bluuebunny/crossref_metadata_2025 huggingface dataset:
Update on the Review of the LLWR 2011 Environmental Safety Case
Authors: Cummings Richard , Huntington Amy , Shevelan John , Baker Andrew J. , Sumerling Trevor , Taylor Frank
Published: February 2014

The UK Low Level Waste Repository has submitted a fully revised Environmental Safety Case (ESC) to the Environment Agency for the continued operation of the site. The Environment Agency is reviewing the submission. As part of the review of the ESC, we have been engaging with the Environment Agency to answer questions and provide further clarification where required. Once the review is complete, LLWR will apply for a revised permit for the continued operation of the site. We are required by our current Permit to operate the site in accordance with the assumptions of the ESC. We have developed a process for the implementation and maintenance of the ESC as a ‘live’ safety case under formal change control, and the development of waste acceptance arrangements identified as necessary to ensure that the repository is operated in a safe and optimised way, consistent with the assumptions and results of the ESC. Engagement with waste consignors has been essential in the development of revised waste acceptance criteria. Additional work has also been carried out in the development of an Article 37 submission, presenting the ESC to local stakeholders and developing of waste emplacement strategies.
Radioactive Waste Characterization and Tracking System at the French Atomic Energy Commission
Authors: Lamotte Hervé , Torreblanca Luc , Vilarel Suzel
Published: February 2014

CARAIBES is the traceability information system in radioactive solid waste management of the French Atomic Energy Commission, (CEA). Initially developed for the reprocessing facility of low and intermediate level waste of the centre of Cadarache, it has been gradually enhanced to manage very low level waste and the facilities of the centres of Saclay and Fontenay-aux-Roses. This software is directly connectable to characterization equipment as X-Ray measurement, gamma spectrometric measurement and/or neutronic measurement. After 15 years of operation, a new information system should replace it in 2014. This new CARAIBES V2 will make the functional and technical improvements resulting from the CARAIBES feedback but will also allow to manage all produced and/or treated waste in the Atomic Energy Commission and to ensure the traceability of waste and waste treatment before and after its packaging. From the analysis of the processes, waste management was described as an iterative process where, in every step, the facility in charge of a waste will apply a treatment to it (packaging, association or division) and will characterize it in order to evacuate it towards the next facility. Associated with a strongly customizable information system, this process allows the management of all kind of radioactive waste, including radioactive liquids. The final objectives are: to share a unique software and only one database for all CEA’s facilities involved in radioactive waste management, to be able to track the radioactive waste through all steps of its process, to meet the user’s needs as well as the regulatory requirements.
Development of Low-pH Cements for Immobilisation of Intermediate Level Radioactive Waste: Achievements and Challenges
Authors: Martins de Freitas Regeane , Al-Tabbaa Abir
Published: February 2014

Although cementation is a widely recognized solidification/stabilization process for immobilisation of Intermediate Level Radioactive Waste (ILRW), the low resistance to hyperalkaline pore waters compromises the effectiveness of the process when Portland Cement (PC) is employed. Moreover the manufacture of PC is responsible for significant CO2 emissions. In this context, low pH cements are environmentally more suitable and have emerged as a potential alternative for obtaining secure waste forms. This paper summarises the achievements on development of low-pH cements and the challenges of using these new materials for the ILRW immobilisation. The performance of waste forms is also discussed in terms of radionuclides release. Reactive magnesium oxide and magnesium phosphate cements are emphasised as they feature important advantages such as consumption of available constituents for controlling acid-base reactions, reduced permeability and higher density. Additionally, in order to identify new opportunities for study, the long-term modelling approach is also briefly discussed.
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.
Characterising Nuclear Simulant Suspensions In Situ With an Acoustic Backscatter System
Authors: Bux Jaiyana , Hunter Timothy N. , Paul Neepa , Dodds Jonathan M. , Peakall Jeffrey , Biggs Simon R.
Published: February 2014

In situ characterisation of radiotoxic sludges and slurries is critical to numerous operations including those involving their transport and retrieval. An inexpensive, flexible acoustic backscatter system has been employed for the first time here to a 4/10th scale active storage tank comprising of a nuclear simulant suspension, to verify its application. Intricate suspension characteristics and tank operation features emerged.
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.
A Reform in the Helium Purification System of the HTR-10: γ Dose Rate Measurement and Suggestions for Decommissioning
Authors: Xie Feng , Li Hong , Cao Jianzhu , Yu Suyuan , Zhang Liguo , Li Wenqian , Fang Sheng
Published: February 2014

A reform will be implemented in the helium purification system of the 10MW High Temperature Gas-cooled Test Reactor (HTR-10) in China. The measurement of the γ dose rates of facilities, including valves, pipes, dust filter, etc., in the purification system of the HTR-10, has been performed. The results indicated that most radiation nuclides are concentrated in the dust filter and facilities at the entrance of the helium purification system upstream of the dust filter. Other facilities have the same γ dose rate level as the background. Based on the previous study and experiences in AVR, the measurement results can be understood that the radioactive dust carried by the helium gas was filtered by the dust filter. It provides important insights for the decontamination and decommissioning of facilities in the primary loop, especially in the helium purification system of the HTR-10 as well as the High Temperature Reactor-Pebblebed Modules (HTR-PM).
Measurement of Activity of Spent Ionizing Sources Based on Cs-137 and Co-60 Placed in Well-Type Storage Facilities
Authors: Diordiy Mikhail N. , Nikolaev Oleg A. , Karlina Olga K. , Semenov Valeriy E. , Barinova Ella A.
Published: February 2014

In this paper the method of activity measurement of spent radionuclide sources (SRS) based on Co-60 and Cs-137 placed in well-type storage facilities, in order to forecast the SRS activity which can be placed at this facility is described. The distribution of the absorbed dose through the absorption material is measured with the help of the described absorption spectrometer. This information allows determining the activity of individual radionuclides with the help of mathematical methods.
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.
The Impact of Council Directive 2011/70/EURATOM and IAEA Joint Convention Review Meetings on the Ongoing Establishment of the Portuguese Regulatory Framework and on the Future of National Radioactive Waste
Authors: Paiva Isabel , Trindade Romão B.
Published: February 2014

Council Directive 2011/70/EURATOM of 19 July 2011, establishing a Community framework for the responsible and safe management of spent fuel and radioactive waste will enter in force August 2013 in all EU Member States. Portugal has already started preparing its legislative framework to accommodate the new legislative piece. However, the first report of Portugal to the Joint Convention on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste Management of the IAEA, in Vienna, 2012, has shown that Portugal still has many steps to overcome to establish a successful and effective basic regulatory framework. The existence of many competent authorities related to the radiological protection area and a newly independent commission that is still looking on how to fulfill its regulator role in other areas such as the radioactive waste management makes quite challenging the full application of the new directive as well as compliance that Portugal will have to show in the next Joint Convention review meeting in order to meet the obligations of the Convention. In this paper, the reality of the regulatory Portuguese framework on radiological protection, nuclear safety and radioactive waste management is presented. Discussion of the future impact of the new legislation and its consequences such as the need to setup the national program on radioactive waste management is critical discussed.
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.
The Status of LILW Disposal Facility Construction in Korea
Authors: Kim Min-Seok , Chung Myung-Sub , Park Kyu-Wan
Published: February 2014

In this paper, we discuss the experiences during the construction of the first LILW disposal facility in South Korea. In December 2005, the South Korean Government designated Gyeongju-city as a host city of Low- and Intermediate-Level Radioactive Waste (LILW) disposal site through local referendums held in regions whose local governments had applied to host disposal facility in accordance with the site selection procedures.

The LILW disposal facility is being constructed in Bongilri, Yangbuk-myeon, Gyeongju. The official name of the disposal facility is called ‘Wolsong Low and Intermediate Level Radioactive Waste Disposal Center (LILW Disposal Center)’. It can dispose of 800,000 drums of radioactive wastes in a site of 2,100,000 square meters. At the first stage, LILW repository of underground silo type with disposal capacity of 100,000 drums is under construction expected to be completed by June of 2014.

The Wolsong Low and Intermediate Level Radioactive Waste Disposal Center consists of surface facilities and underground facilities. The surface facilities include a reception and inspection facility, an interim storage facility, a radioactive waste treatment building, and supporting facilities such as main control center, equipment & maintenance shop. The underground facilities consist of a construction tunnel for transport of construction equipment and materials, an operation tunnel for transport of radioactive waste, an entrance shaft for workers, and six silos for final disposal of radioactive waste. As of Dec. 2012, the overall project progress rate is 93.8%.
Used Fuel Management System Architecture and Interface Analyses
Authors: Nutt Mark , Howard Robert , Busch Ingrid , Carter Joe , Delley Alexcia , Kalinina Elena , Hardin Ernest , Cotton Thomas
Published: February 2014

Preliminary system-level analyses of the interfaces between at-reactor used fuel management, consolidated storage facilities, and disposal facilities, along with the development of supporting logistics simulation tools, have been initiated to provide the U.S. Department of Energy (DOE) and other stakeholders with information regarding the various alternatives for managing used nuclear fuel (UNF) generated by the current fleet of light water reactors operating in the United States. An important UNF management system interface consideration is the need for ultimate disposal of UNF assemblies contained in waste packages that are sized to be compatible with different geologic media. Thermal analyses indicate that waste package sizes for the geologic media under consideration by the Used Fuel Disposition Campaign may be significantly smaller than the canisters being used for on-site dry storage by the nuclear utilities. Therefore, at some point along the UNF disposition pathway, there could be a need to repackage fuel assemblies already loaded and being loaded into the dry storage canisters currently in use. The implications of where and when the packaging or repackaging of commercial UNF will occur are key questions being addressed in this evaluation. The analysis demonstrated that thermal considerations will have a major impact on the operation of the system and that acceptance priority, rates, and facility start dates have significant system implications.
Dismantling the Nuclear Research Reactor Thetis
Authors: Michiels P.
Published: February 2014

The research reactor Thetis, in service since 1967 and stopped in 2003, is part of the laboratories of the institution of nuclear science of the University of Ghent. The reactor, of the pool-type, was used as a neutron-source for the production of radio-isotopes and for activation-analyses.

The reactor is situated in a water pool with inner diameter of 3 m. and a depth of 7.5 m. The reactor core is situated 5.3m under water level. Besides the reactor, the pool contains pneumatic loops, handling tools, graphite blocks for neutron moderation and other experimental equipment. The building houses storage rooms for fissile material and sources, a pneumatic circuit for transportation of samples, primary and secondary cooling circuits, water cleaning resin circuits, a ventilation system and other necessary devices. Because of the experimental character of the reactor, laboratories with glove boxes and other tools were needed and are included in the dismantling program. The building is in 3 levels with a crawl-space.

The ground-floor contains the ventilation installation, the purification circuits with tanks, cooling circuits and pneumatic transport system. On the first floor, around the reactor hall, the control-room, visiting area, end-station for pneumatic transport, waste-storage room, fuel storage-room and the labs are located. The second floor contains a few laboratories and end stations of the two high speed transfer tubes. The lowest level of the pool is situated under ground level. The reactor has been operated at a power of 150 kW and had a max operating power of 250 kW.

Belgoprocess has been selected to decommission the reactor, the labs, storage halls and associated circuits to free release the building for conventional reuse and for the removal of all its internals as legal defined.

Besides the dose-rate risk and contamination risk, there is also an asbestos risk of contamination. During construction of the installation, asbestos-containing materials were used, which must be removed in controlled conditions. The ventilation system is considered free from nuclear contamination but it contains asbestos.

This paper covers the organization of the dismantling work, the technical execution aspect and conclusions already known (dismantling is ongoing as this is written).
Impact of Advanced Fuel Cycles on Uncertainty Associated With Geologic Repositories
Authors: Rechard Rob P. , Lee Joon , Sutton Mark , Greenberg Harris R. , Robinson Bruce A. , Nutt W. Mark
Published: February 2014

This paper provides a qualitative evaluation of the impact of advanced fuel cycles, particularly partition and transmutation of actinides, on the uncertainty associated with geologic disposal. Based on the discussion, advanced fuel cycles, will not materially alter (1) the repository performance, (2) the spread in dose results around the mean, (3) the modeling effort to include significant features, events, and processes in the performance assessment, or (4) the characterization of uncertainty associated with a geologic disposal system in the regulatory environment of the United States.
Experiment on the Treatment of Waste Extraction Solvent From the Molybdenum-99 Process
Authors: Hsiao Hsien-Ming , Hu Chang-Liang , Chien Kuang-Li , Lee Wen-Chen , Wei Tsong-Yang
Published: February 2014

In the Mo-99 (Molybdenum-99) isotope extraction test process for radiopharmaceutical applications, organic solvent is used to extract Mo-99 from an irradiated UO2 dissolution. The extraction solvent was stored when the test work was stopped. A total of about 120 liters of waste solvent was stored at INER (Institute of Nuclear Energy Research, Taiwan). The extraction solvent consisted of 5% di-(2-ethylhexyl)-phosphoric acid (D2EHPA) and kerosene. The radionuclides found in the waste solvent include Cs-137, Am-241, Tc-99, and Sr-90, which give off gross alpha and beta radioactivity of 1898 and 471 Bq/ml, respectively. This study aims to remove radionuclides from the waste solvent using sodium carbonate and sodium hydroxide solutions in different concentrations. After mixing the waste solvent with the alkaline solution followed by settling, a third phase other than organic and aqueous phase appeared which is expected due to the saponification reaction. The experimental results showed that increasing the number of washing and the alkaline solution concentration could enhance the radionuclides removal rate. An optimal removal method was proposed using 2M Na2CO3 solution twice followed by 1M NaOH solution one time for the third phase generated early in the mixing stages. The remaining gross alpha and beta radioactivity of the treated organic solvent was 2 and 3 Bq/ml, respectively. The treated solvent could be stabilized by ashing at 500°C and then immobilized. The alkaline solution would be neutralized by hydrochloric or nitric acid and then treated using a variety of adsorbents or bone char via adsorption to remove nuclides to meet the wastewater discharge limitation.
Prediction of Radiation Doses During the Dismantling of the Pressurized Tank From Emergency Core Cooling System of RBMK-1500 Reactor
Authors: Simonis A. , Poskas P. , Poskas G.
Published: February 2014

Preparation for the decommissioning of the Ignalina Nuclear Power Plant involves multiple problems. Personnel radiation safety during the performance of dismantling activities is one of them. In order to assess the optimal personnel radiation safety, the modelling is performed for large components by the means of computer code “VISIPLAN 3D ALARA Planning tool” developed by SCK CEN (Belgium).

Modelling results of radiation doses during the dismantling of the pressurized tank from the emergency core cooling system (ECCS PT) of RBMK-1500 reactor are presented in this paper. The mass of one ECCS PT is approximately 47.6 tons. The radiological surveys indicate that the inner surface of the ECCS PT is contaminated with radioactive products of corrosion and sediments due to the radioactive water.

The assessment of workers exposure was performed to comply with ALARA. The effective doses to the workers were modeled for different strategies of ECCS PT dismantling. The impact of dismantling tools and shielding types and extract ventilation flow rate during the dismantling of ECCS PT on effective doses were analyzed. The total effective personnel doses were obtained by summarizing the effective personnel doses from various sources of exposure, i. e., direct radiation from radioactive equipment, internal radiation due to inhalation of radioactive aerosols, and direct radiation from radioactive aerosols arising during hot cutting in premises.
Surface and Subsurface Decontamination Technology
Authors: Tuck Matt , Wray Beth , Musgrave Mark
Published: February 2014

A number of proven technologies applicable to the chemical and physical decontamination of radioactive and non-radioactive contaminants within the environmental remediation and radiological waste management sectors exist. Previous work generally acknowledges that these methods have limitations such as production of large volumes of waste, destruction of the substrate, complex safety considerations [1a] and application of special precautions to meet disposal acceptance criteria [2].

A method that removes a variety of contaminants from the surface and subsurface of porous materials, with minimal contaminated waste arisings, is highly desirable.

TechXtract® is a patented, sequential chemical extraction process developed to remove radionuclides, PCBs, and other hazardous organic and inorganic substances from solid materials such as concrete, brick, steel, and exotic metals [3]. The technology uses multifarious task-specific chemical formulations and engineered applications to achieve surface penetration and removal of the contaminants from the atomic voids of metals and other substrates, or the capillaries and gel pores of concretes.

TechXtract® is proven to remove a variety of contaminants from various substrates, allowing free release of the substrate as waste for disposal, or re-use, whilst producing minimal secondary waste.

Data from testing of TechXtract’s capabilities and evidencing the technology’s efficacy during site based applied research and development is presented here.
Agricultural Land Management Options Following Large-Scale Environmental Contamination: Evaluation for Fukushima Affected Agricultural Land
Authors: Vandenhove Hildegarde
Published: February 2014

The accident at the Fukushima Daiichi Nuclear Power Plant has raised questions about the accumulation of radionuclides in soils, the transfer in the foodchain and the possibility of continued restricted future land use. This paper summarizes what is generally understood about the application of agricultural countermeasures as a land management option to reduce the radionuclides transfer in the food chain and to facilitate the return of potentially affected soils to agricultural practices in areas impacted by a nuclear accident.
Geostatistical Mapping of Cs-137 Contamination Depth in Building Structures by Integrating ISOCS Measurements of Different Spatial Supports
Authors: Rogiers B. , Boden S. , Jacques D.
Published: February 2014

Reliable methods to determine the contamination depth in nuclear building structures are very much needed for minimizing the radioactive waste volume and the decontamination workload. This paper investigates the geostatistical integration of in situ gamma-ray spectroscopy measurements of different spatial supports. A case study is presented from the BR3 decommissioning project, yielding an estimated reduction of waste volume of ∼35%, and recommendations are made for future application of the proposed methodology.
Design Innovation for the Management of Alpha Contaminated Unserviceable Glove Boxes
Authors: Sandhanshive Devendra , Shendge Shivaji , Pol S. K. , Nair K. N. S. , Wattal P. K. , Sarjekhan I. A. , Kumar Arun
Published: February 2014

With the maturing of nuclear industry, there is an added burden on the Back End of fuel cycle. Radioactive facilities are in the need for refurbishment. This paper describes the steps adopted for managing such alpha contaminated unserviceable Glove Boxes.

The first step consisted of in-situ encasement of individual Glove Boxes, encountering the challenges of low head room and space congestion in these laboratories with cognizance to regulatory requirement related to radiation safety. The second step was removal, transfer and placement of encased Glove Boxes in a dedicated facility under continuous surveillance. The glove boxes will remain stored in this facility until arrangements are completed for dismantling and volume reduction in another facility dedicated for this purpose. The final step is the development of an appropriate technique for dismantling/cutting of Glove Boxes in an alpha-tight facility constructed to prevent spread of airborne activity, collection of cut pieces, subsequent decontamination of these metallic wastes and disposal. The paper describes the design scheme for dismantling of the glove box. The description also includes hands on evaluation of tools and gadgets in pilot set-up with a view to incorporating the most credible choice in an upcoming active facility.
A Step-by-Step Development of Radioactive Waste Management Strategy
Authors: Lytsy Tero , Ylä-Mella Mia
Published: February 2014

Fennovoima is Finland’s third nuclear power company founded in 2007. Fennovoima will build a new nuclear power plant, Hanhikivi 1, in Pyhäjoki municipality located in Northern Finland. Currently Fennovoima is evaluating bids from reactor suppliers. Originally two suppliers, AREVA and Toshiba Heavy Industries were invited to bid for the plant but later also a bid from Rosatom were invited. The plant supplier will be selected in 2013. Platom is a Finnish company with extensive experience in radioactive waste management. In the beginning of 2008, Platom was assigned by Fennovoima as radioactive waste management consultant and to develop radioactive waste management strategy and to support Fennovoima in negotiations with plant suppliers in waste management aspects.

Since Fennovoima has started, there have been some changes to the regulatory requirements which had to be taken into account while developing the strategy. One important change is due to the Government Degree 27.11.2008/736 which establishes a new waste category of very low-level waste following international development and the “new” IAEA waste classification. Other important change was introduced by the change 342/2008 to the Nuclear Energy Act allowing some nuclear waste to be delivered to another country for treatment. These and many more requirements had to be considered when developing the strategy.

Early work by Platom was mostly done to support development of application for Decision-in-Principle. This was the first important licensing step for Fennovoima. Work started with studies and plans which were used by Fennovoima to gather the know-how required to draw up the application. Descriptions of waste streams and waste management technologies were developed as well as preliminary waste inventories and studies for final disposal, including preliminary dimensioning of the repository facilities. Based on these plans nuclear regulator’s preliminary safety assessment was performed. The work was successfully completed as in 2010 the Government of Finland granted Fennovoima a permit to build a new nuclear power plant.

The work continued in 2012 when bids were received for EPR and ABWR type reactors. Plans developed earlier were elaborated into a detailed strategy and the proposed waste management solutions were evaluated. This allowed feedback from the bids into the strategy and to the bid specification and they both were developed. Also waste inventories were elaborated which allowed development of reactor design specific disposal strategies.

The main objective of work has been to establish basis for safe and efficient radioactive waste management which meets all the relevant national and international recommendations, requirements and regulations, takes advantage of best available and state-of-art technologies and offers fit-for-purpose solutions. All the work was performed to accommodate requirement based management system.