation, managing integrated but flexible production schedules, and encouraging workface problem solving.
The combined impact of beginning to resolve the technical challenges and focussing on the delivery culture has resulted in the facility making a step change towards becoming fully retrievals operations focussed.
Long-Term Safety Assessment of Trench-Type Surface Repository at Chernobyl, Ukraine — Computer Model and Comparison With Results From Simplified Models
Authors: Haverkamp B. , Krone J. , Shybetskyi I. 
Published: February 2014

The Radioactive Waste Disposal Facility (RWDF) Buryakovka was constructed in 1986 as part of the intervention measures after the accident at Chernobyl NPP (ChNPP). Today, the surface repository for solid low and intermediate level waste (LILW) is still being operated but its maximum capacity is nearly reached. Long-existing plans for increasing the capacity of the facility shall be implemented in the framework of the European Commission INSC Programme (Instrument for Nuclear Safety Co-operation).
Within the first phase of this project, DBE TECHNOLOGY GmbH prepared a safety analysis report of the facility in its current state (SAR) and a preliminary safety analysis report (PSAR) for a future extended facility based on the planned enlargement. In addition to a detailed mathematical model, also simplified models have been developed to verify results of the former one and enhance confidence in the results. Comparison of the results show that — depending on the boundary conditions — simplifications like modeling the multi trench repository as one generic trench might have very limited influence on the overall results compared to the general uncertainties associated with respective long-term calculations. In addition to their value in regard to verification of more complex models which is important to increase confidence in the overall results, such simplified models can also offer the possibility to carry out time consuming calculations like probabilistic calculations or detailed sensitivity analysis in an economic manner.
Improvements Realized on the Cryogenic Solution Called NiThrow™
Authors: Objois Laurent , Moggia Fabrice , Toulemonde Valérie , Varet Thierry , Richard Frédéric , Benchikhoune Fernand 
Published: February 2014

The economic, environmental, social and societal constraints have pushed the nuclear industry to develop new dismantling and decommissioning (D&D) techniques in order to meet the current industrial requirements. Since 5 years, AREVA, in partnership with Air Liquide, has been working on a new cryogenic solution that can be used to achieve the main important D&D operations (i.e. cutting, surface decontamination and concrete scabbling). This solution, called NiThrow™, is based on the Nitrocision LLC concept and consists in spraying highly pressurized liquid nitrogen (3500 bar) at a very low temperature (−140°C) onto a surface. The main advantage in using liquid nitrogen instead of water is due to its rapid conversion into gaseous nitrogen that will avoid the generation of liquid effluent. The preliminary trials made in France at Saint Ouen l’Aumône and at the AREVA SICN facility by using NiThrow™ appeared to be very interesting and promising in terms of effectiveness, reduction of the dosimetry (i.e. ALARA principle) and adaptability to the different types of surfaces and materials to be treated. On an other hand, they also highlighted a couple of defects like a lack of reliability that is not compatible with a safety use in a nuclear environment.
In 2009, with the aim of fixing all these issues, AREVA decided to start an extensive R&D program. This work has been essentially focused on the lancing tool, the insulation and the flexibility of the liquid nitrogen distribution network, but also on the design and the manufacturing of a waste collection unit and a dedicated carrier. Today, these R&D efforts have been totally completed. and resulted to the publication of 3 patents. Also, the collaboration, with Air Liquide and the American company Conco allows AREVA to propose to its customers a safe and reliable solution for their cleaning operations in France and all over the world.
Radionuclide Inventory of Vitrified Waste After Spent Nuclear Fuel Reprocessing at La Hague
Authors: Meleshyn A. , Noseck U. 
Published: February 2014

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 14C, 36Cl, 79Se, and 129I), 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.
Brennilis Decommissioning: The Case of the Treatment Station and the Discharge Channel Remediation
Authors: Ollivier Dehaye Catherine , Pellenz Gilles 
Published: February 2014

The site of the Monts d’Arrée located in Brennilis in the area of Brittany in France is a former 70 MWe heavy water reactor, that was jointly operated by CEA and EDF from 1967 until 1985. It was an industrial prototype. EDF is now in charge of its decommissioning.
This paper presents the management plan and the remediation objectives that were approved by the French Nuclear Authority for the channel remediation and the first studies carried out for the future Treatment Station remediation.
The Involvement of SCK•CEN in Decommissioning Projects
Authors: Noynaert Luc , Dadoumont Jérôme , Cornelissen René , Van den Dungen Kurt 
Published: February 2014

SCK•CEN is the Belgian Nuclear Research Centre. Founded in the mid-fifties, it has accumulated experience and know-how in all fields of the nuclear power production: in the neutronics calculation, radiation protection, waste management, fuel performance and analysis, nuclear measurements, radiochemistry, reactor technology, etc. Since 1989, SCK•CEN has launched Decommissioning activities to deal with the Technical Liabilities created by 40 years of operation. The main projects started were:
• the dismantling of the BR3 PWR reactor;
• the dismantling of active laboratories and the decontamination of buildings for unrestricted reuse;
• the management of the waste arising from the refurbishment activities of the BR2, especially the management of the high active beryllium matrix.
In 1989, the BR3 reactor, a Pressurized Water Reactor, was selected by the European Commission as one of the four pilot dismantling projects in the framework of the EC five year RTD program on dismantling nuclear installations. Through this project, SCK•CEN has built up a broad know-how in dismantling and decommissioning operations. This know how concerns the decontamination for dose rate reduction and/or free release of materials, teleoperated techniques for cutting highly activated components of a reactor, concrete decontamination techniques, characterization techniques of radioactive waste or for free release of components and development of decommissioning management and recordkeeping of material streams and of nuclear material accountancy.
SCK•CEN is now actively involved in other decommissioning projects in Belgium and in expertise abroad.
After giving an overview of the main achievements and the perspectives of the decommissioning of the BR3 reactor, the paper intends to present the involvement of SCK•CEN in the other decommissioning projects and to give an overview of our activities and capacities.
Developing a Concept for a National Used Fuel Interim Storage Facility in the United States
Authors: Lewis Donald Wayne 
Published: February 2014

In the United States (U.S.) the nuclear waste issue has plagued the nuclear industry for decades. Originally, spent fuel was to be reprocessed but with the threat of nuclear proliferation, spent fuel reprocessing has been eliminated, at least for now. In 1983, the Nuclear Waste Policy Act of 1982 [1] was established, authorizing development of one or more spent fuel and high-level nuclear waste geological repositories and a consolidated national storage facility, called a “Monitored Retrievable Storage” facility, that could store the spent nuclear fuel until it could be placed into the geological repository. Plans were under way to build a geological repository, Yucca Mountain, but with the decision by President Obama to terminate the development of Yucca Mountain, a consolidated national storage facility that can store spent fuel for an interim period until a new repository is established has become very important. Since reactor sites have not been able to wait for the government to come up with a storage or disposal location, spent fuel remains in wet or dry storage at each nuclear plant.
The purpose of this paper is to present a concept developed to address the DOE’s goals stated above. This concept was developed over the past few months by collaboration between the DOE and industry experts that have experience in designing spent nuclear fuel facilities.
The paper examines the current spent fuel storage conditions at shutdown reactor sites, operating reactor sites, and the type of storage systems (transportable versus non-transportable, welded or bolted). The concept lays out the basis for a pilot storage facility to house spent fuel from shutdown reactor sites and then how the pilot facility can be enlarged to a larger full scale consolidated interim storage facility.
Intra-Tester Reliability of Hand-Held Dynamometry and Strap-Mounted Dynamometry for Assessment of Ankle Strength
Authors: McGirr Kate Anne , Kennedy Thomas , Mølgaard Carsten M. , Rathleff Michael Skovdal 
Published: February 2014

Context:Measurement of ankle evertor and invertor strength is important for assessment of ankle sprain risk and assessment of restoration of normal function after an ankle sprain.Objective:To compare the intra-tester reliability of ankle evertor and invertor strength measurements derived from hand-held dynamometry (HHD) and strap-mounted dynamometry (SMD) for both make-test and break-test techniques.Participants:25 healthy individuals.Main Outcome Measure:Repeated measurements of isometric evertor and invertor strength derived from both HHD and SMD devices using both make-test and break-test techniques.Results:High make-test reliability values were found for both inversion and eversion force measured by both the HHD and SMD devices (ICC = 0.80 to ICC = 0.88). The correlation coeffcients between HHD and SMD measurements for the inversion make-test, the inversion break-test, and the eversion make-test ranged from r = 0.74 to r = 0.87, but the correlation between HHD and SMD measurements for the eversion break-test was poor.Conclusion:The fndings suggest that testers with limited experience can obtain reliable measurements of ankle evertor and invertor strength using either HHD or SMD.
Correlation Between Glycemic Control and Physical Activity Level in Adolescents and Children With Type 1 Diabetes
Authors: Miculis Cristiane Petra , De Campos Wagner , Boguszewski Margaret Cristina da Silva 
Published: February 2014

Background:The aim of this study was to correlate glycemic control (GC) and variables of physical activity levels (PAL) in children with type 1 diabetes mellitus (T1DM).Methods:Fifty children and adolescents with T1DM were selected. Personal and medical data for the patients were collected. Physical evaluations of body weight and sexual maturation were undertaken. Bouchard’s questionnaire was applied to evaluate PAL as well as for time spent on physical activities.Results:Sixty-four percent of the subjects were sexually mature. Differences were observed between females and males in insulin dose, duration of light physical activity, and sleeping time (P < .05). Ninety percent presented poor GC and 80% had a low PAL. Fasting blood glucose (FBG) was significantly correlated with PAL, with sedentary time, and with sleeping time. Glycated hemoglobin (HbA1c) was significantly correlated with sedentary time and sleeping time. Among the three groups of PAL (insufficient × moderate × active) there were differences in HbA1c (%), FBG (mg/dL), duration of disease (years), and insulin dose (UI/kg/day) (P < 0.001).Conclusion:GC was significantly correlated with PAL. Among the three groups of physical activity level, the most active group was seen to have the best GC.