bluuebunny/crossref_metadata_2025_split huggingface dataset:
Fundamental Study on Mitigation of Rocking Motion of a Cask System by a Gyro System
Authors: Ito Tomohiro , Ishikawa Yasumasa , Shintani Atsuhiko , Nakagawa Chihiro 
Published: February 2014

In Japan, ensuring the structural integrity of cask systems during seismic events is becoming increasingly important. Cask systems, which are free-standing cylindrical structures that contain spent fuel assemblies, are believed to exhibit rocking motions under strong seismic excitations. Thus far, analytical studies conducted by the authors have indicated that cask systems subjected to strong seismic motions, undergo large rocking motions, and, in the worst case, may overturn and collapse. Therefore, reducing the rocking motions of casks to avoid overturning and consequent contamination of radioactive substances is critical.
To suppress rocking motions for heavy free-standing structures such as cask systems, we propose a rocking motion suppression system that employs a gyro system. This system is installed in the free-standing structure. A previous analytical study showed that this system largely mitigates rocking motion.
In the present study, we fabricated a fundamental cask system model and a gyro system. By using the cask system model without a gyro system, free vibration tests and shaking table tests were conducted to understand the basic characteristics of rocking motion and responses under base excitations. Analyses were also conducted to confirm the validity of the analytical model for rocking motion comparing with the experimental data. Moreover, analyses for the cask system with the gyro system were conducted. From these results, we evaluated the ability of the gyro system to mitigate rocking motion.
Tailoring of the CARE Concept for Practicality, Safety and Robustness
Authors: Kawamura Hideki , McKinley Ian G. 
Published: February 2014

The CAvern REtrievable (CARE) concept was originally developed to provide flexibility in tailoring disposal of vitrified HLW to small volunteer host sites, where high emplacement density leads to significant challenges in terms of the management of radiogenic heat. Including an extended storage period in underground caverns prior to sealing solved the heat issue, but also provided benefits in terms of a slow decision-making process that provided more opportunities to engage stakeholders.
Recently, direct disposal of spent fuel is being considered in Japan and this gives more severe challenges for heat management, even for conventional disposal concepts. Due to great uncertainties in the future of nuclear power in Japan, this may also be associated with a desire to preserve the option of retrieval for extended periods. The basic CARE concept has thus been re-assessed to consider these aspects — and also the need to be able to develop a robust safety case that can be readily communicated to stakeholders, who are more sensitive to nuclear issues than they were in the past.
Plant Security During Decommissioning; Challenges and Lessons Learned From German Phase Out Decision
Authors: Renner Andrea , Esch Markus 
Published: February 2014

Purpose of this paper is to point out the security challenges that may occur during the decommissioning, based on the issues and lessons learned from the German phase out decision. Though national regulations may be different in other countries the basic problems and issues will be the same. Therefore presented solutions will be applicable in other countries as well.
The radioactive material remaining at the NPP during decommissioning has the most influence on how the security measures have to be designed. The radioactive material defines the risk potential of the plant and this determines the needed security level. The following aspects have been challenging in Germany:
• Scenarios varying from those, used for plants in operation, due to changed operating conditions
• Spent fuel will stay in the spent fuel pool for a quite long period before it can be removed from the plant. Risk potential of the plant stays high and requires a high level of security measures
• Security measures according to the existing operating license have to stay in place as they are, unless the first license for decommissioning is given respective the spent fuel is removed from the plant site. This even led to the question if improvements of security measures, planned and announced with focus on a plant remaining in operation for another couple of years, need to be done although they will not be required after removing the spent fuel from the plant.
A further important aspect for the security design is the fact that a plant under decommissioning has completely different and strongly varying operating procedures, compared to the stable ones of an operating plant. This leads to different needs concerning workspace, infrastructure on plant site, access to buildings etc. An optimized and highly flexible security concept is needed to ensure an adequate level of security as well as an efficient decommissioning. A deep analysis of the vital plant functions, depending on the different decommissioning stages, is required to determine the vital equipment, its location and its need for protection.
Investigation of Processing Advanced Waste Forms in Cold Crucible Induction Melter Systems
Authors: Roach Jay A. 
Published: February 2014

The Cold Crucible Induction Melter (CCIM) technology has been identified in several independent reviews and studies as a potential alternative or supplemental vitrification technology for specific U.S. Department of Energy (DOE) high level radioactive waste (HLW) streams. These inventories have unique chemistries, such as high alumina, high iron, high chromium, etc., that are not amenable to efficient immobilization in conventional borosilicate glass (BSG) matrices. Advanced waste forms, such as iron phosphate glass (FePG), high crystalline BSG (HC-BSG), and alumino-silicate glass-ceramic (GC) have been shown to provide significantly improved immobilization matrices for these challenging waste streams, including dramatically improved waste loadings; however, they are difficult to efficiently process in the baseline joule-heated ceramic melters (JHCMs). The CCIM technology can effectively process these advanced waste forms at the desired high waste loadings.
This paper describes the most recent collaborative activities between the DOE Office of Environmental Management (DOE-EM) and various Russian institutions in modeling, development and testing of advanced waste forms in CCIMs, including FePG and HC-BSG formulations. Additionally, past interactions with other international partners, including France and Korea, are described. Finally, recent enhancements to CCIM processing related to modeling, draining techniques and automated control, which were collaboratively developed, will be discussed.
Westinghouse Waste Simulation and Optimization Software Tool
Authors: Mennicken Kim , Aign Jörg 
Published: February 2014

Applications for dynamic simulation can be found in virtually all areas of process engineering. The tangible benefits of using dynamic simulation can be seen in tighter design, smoother start-ups and optimized operation. Thus, proper implementation of dynamic simulation can deliver substantial benefits. These benefits are typically derived from improved process understanding. Simulation gives confidence in evidence based decisions and enables users to try out lots of ‘what if’ scenarios until one is sure that a decision is the right one.
In radioactive waste treatment tasks different kinds of waste with different volumes and properties have to be treated, e.g. from NPP operation or D&D activities. Finding a commercially and technically optimized waste treatment concept is a time consuming and difficult task.
The Westinghouse Waste Simulation and Optimization Software Tool will enable the user to quickly generate reliable simulation models of various process applications based on equipment modules. These modules can be built with ease and be integrated into the simulation model. This capability ensures that this tool is applicable to typical waste treatment tasks.
The identified waste streams and the selected treatment methods are the basis of the simulation and optimization software. After implementing suitable equipment data into the model, process requirements and waste treatment data are fed into the simulation to finally generate primary simulation results. A sensitivity analysis of automated optimization features of the software generates the lowest possible lifecycle cost for the simulated waste stream.
In combination with proven waste management equipments and integrated waste management solutions, this tool provides reliable qualitative results that lead to an effective planning and minimizes the total project planning risk of any waste management activity. It is thus the ideal tool for designing a waste treatment facility in an optimum manner, taking account of the detailed waste stream and specific requirements.
Estimation of Carbon 14 Inventory in Hull and End-Piece Wastes From Japanese Commercial Reprocessing Operation
Authors: Sakuragi Tomofumi , Tanabe Hiromi , Hirose Emiko , Sakashita Akira , Nishimura Tsutomu 
Published: February 2014

Hull and end-piece wastes generated from reprocessing plant operations are expected to be disposed of in a deep underground repository as Group 2 TRU wastes under the Japanese classification system. The activated metals that compose the spent fuel assemblies such as Zircaloy claddings and stainless steel nozzles are mixed and compressed after fuel dissolution, and then stuffed into stainless steel canisters.
Carbon 14 is a typical activated product in the hulls and end-pieces and is mainly generated by the 14N(n,p)14C reaction. In the previous safety assessment of the TRU waste in Japan, the radionuclides inventory was calculated by ORIGEN-2 code. Some conservative assumptions and preliminary estimates were used in this calculation. For example, total radionuclides generated from a single type of fuel assembly (45 GWd/tU for a PWR unit), and the thickness of the Zircaloy oxide film on the hulls (80 μm) were both overestimated. The second assumption in particular has a large effect on exposure dose evaluation. Therefore, it is essential to have a realistic source term evaluation regarding such items as the C-14 inventory and its distribution to waste parts.
In the present study, a C-14 inventory of the hull and end-piece wastes from the operation of a commercial reprocessing plant in Japan corresponding to 32,000 tU (16,000 tU in each BWR and PWR) was calculated. Analysis using individual irradiation conditions and fuel characteristics was conducted on 6 types of fuel assemblies for BWRs and 12 types for PWRs (4 pile types × 3 burnup limits). The oxide film thickness data for each fuel type cladding were obtained from the published literature. Activation calculations were performed by using ORIGEN-2 code. For the amount of spent assembly and other waste characteristics, representative values were assumed based on the published literature. As a preliminary experiment, C-14 in irradiated BWR claddings was measured and found to be consistent with the calculated activation.
The total C-14 inventory was estimated as 4.46×1014 Bq, consisting of 2.58×1014 Bq for BWRs and 1.87×1014 Bq for PWRs, and is consistent with the safety assessment of 4.4×1014 Bq. However, the distribution of the C-14 inventory to hull oxide, which was estimated under the assumption of instantaneous radionuclide release in the safety assessment, decreased from 5.72×1013 Bq (13% of the total) in the previous assessment to 1.30×1013 Bq (2.9% of the total; consisting of 1.48×1012 for BWRs and 1.15×1013 for PWRs). In other words, the exposure dose peak is reduced to approximate 25% of its previous value due to the use of detailed oxide film data that the BWR cladding has a thin oxide film. Other instantaneous release components for C-14 such as the fuel residual were negligible.
Remediation of Uranium Impacted Sediments in a Watercourse
Authors: Shephard Eugene , Walter Nelson , Downey Heath , Collopy Peter , Conant John 
Published: February 2014

In 2009, remediation was initiated for a non-operational fuel cycle facility previously used for government contract work located in Windsor, Connecticut, USA. Radiological contaminants consisted primarily of high enriched uranium (HEU). Other radionuclides encountered in relatively minor amounts in certain areas of the clean-up included Co-60, Cs-137, Ra-226, Th-232 and low enriched uranium (LEU). Between 2009 and the spring of 2011, remediation efforts were focused on demolition of contaminated buildings and removal of contaminated soil. In the late spring of 2011, the last phase of remediation commenced involving the removal of contaminated sediments from portions of a 1,200 meter long gaining stream.
Planning and preparation for remediation of the stream began in 2009 with submittal of permit applications to undertake construction activities in a wetland area. The permitting process was lengthy and involved securing permits from multiple agencies. However, early and frequent communication with stakeholders played an integral role in efficiently obtaining the permit approvals. Frequent communication with stakeholders throughout the planning and remediation process also proved to be a key factor in timely completion of the project.
The remediation of the stream involved the use of temporary bladder berms to divert surface water flow, water diversion piping, a sediment vacuum removal system, excavation of sediments using small front-end loaders, sediment dewatering, and waste packaging, transportation and disposal. Many safeguards were employed to protect several species of concern in the work area, water management during project activities, challenges encountered during the project, methods of Final Status Survey, and stream restoration.
Effects of Precipitation on Surface-Scan Gamma Ray Survey Results: Case Study
Authors: Dorrell N. 
Published: February 2014

It has been known for many years that certain weather events (e.g., precipitation, low barometric pressure, etc.) can affect the results of outdoor gamma-ray surveys, particularly those where gamma spectroscopy is being used for the detection of uranium and its progeny. These effects are a result of a natural phenomenon that produces anomalous results that are contrary to the true concentrations present at the survey site. Gamma-ray survey results sometimes overestimate uranium concentrations during and immediately following rain or snowfall events.
The effects that a precipitation event has upon a drive-over gamma-ray survey are discussed in this paper. Surveys were conducted using a sensitive array of sodium iodide (NaI) detectors mounted to an all-terrain vehicle in late fall/early winter where snow was encountered. Isotope-specific measurements taken before and during precipitation events are compared and visually presented in iso-contour maps.
Demonstration Test Results of Organic Materials’ Volumetric Reduction Using Bio-Ethanol, Thermal Decomposition and Burning
Authors: Tagawa Akihiro , Watanabe Masahisa 
Published: February 2014

To discover technologies that can be utilized for decontamination work and verify their effects, economic feasibility, safety, and other factors, the Ministry of the Environment launched the “FY2011 Decontamination Technology Demonstrations Project” to publicly solicit decontamination technologies that would be verified in demonstration tests and adopted 22 candidates. JAEA was commissioned by the Ministry of the Environment to provide technical assistance related to these demonstrations. This paper describes the volume reduction due to bio-ethanol, thermal decomposition and burning of organic materials in this report. The purpose of this study is that to evaluate a technique that can be used as biomass energy source, while performing volume reduction of contamination organic matter generated by decontamination. An important point of volume reduction technology of contaminated organic matter, is to evaluate the mass balance in the system. Then, confirming the mass balance of radioactive material and where to stay is important. The things that are common to all technologies, are ensuring that the radioactive cesium is not released as exhaust gas, etc.. In addition, it evaluates the cost balance and energy balance in order to understand the applicability to the decontamination of volume reduction technology.
The radioactive cesium remains in the carbides when organic materials are carbonized, and radioactive cesium does not transfer to bio-ethanol when organic materials are processed for bio-ethanol production. While plant operating costs are greater if radioactive materials need to be treated, if income is expected by business such as power generation, depreciation may be calculated over approximately 15 years.
Potential Scenarios for Broadening Stakeholder Involvement in the Implementing Geological Disposal Technology Platform
Authors: Martell Meritxell , Bergmans Anne 
Published: February 2014

This paper analyses the potential for the involvement of different types of stakeholders in the Implementing Geological Disposal Technology Platform (IGD-TP). This analysis was conducted as part of the InSOTEC project, a three-year (2011–2014) collaborative research project funded under the 7th Euratom Framework Programme (Grant Agreement nr. 269906).
In our analysis, we consider the extent to which the IGD-TP’s practice as regards to stakeholder involvement matches its discourse, and what potential for improvement exists given its structural organisation as a European Technology Platform (ETPs). Technology Platforms (TPs) can be understood as knowledge networks, deliberately set up to influence (research) policy in a specific domain. We therefore use knowledge networks as a conceptual approach and look at the IGD-TP as a complex network which includes actors, knowledge and practices across different countries, focusing on a very specific topic (i.e. implementing geological disposal). We compare the way different stakeholders are involved in the IGD-TP to the practice of other ETPs, and explore how the IGD-TP is viewed by its members and by outsiders to the platform.
Applying Callon’s framework of knowledge co-production (1999) we come to define different degrees of interaction between science, society and policy in view of defining research and development (R&D) priorities [1]. Subsequently we describe how these interactions could be conceptualised and interpreted for the IGD-TP. The current approach of the IGD-TP can be mainly understood as classical model involving mainly expert stakeholders and scientists. Where there seems to be a good representation among IGD-TP members of industry, research institutes, and some members of the academic community this is not the case for other types of stakeholders, such as public authorities or civil society. At this stage, the overall approach of the IGD-TP would seem to restrict the scope of stakeholder involvement, as it narrows participation down to uniquely technology experts, hindering socio-technical manifestations. Our analysis nevertheless shows that there is room for engaging with a broader range of stakeholders in the field of radioactive waste management, if this is the intention of the IGD-TP. However, this would require a commitment to developing a common knowledge base including other stakeholders through a process of mutual adjustment and negotiation.
How an Integrated Change Programme Has Accelerated the Reduction in High Hazard Nuclear Facilities at Sellafield
Authors: Mackintosh Angela 
Published: February 2014

For over five decades the Sellafield Site has been central to the UK’s nuclear programme. Now operated by Sellafield Ltd, under the management of Parent Body Organisation Nuclear Management Partners (NMP), a consortium of URS Washington Division, AMEC and AREVA is focussed on the decommissioning of historical facilities.
When Decommissioning commenced in the late 1980’s the site focus at that time was on commercial reprocessing and waste management. Now through the implementation of a company change programme, emphasis has shifted towards accelerated risk and hazard reduction of degraded legacy plants with nuclear inventory whilst ensuring value for money for the customer, the Nuclear Decommissioning Authority.
This paper will describe the management success by the Site owners in delivering a successful change programme.
The paper will explain how the site has transitioned to the INPO Standard Nuclear Performance Model (SNPM) and how through the use of a change maturity matrix has contributed to the accelerated reduction in high risk high hazard nuclear facilities.
The paper will explain in detail how the Decommissioning Programme Office has facilitated and coordinated the Governance and assured delivery of the change plan and how successful application of visual management has aided the communication of its progress.
Finally, the paper will discuss how the Delivery Schedules have proved critical for presenting the change plan to Key Stakeholders, Government Owners and Powerful Regulators. Overall, this paper provides an insight into how a massive change programme is being managed within one of the world’s highest regulated industries.
Decommissioning of Magnox Ltd Fuel Cooling Pond Facilities in the UK
Authors: Bertoncini Carlo 
Published: February 2014

Magnox reactors were the first generation of nuclear power stations built in the UK; ten sites in total, of which, nine had wet fuel routes with cooling ponds. Five ponds are currently in a decommissioning phase; this paper will focus primarily on Hunterston-A (HNA) Site and the central programme of work which governs its management.
During its operation, the Cartridge Cooling Pond at HNA was used to receive the spent fuel discharged from the Site’s two reactors, it was then stored for cooling purposes prior to dispatch off site. The current decommissioning phase focusses on draining the 6500m3 pond. Due to the Site’s limited caesium removal facilities, a stand-alone effluent treatment plant was constructed to improve abatement and reduce the pond activity from 200 to 0.7 Bq/ml (β). This was necessary due to increased environmental standards introduced since the site had ceased generation ten years previously.
Early characterisation and experience from other sites concluded that if the pond were to be drained without any treatment to the walls, doses to the Operators, during subsequent decommissioning works, would routinely be in excess of 1mSv.hr−1(γ). An opportunity was realised within the Ponds Programme that if the surface layer of the pond walls were to be removed during drain-down, ambient dose rates would be reduced by a factor of 10; this would allow for more cost-effective decommissioning options in the future. Ultra-high pressure water jetting was tested and proved to yield a ∼95% total-activity reduction on treated surfaces. Challenges were overcome in providing safe and secure access to Decommissioning Operators to perform this operation by means of floating platforms on the surface of the pond.
As strategies to clear facilities to exemption levels are becoming both cost prohibitive and not reasonably practicable, work is now underway in the Programme to determine the optimum condition for entry into long-term quiescent storage, prior to final demolition. This paper will discuss the strategy and techniques which led to Magnox Ltd ponds to be of national and international interest to the nuclear community.