bluuebunny/crossref_metadata_2025 huggingface dataset:
Advanced Simulation Capability for Environmental Management (ASCEM)
Authors: Dixon Paul , Williamson Mark , Freshley Mark , Moulton David , Gorton Ian , Keating Elizabeth , Collazo Yvette , Gerdes Kurt , Meza Juan
Published: August 2012

The United States Department Energy (DOE) Office of Environmental Management (EM) determined that uniform application of advanced modeling in the subsurface could help reduce the cost and risks associated with its environmental cleanup mission. In response to this determination, the EM Office of Technology Innovation and Development (OTID), Groundwater and Soil Remediation (GW&S) began the program Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is a state-of-the-art scientific tool and approach for integrating data and scientific understanding to enable prediction of contaminant fate and transport in natural and engineered systems. This initiative supports the reduction of uncertainties and risks associated with EM’s environmental cleanup and closure programs through better understanding and quantifying the subsurface flow and contaminant transport behavior in complex geological systems. This involves the long-term performance of engineered components, including cementitious materials in nuclear waste disposal facilities that may be sources for future contamination of the subsurface. This paper describes the ASCEM tools and approach and the ASCEM programmatic accomplishments completed in 2010 including recent advances and technology transfer.
Managing Aging Effects on Used Fuel Dry Cask for Very Long-Term Storage
Authors: Chopra Omesh , Diercks Dwight , Ma David , Shah Vikram , Tam Shiu-Wing , Fabian Ralph , Liu Yung , Nutt Mark
Published: August 2012

The cancellation of the Yucca Mountain repository program in the Unites States raises the prospect of very long-term storage (i.e., >120 years) and deferred transportation of used fuel at the nuclear power plant sites. While long-term storage of used nuclear fuel in dry cask storage systems (DCSSs) at Independent Spent Fuel Storage Installations (ISFSIs) is already a standard practice among U.S. utilities, recent rule-making activities of the U.S. Nuclear Regulatory Commission (NRC) indicated additional flexibility for the NRC licensees of ISFSIs and certificate holders of the DCSSs to request initial and renewal terms for up to 40 years. The proposed rule also adds a requirement that renewal applicants must provide descriptions of aging management programs (AMPs) and time-limited aging analyses (TLAAs) to ensure that the structures, systems, and components (SSCs) that are important to safety in the DCSSs will perform as designed under the extended license terms. This paper examines issues related to managing aging effects on DCSSs for very long-term storage (VLTS) of used fuels, capitalizing on the extensive knowledge and experience accumulated from the work on aging research and life cycle management at Argonne National Laboratory (ANL) over the last 30 years. The technical basis for acceptable AMPs and TLAAs is described, as are generic AMPs and TLAAs that are being developed by Argonne under the support of the U.S. Department of Energy (DOE) Used Fuel Disposition Campaign for R&D on extended long-term storage and transportation.
New Free-Release and Sorting Monitors Developed for NPP A-1 Decommissioning, Slovakia
Authors: Sla´vik Ondrej , Slaninka Alojz , Lisˇtjak Martin , Krava´rik Kamil , Pe´ly Igor
Published: August 2012

A pilot free-release monitoring post with a 600 L container monitor was developed and metrological tested within the 2nd stage of NPP A1 decommissioning project. In order to reduce the volume of contaminated soil monitoring a conceptual design of fast sorting loader shovel monitor (loader’s spoon) was proposed and tested within the project as well. The free-release monitoring post makes use of a pair of electrically cooled lead shielded semiconductor detectors placed into a mounting rack ensuring measurements in either horizontal (container monitoring) or vertical (drum monitoring) counting geometry. For evaluation of measured HPGe spectra the Canberra – Packard ISOCS detection efficiency calculation code was used. A loader is used to change the measured side of the 600 L container. For metrological certification of this monitor a special prototype test container with 24 rod sources inside a regular grid was necessary to design and to use. The mentioned above vertical counting geometry together with an additional drum rotator ensures standard free release monitoring of materials in 200 l rotating drums. Successful metrological qualification of the both counting geometries at SMU Bratislava showed 20% accuracy class. A pair of NaI(Tl) detectors and a measurement and navigation frame ensuring loader shovel fixation in counting position are used for the fast sorting monitoring. The navigation of loader and its shovel to the counting geometry should be as fast as possible. The monitored results shall be indicated by a prompt light indication system (apart from storing on HDD). MCNP 5 calculation code was used for assessment of gross gamma 137Cs detection efficiency. The estimated MDA for a pair of 2″ × 2″ Na(Tl) detectors and 30 s acquisition time is about 90 Bq/kg. However, due to the counting geometry deviations from calculated values the uncertainty of measurements can be relatively high. Hence, the system is applicable for sorting monitoring only.
Sodium Nonatitanate for Removal Radioactive Strontium From Aqueous Contaminated Effluent
Authors: Merceille Aure´lie , Grandjean Agne`s , Barre´ Yves
Published: August 2012

Sodium titanate materials which have a layered structure consisting of titanate layers and exchangeable sodium ions are promising inorganic ionic exchangers for strontium adsorption from aqueous solution. The materials used in this study include samples synthesized in a hydrothermal method at different temperatures between 60°C and 200°C. Batch kinetics experiments for strontium removal from aqueous solution were performed. Kinetics data were fitted by using a pseudo second order reaction model and a diffusive model. It was shown that the sorption process occurs in one or two diffusion-controlled steps that depend on the synthesis temperature. The strontium extraction capacity depends on the synthesis temperature with a maximum for sample synthesis at 100°C. This sample has good properties of decontamination at laboratory scale.
Validation of Correlations Between Nd Isotopes and Difficult-to-Measure Nuclides Predicted With Burn-Up Calculation Code by Postirradiation Examination
Authors: Asai Shiho , Okumura Keisuke , Hanzawa Yukiko , Suzuki Hideya , Toshimitsu Masaaki , Inagawa Jun , Kimura Takaumi , Kaneko Satoru , Suzuki Kensuke
Published: August 2012

Correlations between Nd isotopes and difficult-to-measure (DTM) nuclides, such as Se-79, Tc-99, Sn-126, and Cs-135, predicted using a calculation code have been validated by postirradiation examination (PIE). The calculation was performed with a burnup calculation code, MVP-BURN, using the updated nuclear data library JENDL-4.0. An irradiated PWR fuel with a burnup of 44.9 GWd/t and a cooling time of 7458 days was used as a standard sample. The concentrations of Nd isotopes in the sample solutions were determined by inductively coupled plasma mass spectrometry (ICP-MS) after purification by two consecutive anion-exchange separations. The ratios of Se-79 and Cs-135 to Nd isotopes (Nd-145, Nd-146, and Nd-145+Nd-146) calculated using MVP-BURN were in good agreement with PIE values within the deviations of 8% and 6%, respectively. This indicates that these calculated ratios are applicable to the scaling factors of Se-79 and Cs-135. For Tc-99 and Sn-126, the calculated values were respectively about 50% and 20% higher than the PIE values. These overestimations were mainly caused by the lack of the contribution of an insoluble residue to the measured concentrations.
Technical Know-How of Selection Process for the Horonobe Underground Research Laboratory Area and Site
Authors: Kunimaru Takanori , Takeuchi Ryuji , Matsuzaki Tatsuji
Published: August 2012

This study demonstrates the selection process for the Horonobe URL based on surveys of existing information and geophysical surveys on a regional scale. In addition, preliminary requirements on the geological environment, safety (during construction of the underground facility) and social and environmental constraints were taken into consideration. The technical know-how utilised through the experiences for the site selection is described here. The proposed Horonobe URL site required the existence of argillaceous sedimentary formations and associated groundwater. Further fundamental requirements were appropriate rock mechanical properties and low gas content in the host rock to meet safe underground construction and operation regulations. This led to a stepwise narrowing down from several potential URL areas located completely within the Horonobe District to one candidate URL area and, finally, to a specific URL site. In the URL investigation area (ca. 3 km × 3 km) the main surface-based investigations were conducted as the first step to choosing the actual URL site. This was selected based on establishing fundamental factors related to the geological environment, safety and societal issues. This paper provides an outline of the process utilised in selecting the URL site by taking into consideration technical and social requirements. Thus stepwise approach and experience in selecting the URL site will be applicable when NUMO needs to select a site through literature surveys, and preliminary and detailed investigations in the future.
Technical Know-How of Site Descriptive Modeling for Site Characterization
Authors: Saegusa Hiromitsu , Niizato Tadafumi , Yasue Ken-ichi , Onoe Hironori , Doke Ryosuke
Published: August 2012

The site descriptive model covering the current status of characteristics of geological environment and the site evolution model for estimation of the long-term evolution of site conditions are used to integrate multi-disciplinary investigation results. It is important to evaluate uncertainties in the models, to specify issues regarding the uncertainties and to prioritize the resolution of specified issues, for the planning of site characterization. There is a large quantity of technical know-how in the modeling process. It is important to record the technical know-how with transparency and traceability, since site characterization projects generally need long duration. The transfer of the technical know-how accumulated in the research and development (R&D) phase to the implementation phase is equally important. The aim of this study is to support the planning of initial surface-based site characterizations based on the technical know-how accumulated from the underground research laboratory projects. These projects are broad scientific studies of the deep geological environment and provide a technical basis for the geological disposal of high-level radioactive wastes. In this study, a comprehensive task flow from acquisition of existing data to planning of field investigations through the modeling has been specified. Specific task flow and decision-making process to perform the tasks have been specified.
Remote Radiation and Visual Surveys of the Hot Cell Waste Vault and Spent Fuel Transfer Route at Kazakhstan’s BN-350 Reactor
Authors: Wells David , Herrick Andrew , Klepikov Alexander , Yakovlev Igor , Tur Evgeniy , Knight Collin
Published: August 2012

Kazakhstan’s BN-350 fast reactor was shut down in 1999 and is in the process of being decommissioned in preparation for Safestore. A key achievement during 2010 was the removal of the known inventory of spent nuclear fuel (SNF) from the reactor to off-site secure storage. As a complementary activity, surveys of areas of the fuel discharge route where it was considered possible for fuel pins or fuel residues to have collected over many years of operation were also arranged to confirm that no significant amounts of fuel of remained at the plant. This paper reports on the remote radiation and visual (video and still photograph) surveys undertaken on the following areas of the BN-350 spent fuel route: • The waste storage repository or vault underneath the Post-Irradiation Examination Hot Cell in which non-destructive and destructive examination of irradiated fuels was undertaken throughout the operating life of the reactor. • The fuel transfer and washing cells within which irradiated fuel sub-assemblies were processed and residual sodium coolant removed. • The fuel storage ponds. Man access to several areas (particularly the waste vault) was not possible due to very high radiation levels from stored βγ-active wastes or residual contamination and in these cases specially engineered remote camera and radiation detector deployment systems were developed and used. In other areas, such as the ponds, limited man access was possible under prepared and controlled conditions. The survey results, together with associated radiochemical and radiation dose rate analysis, demonstrated that, despite several recorded handling incidents during Hot Cell operations, the maximum estimated amounts of nuclear material which could remain at BN-350 were sufficiently low to give no major safeguards concerns. The data also provides key information to guide decommissioning and dismantling planning in the future.
Ultrasonic Techniques for the In Situ Characterisation of ‘Legacy’ Waste Sludges and Dispersions
Authors: Hunter Timothy , Biggs Simon , Young James , Fairweather Michael , Peakall Jeff
Published: August 2012

Research being undertaken at the University of Leeds, as part of the DIAMOND university consortium, is exploring the effectiveness of various ultrasonic technologies as in situ probes to characterize and monitor nuclear waste slurries, such as the ‘Legacy’ Magnesium hydroxide sludges found in Sellafield, U.K. Through use of a commercial Acoustic Backscatter Sensor (ABS) with 1–5 MHz transducers, various properties of free-settling oxide simulant sludges were determined. Work was focused upon characterizing essentially ‘static’ sludges (to give prospective use as tools for the wastes in current deposits); although, the sensors also have potential as dispersion monitors during any future processing and storage of the Legacy wastes, as well as many other storage, clarifier or thickener systems across a wide range of industrial processing operations. ABS data of mixed glass powder dispersions was analysed and compared to scattering theory, to understand the correlations between acoustic attenuation and particulate concentration. The ABS was also calibrated to measure changes in average particulate concentration within a settling suspension over time, and showed the depth-wise segregation of the dispersion through the settling column at different particular time intervals. It was found that observed hindered settling also led to an increase in particulate concentration over the sludge zone and significant segregation occurred at moderate time intervals, due to the broad size distribution of the aggregates measured. It is hoped in future that these sensors may be able to be fitted to robotic handlers that have been installed onsite (and previously used for sampling), allowing fully automatic in situ sludge analysis.
A Dialogue-Centric Approach to Environmental-Remediation Decision-Making
Authors: Boucher Laurel , Clark James F.
Published: August 2012

This paper outlines a multi-step approach to streamline and enhance the decision-making process that guides environmental remediation. The inability of the responsible party and the various stakeholders to reach agreement on the remediation plan can delay the remediation, result in financial penalties, and lead to the development of an adversarial stance that inhibits the ability of the parties to work together in a creative and constructive manner. The approach presented by the authors is designed to expand dialogue in a way that moves it beyond technical or fiscal matters by addressing what the authors describe as the “hidden barriers” to productive dialogue. These “hidden barriers” include: self-interests, the perception as to how people are being treated, a lack of clarity or poor management of responsibilities and accountabilities, unclear or convoluted communication protocols, and an underlying tone of conflict and cynicism. A key element of the multi-step approach outlined in this paper is the process of uncovering these “hidden barriers” and addressing them in a way that turns discourse into collaboration. The paper describes a model the authors have used to streamline and enhance the process of creating sustainable agreements both for the U.S. Department of Energy’s Office of Environmental Management as well as the U.S. Department of Defense for a variety of environmental remediation projects. The results of this approach include the acceleration of an environmental clean-up from a projected 19 years to 11 years, the development of innovative technical strategies, the reduction of a major backlog of environmental proposals requiring review and comment, and the distinction accorded one group of being recognized as a model of effective partnering. The approach described has widespread implications not only because its use can be expanded to include a multitude of decision-making applications but also because of the impact it creates by expanding both the management and leadership skills of those who use it.
Advanced Drying System to Meet Radiolysis Criteria of Metallic Waste
Authors: Obata Masamichi , Ooishi Yasushi
Published: August 2012

High dose rate metal waste from core internals, such as BWR channel boxes or control rods are stored in spent fuel pool or site bunker pool. Waste form for final disposal of these high dose rate metal wastes should eliminate water to prevent hydrogen gas generation which can be caused by radiolysis of water. Toshiba’s newly developed drying system for high dose rate metal wastes enable very low content of water even if the waste have a powder layer. Toshiba will provide the total system to fabricate the waste form for high dose rate metal wastes. An operational condition to dry the B4C control rod which is the typical waste with powder layer was estimated based on experimental data.
Evaluation of External Exposure During Building and Operation of Concrete Bridges Constructions That Reuse the Conditionally Released Steels
Authors: Panik Michal , Necas Vladimir
Published: August 2012

This paper presents ongoing results of the project presented at ICEM’10 [1] related to the topics of reusing the conditionally released materials from decommissioning. The subject of the reuse of conditionally released materials in this case is modeling of bridge constructions which reuse the conditionally released steel in the form of reinforcement bars for the concrete bridges. A general approach for the project was presented at ICEM’10. The activities of the project continue in evaluating the individual effective doses from the external exposure based on reused conditionally released steels separately for public and for professionals (the internal exposure will be evaluated in next stages of the project). Evaluated scenarios are related to critical groups of professionals constructing the bridges (worker’s scenarios). The computer code VISIPLAN 3D ALARA 4.0 planning tool was used for the calculation of the individual effective dose for professionals. Various limits of the annual individual effective dose are used for the evaluation of calculation results. The aim of the ongoing modeling is to develop a set of data of maximal radioactivity concentration for individual radionuclides in the conditionally released steel used in the bridges model constructions in order not to exceed the limits for the individual effective dose.
Physical Modelling of Turbulent Jets for Nuclear Sludge Mobilisation
Authors: McKendrick D. , Biggs S. R. , Fairweather M. , Young J.
Published: August 2012

The impingement of a fluid jet onto a surface has broad applications across many industries. Within the UK nuclear industry, during the final stages of fuel reprocessing, impinging fluid jets are utilised to mobilise settled sludge material within storage tanks in preparation for transfer and ultimate immobilisation through vitrification. Despite the extensive applications of impinging jets within the nuclear and other industries, the study of two-phase, particle -laden, impinging jets is limited, and generally restricted to computational modelling. Surprisingly, very little fundamental understanding of the turbulence structure within such fluid flows through experimental investigation is found within the literature. The physical modelling of impinging jet systems could successfully serve to aid computer model validation, determine operating requirements, evaluate plant throughput requirements, optimise process operations and support design. Within this work a method is considered, capable of exploring the effects of process and material variables on the flow phenomena of impinging jets. This is achieved on a number of experimental test rigs of varying scale employing both intrusive and non-intrusive measurement techniques Particle image velocimetry (PIV), ultrasonic Doppler velocity profiling (UDVP) and high speed imaging, through to visual observations and direct measurements, are all techniques that can be deployed. The influence of a number of parameters on the erosion characteristics of sediment beds following application of an axisymmetric impinging jet is presented in detail. Bed erosion is found to be enhanced as the jet height above the sediment bed is increased, due to greater turbulence development. Different erosion characteristics, as jet outlet velocity increased, were found for the particulates tested; sand, fine Mg(OH)2 (test simulant representative of waste sludge, has similar particle size to sand, 200–1000μm) and coarse Mg(OH)2 (1000–2000μm). The crater diameter increased with increasing velocity as expected. However, the effect of the increase in velocity on the crater depth was very different, particularly for the coarse material which was found to re-deposit in the crater when the velocity increased above 1.3 ms−1, most likely due to enhanced re-circulation at the higher velocities.
A Numerical Interpretation of Density Homogenization of Bentonite Material in Wetting Process
Authors: Iizuka Atsushi , Takayama Yusuke , Tachibana Shinya , Ohno Shintaro , Kobayashi Ichizo , Kawai Katsuyuki
Published: August 2012

Due to its low permeability and excellent expansion characteristics, bentonite is an excellent candidate with potential use as a buffer in the disposal of nuclear waste. Its expansion characteristics, activated by wetting, can be interpreted based on the full saturation line, depicted as a unique line on the density and the confining pressure relationship as proposed by Kobayashi et al. (2007). In addition, its elasto-plastic constitutive relation can also be formulated by introducing additional irreversible strain component describing the expansion of the montmorillonite present in the bentonite material. A constitutive model can consistently express the mechanical behavior of the compacted bentonite material from the unsaturated to the fully saturated state. This paper describes the density homogenization process that was conducted through a series of soil-water coupled elasto-plastic finite element simulations. Specifically, bentonite specimens, with different initial densities, were permeated with a constant water head. Stresses and strains developing in bentonite, particularly the density change, were carefully examined. A series of numerical simulations, performed on the two specimens, showed that specimens did not homogenize to a unique value of density upon reaching the fully saturated state. To confirm the simulation results, we carried out a series of experiments. The experimental results also support our simulation results.
Development of an Engineering Design Process and Associated Systems and Procedures for a UK Geological Disposal Facility
Authors: Rendell Philip , O’Grady Henry , Breen Brendan , Clark Alastair , Reece Steve
Published: August 2012

In the United Kingdom the Nuclear Decommissioning Authority (NDA) has been charged with implementing Government policy for the long-term management of higher activity radioactive waste. The UK Government is leading a site selection process based on voluntarism and partnership with local communities interested in hosting such a facility and as set out in the ‘Managing Radioactive Waste Safely’ White Paper (2008). The NDA has set up the Radioactive Waste Management Directorate (RWMD) as the body responsible for planning, building and operating a geological disposal facility (GDF). RWMD will develop into a separately regulated Site Licence Company (SLC) responsible for the construction, operation and closure of the facility. RWMD will be the Design Authority for the GDF; requiring a formal process to ensure that the knowledge and integrity of the design is maintained. In 2010 RWMD published ‘Geological Disposal - Steps towards implementation’ which described the preparatory work that it is undertaking in planning the future work programme, and the phases of work needed to deliver the programme. RWMD has now developed a process for the design of the GDF to support this work. The engineering design process follows a staged approach, encompassing options development, requirements definition, and conceptual and detailed designs. Each stage finishes with a ‘stage gate’ comprising a technical review and a specific set of engineering deliverables. The process is intended to facilitate the development of the most appropriate design of GDF, and to support the higher level needs of both the project and the community engagement programmes. The process incorporates elements of good practices derived from other work programmes; including process mapping, issues and requirements management, and progressive design assurance. A set of design principles have been established, and supporting design guidance notes are being produced. In addition a requirements management system is being implemented for the identification, capture, analysis, update, verification, validation and acceptance of requirements for the GDF. This is to ensure that there are traceable links between requirements, and to identify and record the verification/validation of individual requirements. This paper describes the engineering design process and the supporting documents, systems and procedures. The paper addresses the relationship to the geological disposal programme timeline in ‘Geological Disposal - Steps towards implementation’ and, from there, to the UK Government ‘Managing Radioactive Waste Safely’ Programme. It also describes the next steps in the development of the design process, and some of the lessons learnt to date.
Characterization of Iraq’s Remote Nuclear Facilities for Decommissioning and Waste Management
Authors: Al-Musawi Fouad , Jarjies Adnan , Miller Ross A.
Published: August 2012

The Government of Iraq (GOI) has undertaken efforts to decommission and dismantle former nuclear facilities. The GOI has only preliminary information on some of the former nuclear facilities. This paper will highlight the challenges involved in conducting inspections of the outlying former nuclear facilities in Iraq and present a brief summary of the results of those inspections. The facilities discussed in this paper are located at various sites throughout Iraq, from locations close to Baghdad to those in the north and far western desert areas. Some of the facilities, such as those at the Al Tuwaitha Nuclear Research Center have been visited and characterized. Other facilities, including the following, have not been visited or thoroughly characterized. • Al Jesira, Uranium feed stock production facility; • Adaya, Burial location for contaminated equipment; • Djerf al Naddah, Spent fuel storage facility; • Rashdiya, Centrifuge development center; • Al Qa’im, Uranium (yellowcake) production facility. The visits were conducted to develop an inventory of the buildings/structures that need to be included in decommissioning/dismantlement efforts. The number of buildings, type of construction, size and general condition of the buildings were noted. In addition, attempts were made to determine contamination levels on surfaces, equipment, rubble, etc. This information will be used to support the Iraqi decommissioning and dismantlement project. Because the facilities are scattered throughout the country of Iraq, significant planning and coordination was required to ensure personnel security. Teams consisting of individuals from the Iraqi Ministry of Science and Technology (MoST) and Americans were under military escort when traveling to and visiting the sites. Because of the security issues, time on the ground at each site was limited. This paper will highlight the challenges involved in conducting the inspections of the outlying former nuclear facilities In Iraq and present a brief summary of the results of those inspections.
Determination of the Radionuclide Contamination on the Absheron Peninsula in Azerbaijan
Authors: Vandergraaf Tjalle T. , Mamedov Gudrat G. , Ramazanov Mahammadali A. , Naghiyev Jalal A. , Mehdiyeva Afat A. , Huseynov Nazim A.
Published: August 2012

Much of Azerbaijan’s Absheron Peninsula is contaminated by natural U-series and Th-series radionuclides, released in the production of oil and gas and, to a lesser extent, by anthropogenic radionuclides, including Sr-90 and Cs-137, from local industrial activities and trans-border transport. The region contains a large number of pipelines and artificial lagoons that have been used to retain excess groundwater and oil residues. In spite of the long history of the oil and gas industry, radioecological investigations have not been carried out until recently. The purpose of this project is to determine the extent of radionuclide contamination in the Absheron Peninsula using a combination of radiation field measurements and laboratory analyses of selected samples, focusing on ten routes in the vicinity of Baku. The routes were selected as most likely to have become contaminated over time. Soil samples, taken from surface and to a depth of 1 m, aqueous samples from surface waters and marshes, and aqueous and sediment lagoons that showed elevated dosimetry readings, were prepared for gamma spectrometric analysis. Control samples were taken from non-contaminated areas. Samples of air and surface waters were analyzed for Rn-220 and Rn-222. The data will then be used to assess the potential impact of the contamination on the local population. A total slightly 4000 dosimetric readings were taken during the course of this investigation. Of these, 1366 (34%) exceeded 5.4 μR/h. This level is two standard deviations above the mean of the least contaminated route, the 79-km Baku-Guba route. Along the routes Baku-Shemakha and Baku-Guba where no oil and gas activity had taken place, radiation levels of 5.1 ± 1.5 and 4.2 ± 0.6 μR/h, respectively, were obtained. The readings for the route Baku-Guba were then used as representing negligible contamination to which the readings of the other sites were compared. In contrast, along the routes Baku-Lokbatan and Baku-Surakhani, that have seen oil- and gas-related activity, radiation levels were sometimes two or three orders of magnitude higher. The most highly contaminated sites were those of two abandoned iodine recovery facilities along the route Baku-Surakhani, the Ramani and Surakhani sites where readings up to 1450 μR/h were obtained. The contamination is due mainly to uranium and thorium in the formation water associated with the oil and gas. Radon measurements did not exceed 20 Bq/m3.
Full System Decontamination (FSD) Prior to Decommissioning
Authors: Stiepani Christoph
Published: August 2012

Decontamination prior to decommissioning and dismantling is an internationally accepted approach. Not only does it provide for minimization of personnel dose exposure but also maximization of the material volume available for free release. Since easier dismantling techniques in lower dose areas can be applied, the licensing process is facilitated and the scheduling and budgeting effort is more reliable. The most internationally accepted approach for decontamination prior to decommissioning projects is the Full System Decontamination (FSD). FSD is defined as the chemical decontamination of the primary cooling circuit, in conjunction with the main auxiliary systems. AREVA has long-term experience with Full System Decontamination for return to service of operating nuclear power plants as well as for decommissioning after shutdown. Since 1976, AREVA has performed over 500 decontamination applications and from 1986 on, decontaminations prior to decommissioning projects which comprise virtually all nuclear power plant (NPP) designs and plant conditions: • NPP designs: HPWR, PWR, and BWR by AREVA, Westinghouse, ABB and GE; • Decontaminations performed shortly after final shutdown or several years later, and even after re-opening safe enclosure; • High alpha inventory and or low gamma/alpha ratio; • Main coolant chemistry (e.g. with and without Zn injection during operation). Fifteen decontaminations prior to decommissioning projects have been performed successfully to date. The lessons learned of each project were consequently implemented for the next project. AREVA NP has developed a fully comprehensive approach for decontamination based on the CORD® (Chemical Oxidation Reduction Decontamination) Family, applied using the in-house designed decontamination equipment AMDA® (Automatic Modular Decontamination Appliance). The Decontamination Concept for Decommissioning (DCD) will be outlined in this paper. Based on highlights of previous FSDs performed prior to decommissioning the AREVA concept for FSD and DCD will be outlined: • Application window; • Decontamination area; • Waste considerations; • Positive results for subsequent decommissioning and dismantling activities.
Mitigation of Sliding Motion of a Cask-Canister by Fluid-Structure Interaction in an Annular Region
Authors: Ito Tomohiro , Fujiwara Yoshihiro , Shintani Atsuhiko , Nakagawa Chihiro , Furuta Kazuhisa
Published: August 2012

The cask-canister system is a coaxial circular cylindrical structure in which several spent fuels are installed. This system is a free-standing structure thus, it is very important to reduce sliding motion for very large seismic excitations. In this study, we propose a mitigation method for sliding motion. Water is installed in an annular region between a cask and a canister. The equations of motion are derived taking fluid-structure interaction into consideration for nonlinear sliding motion analyses. Based on these equations, mitigation effects of sliding motions are studied analytically. Furthermore, a fundamental test model of a cask-canister system is fabricated and shaking table tests are conducted. From the analytical and test results, sliding motion mitigation effects are investigated.
Planning for Decommissioning of Nuclear Facilities: Some Views of Younger Stakeholders
Authors: Labor Bea , Lindskog Staffan
Published: August 2012

One of the prerequisite for nuclear power to be regarded as a sustainable energy source is that no burden be put on future generations. One important step in the financial and technical planning for decommissioning is therefore to investigate, describe and explain the opinions and values of especially the younger stakeholders. In this article, some results from current field studies in Poland and Slovakia are presented. The aggregated survey is based on 1444 personal interviews in four towns in Poland and one geographic area in Slovakia with a near 100% response rate. The main conclusions from this field study may be summarised as follows: • Sustainable energy sources are favoured (nuclear power may be included if waste is managed properly). • Nuclear power is seen as a potential future semi-sustainable energy source by nearly 1/4th of the respondents. • The values are based on safety and environmental aspects. • The polluter pays principle (extended polluter responsibility) is accepted. • There are doubts regarding the compliance with these principles in the implementation of the disposal of the nuclear residues. • Means and modes of communication with the younger stakeholders need further development. • No difference in views was found between the sexes. • Nearly one 1/5th of the respondents are open to accept reprocessing.
Standardizing of LLW/ILW Treatment for Russian Type Pressure Water Reactors
Authors: Bu¨ttner Klaus , Slametschka Rainer
Published: August 2012

According to the framework program of the Russian Federation released in 2010 it is planned to build up additional nuclear capacity of approximately 20 GWe. In addition, contracts were signed for several new-builds throughout the world. The related design efforts together with the restricted design capacities in Russia led Rosenergoatom (Client) and Atomstroyexport (Co-Client) to the set up of a so-called unified project “TOI”. “TOI” means translated “Technically Optimized reactor designed with Integrated databases”. NUKEM Technologies owned by Atomstroyexport was selected to develop the standardized technically optimized design for the treatment of the generated LLW and ILW. Special emphasis was made to reduce the waste volume for storage after treatment to below 50 m3 per year and unit. The selected treatment technologies compromise sorting, fragmentation, thermal treatment, compaction, evaporation, cementation etc. In 2010 NUKEM Technologies together with Atomenergoprojekt (Main Contractor) developed the technical concept for the treatment of liquid and solid LLW and ILW generated within the reactor as well as in the auxiliary buildings. In an iterative process the waste amounts generated and the respective treatment methods were discussed and agreed. The next step will be to further develop the design. The basic design (technical project) should be finalized until end of 2012.
Application of Coupled Thermo-Hydro-Mechanical-Chemical (THMC) Processes in Hydrothermal Systems to Processes Near a High-Level Nuclear Waste Repository
Authors: Peter Geoffrey J.
Published: August 2012

Modeling of coupled processes in the geology near a high-level nuclear waste repository is similar to the modeling of coupled Thermo-Hydro-Mechanical-Chemical (THMC) processes that occur in magma-hydrothermal systems. Former Professor Denis Norton and his colleagues at the Geoscience Department at University of Arizona studied magma-hydrothermal systems extensively. These hydro-thermal codes were verified by obtaining excellent matches between calculated δ18O–values and measured δ18O–values in three principal rock units: basalt, gabbro, and gneiss. This paper reviews the concept of transport theory used in the formulation of the conservation principle used to model the hydrothermal systems. In addition, the paper reviews conservation of mass, momentum, energy, and chemical component equations as applied to the multicomponent-multiphase systems related to hydrothermal systems and obtains parallels to reaction rates and radionuclide transport in the geology of a high level nuclear water repository. Further, this paper compares published results obtained by other researchers modeling coupled THMC process in the geology of high-level nuclear waste repositories.
Measurement of Plutonium Contamination Through Paint Using a Fidler Probe
Authors: Clark Robert , Burgess Pete , Croudace Ian
Published: August 2012

Alpha contamination detection usually relies on good surface conditions — unobstructed, smooth, clean and flat. However, in many circumstances, the conditions may be difficult, or it may be that there is a significant chance that contamination has been deliberately or accidentally painted over. Hence an alternative method of monitoring would be useful. In some areas of fuel manufacture and reprocessing, and in weapons manufacture, potential contamination is dominated by plutonium isotopes and by ingrown Am-241, derived from the beta decay of Pu-241. Plutonium is often thought of as a mainly an alpha emitter, together with a few weak betas detectable by direct means only with extreme difficulty. However, the alpha emitting isotopes of plutonium also emit significant L x-rays in the 11 to 20 keV energy range, as does Am-241, which also emits a 60 keV gamma with a 36% probability. These X-rays are unattenuated to any extent in air over a range of 1 metre. They also penetrate paint significantly, which makes them detectable by a suitable probe. The Fidler probe was designed as an efficient detector of these X and gamma radiations. The window is thin beryllium and the scintillator is thin sodium iodide. This leads to a very efficient detection of both the X-rays in question and the 60 keV gamma radiation while keeping the background as low as possible. The signal from such a detector can be processed in several ways — gross counting above a threshold, counting in regions of interest or full spectrometry. The advantages of the latter include the minimisation of background, easy background correction, the ability to use the recorded X-ray spectrum to correct the measured counts and the identification of the presence of other gamma emitters. Nuvia Ltd have been developing techniques to look at the spectrum from Fidler probes and derive a calibration factor, either from using the 60 keV gamma measurement only where the fingerprint is well known and stable, or by using samples of paint and small area sources to actually measure the transmission factor where the Am-241 fraction is variable. Using these techniques, it is possible to obtain a good estimate of the plutonium contamination level under paint over a range of conditions.
The Cige´o Industrial Geological Repository Project
Authors: Labalette Thibaud , Harman Alain , Dupuis Marie-Claude
Published: August 2012

The Planning Act of 28 June 2006 prescribed that a reversible repository in a deep geological formation be chosen as the reference solution for the long-term management of high-level and intermediate-level long-lived radioactive waste. It also entrusted the responsibility of further studies and investigations on the siting and design of the new repository upon the French Radioactive Waste Management Agency (Agence nationale pour la gestion des de´chets radioactifs – Andra), in order for the review of the creation-licence application to start in 2015 and, subject to its approval, the commissioning of the new repository in 2025. In late 2009, Andra submitted to the French government proposals concerning the implementation and the design of Cige´o (Centre industriel de stockage ge´ologique). A significant step of the project was completed with the delineation of an interest zone for the construction of the repositor’s underground facilities in 2010. This year, Andra has launched a new dialogue phase with local actors in order to clarify the implementation scenarios on the surface. The selected site will be validated after the public debate that is now scheduled for the first half of 2013. This debate will be organized by the National Public Debate Committee (Commission nationale du de´bat public). In parallel, the State is leading the preparation of an territorial development scheme, which will be presented during the public debate. The 2009 milestone also constitutes a new step in the progressive design process of the repository. After the 1998, 2001 and 2005 iterations, which focused mainly on the long-term safety of the repository, the Dossier 2009 highlighted its operational safety, with due account of the non-typical characteristics of an underground nuclear facility. It incorporates the first results of the repository-optimisation studies, which started in 2006 and will continue in the future. The reversibility options for the repository constitute proposals in terms of added flexibility in repository management and in package-recovery levels. They orient the design of the repository in order to promote those reversibility components. They contribute to the dialogue with stakeholders in the preparation of the public debate and of the future act on the reversibility conditions of the repository. The development of the repository shall be achieved over a long period, around the century. Hence, the designer will acquire additional knowledge at every new development of the project, notably during Phase 1, which he may reuse during the following phase, in order, for instance, to optimise the project. This process is part of the approach proposed by Andra in 2009 pursuant to the reversibility principle.
A Practical Approach to Proving Waste Metals Suitable for Consignment as Radiologically Exempt Materials
Authors: Carvel Iain , Gunn Richard D. , Orr Christopher H. , Strange Robin
Published: August 2012

Building 220 at Harwell was built by the Ministry of Works as a Radiochemical Research and Development facility in the latter part of the 1940’s. The facility has been operational since 1949 and has been extended several times, most notably the Plutonium Glove Box Wing in the 1950’s and the Remote Handling Wing in the 1980’s. Only the Remote Handling wing remains operational, processing Historic Waste which is being recovered from storage holes elsewhere on site. The remainder of the facility is undergoing progressive strip out and decommissioning. In the Plutonium Wing and associated areas the waste ‘fingerprint’ (nuclide vector) consists predominately of alpha emitting radionuclides. Decommissioning and Decontamination (D&D) operations often result in the production of large volumes of scrap metal waste with little or no radioactive contamination. Proving that the waste is clean can be costly and time consuming, as the shape and size of the metallic waste items often means that it is difficult or impossible to monitor all surfaces using conventional hand-held survey meters. This is a particular problem for alpha contamination measurement. Traditional radiological surveying techniques are very labour intensive and involve surveyors checking every surface using hand held instruments and smear sampling the hard to access areas. Even then 100% monitoring cannot be guaranteed. An alternative to traditional methods is the Long Range Alpha Detection (LRAD) technique which remotely detects and measures secondary ionization created in air by alpha particle interactions, allowing extremely low levels of alpha contamination to be measured. A survey system, IonSens®, using the LRAD technique, was developed by BNFL Instruments Ltd (now Babcock Nuclear) which allows rapid surveying of scrap metal for alpha contamination at very low levels. Two versions of this system exist but both essentially comprise a measurement chamber into which scrap metal is placed and sealed while a measurement is carried out. Research Sites Restoration Ltd (RSRL) has purchased an IonSens® system with a single 2m long measurement chamber. This instrument is optimised for clearance monitoring of pipework of up to 150mm diameter. Testing and use of support baskets has enabled the instrument to be used as a clearance monitor for a wide range of metallic waste including hand tools. This paper illustrates how the instrument, and a robust consignment procedure, has been used to justify the clearance of several tonnes of metal wastes for recycling rather than paying for disposal of this material as low level waste.