bluuebunny/crossref_metadata_2025 huggingface dataset:
Recent Progress in Low-Level Gamma Imaging
Authors: Mahe´ C. , Lamadie F. , Girones Ph. , Le Goaller C.
Published: June 2009

The CEA’s Aladin gamma imaging system has been operated successfully for several years in nuclear plants and during decommissioning projects with additional tools such as gamma spectrometry detectors and dose rate probes. The radiological information supplied by these devices is becoming increasingly useful for establishing robust and optimized decommissioning scenarios. Recent technical improvements allow this gamma imaging system to be operated in low-level applications and with shorter acquisition times suitable for decommissioning projects. The compact portable system can be used in places inaccessible to operators. It is quick and easy to implement, notably for onsite component characterization. Feasibility trials and in situ measurements were recently carried out under low-level conditions, mainly on waste packages and glove boxes for decommissioning projects. This paper describes recent low-level in situ applications. These characterization campaigns mainly concerned gamma emitters with γ energy < 700 keV. In many cases, the localization of hot spots by gamma camera was confirmed by additional measurements such as dose rate mapping and gamma spectrometry measurements. These complementary techniques associated with advanced calculation codes (MCNP, Mercure 6.2, Visiplan and Siren) offer a mobile and compact tool for specific assessment of waste packages and glove boxes.
Acoustic Emission Monitoring of Cement-Based Structures Immobilising Radioactive Waste
Authors: Spasova L. M. , Ojovan M. I. , Hayes M. , Godfrey H.
Published: June 2009

The long term performance of cementitious structures immobilising radioactive waste can be affected by physical and chemical processes within the encapsulating materials such as formation of new phases (e.g., vaterite, brucite), degradation of cement phases (e.g., CSH gel, portlandite), degradation of some waste components (e.g., organics), corrosion of metallic constituents (aluminium, magnesium), gas emission, further hydration etc. The corrosion of metals in the high pH cementitious environment is of especial concern as it can potentially cause wasteform cracking. One of the perspective non-destructive methods used to monitor and assess the mechanical properties of materials and structures is based on an acoustic emission (AE) technique. In this study an AE non-destructive technique was used to evaluate the mechanical performance of cementitious structures with encapsulated metallic waste such as aluminium. AE signals generated as a result of aluminium corrosion in a small-size blast furnace slag (BFS)/ordinary Portland cement (OPC) sample were detected, recorded and analysed. A procedure for AE data analysis including conventional parameter-based AE approach and signal-based analysis was applied and demonstrated to provide information on the aluminium corrosion process and its impact on the mechanical performance of the encapsulating cement matrix.
Site Descriptive Modeling as a Part of Site Characterization in Sweden: Concluding the Surface Based Investigations
Authors: Andersson Johan , Skagius Kristina , Winberg Anders , Stro¨m Anders , Lindborg Tobias
Published: June 2009

The Swedish Nuclear Fuel and Waste Management Co., SKB, is currently finalizing its surface based site investigations for the final repository for spent nuclear fuel in the municipalities of O¨sthammar (the Forsmark area) and Oskarshamn (the Simpevar/Laxemar area). The investigation data are assessed into a Site Descriptive Model, constituting a synthesis of geology, rock mechanics, thermal properties, hydrogeology, hydrogeochemistry, transport properties and a surface system description. Site data constitute a wide range of different measurement results. These data both need to be checked for consistency and to be interpreted into a format more amenable for three-dimensional modeling. The three-dimensional modeling (i.e. estimating the distribution of parameter values in space) is made in a sequence where the geometrical framework is taken from the geological models and in turn used by the rock mechanics, thermal and hydrogeological modeling. These disciplines in turn are partly interrelated, and also provide feedback to the geological modeling, especially if the geological description appears unreasonable when assessed together with the other data. Procedures for assessing the uncertainties and the confidence in the modeling have been developed during the course of the site modeling. These assessments also provide key input to the completion of the site investigation program.
Cost Effective ER Data Acquisition Using a Dynamic Characterization Strategy
Authors: Skinner Kenneth M. , Salpas Peter A.
Published: June 2009

The important first step in remediating contaminated sites is completing characterization. The process for characterization of natural environmental media (i.e., soil, sediment, surface water, groundwater) involves three basic steps: (1) develop a plan, (2) implement the plan by collecting information necessary to define the nature and extent of contaminants in the natural media, and (3) integrate, interpret and report the results. Because of budgetary constraints, these three steps are typically applied linearly with the expectation that by the end of one application of the process the site will be characterized with sufficient resolution to make decisions about remedial actions. Our experience over the past 13 years at a complex site in Tennessee has shown that this linear approach to characterization does not produce the desired resolution. Because characterization is typically a process of defining unknowns the inflexible nature of the linear approach makes it impractical to react as the conceptual understanding of site contaminants changes in response to the acquisition of new data. An alternative, flexible approach to characterization has been developed based on lessons learned. Over the past 3 years the flexible approach has cost-effectively produced the information needed for decision making.
Oil Immobilization Program at Sellafield: An Innovative Approach
Authors: Cassidy Helen
Published: June 2009

Non-standard wastes — those defined as being both hazardous waste under the United Kingdom Hazardous Waste Regulations 2005 [1] and radioactive under the Radioactive Substances Act 1993 [2] — pose particular, unique challenges for radioactive waste management organizations [3]. Treatment and disposal routes for such wastes are limited, in some cases non existent, and generally not cost effective. A non-standard waste of particular concern in the United Kingdom, and indeed on the Sellafield site, is that of radiologically contaminated waste oil. The optioning process for treatment of bulk contaminated waste oil on the Sellafield site has assessed a range of options including incineration, chemical decontamination, physical decontamination and immobilization. Immobilization has proved to be a potentially useful option for oil wastestreams that fail to meet waste acceptance criteria for incineration facilities. Experimental development work has been undertaken at Sellafield during 2006 to test the suitability of an innovative technology for the solidification of waste oil with a cross section of wastestreams from the site. These trials have demonstrated that this polymer system is able to successfully immobilize a range of aged, chemically and physically diverse contaminated oil wastestreams and thus provide a potential solution to the disposal problem posed by this wastestream.
Radiological Characterization of a Vitrification Facility for Decommissioning
Authors: Asou M. , Le Goaller C. , Martin F.
Published: June 2009

Cleanup operations in the Marcoule Vitrification Facility (AVM) will start in 2007 [ref 1]. This plant includes 20 highly irradiating storage tanks for high-level liquid waste before vitrification. The objective of the cleanup phase is to significantly decrease the amount of highly radioactive waste resulting from dismantling. A comprehensive radiological survey of the plant was initiated in 2000. Most of the tanks were characterized using advanced technologies: gamma imaging, CdZnTe gamma spectroscopy, dose rate measurements and 3D calculations codes. At the same time, inspections were conducted to develop 3D geometrical models of the tanks. The techniques used and the main results obtained are described as well as lessons learned from these operations. The rinsing program was defined in 2006. Decontamination operations are expected to begin in 2007, and radiological surveys will be followed up to monitor the efficiency of the decontamination process. Specific rinsing of all tanks and equipment will be carried out from 2007 to 2009. Concentrated liquid solutions will be vitrified between 2008 and 2010; the decommissioning of AVM will be delayed until the end of 2010. This strategy aims at producing less than 5% “B” type (long-lived intermediate-level) waste from the decommissioning operations, as well as reducing the dose rate and risks by simplified remote dismantling. The paper reviews the main options selected for decontamination, as well as the radiological characterization strategy. Some cost-related aspects will also be analyzed.
HEPA Filter Performance Under Adverse Conditions
Authors: Parsons Michael , Hogancamp Kristina , Alderman Steven , Waggoner Charles
Published: June 2009

This study involved challenging nuclear grade high-efficiency particulate air (HEPA) filters under a variety of conditions that can arise in Department of Energy (DOE) applications such as: low or high RH, controlled and uncontrolled challenge, and filters with physically damaged media or seals (i.e., leaks). Reported findings correlate filter function as measured by traditional differential pressure techniques in comparison with simultaneous instrumental determination of up and down stream PM concentrations. Additionally, emission rates and failure signatures will be discussed for filters that have either failed or exceeded their usable lifetime. Significant findings from this effort include the use of thermocouples up and down stream of the filter housing to detect the presence of moisture. Also demonstrated in the moisture challenge series of tests is the effect of repeated wetting of the filter. This produces a phenomenon referred to as transient failure before the tensile strength of the media weakens to the point of physical failure. An evaluation of the effect of particle size distribution of the challenge aerosol on loading capacity of filters is also included. Results for soot and two size distributions of KCl are reported. Loading capacities for filters ranged from approximately 70 g of soot to nearly 900 g for the larger particle size distribution of KCl.
The Environmental Constraint Needs for Design Improvements to the Saligny I/LLW-Repository Near Cernavoda NPP
Authors: Barariu Gheorghe
Published: June 2009

The paper presents the new perspectives on the development of the L/ILW Final Repository Project which will be built near Cernavoda NPP. The Repository is designed to satisfy the main performance objectives in accordance to IAEA recommendation. Starting in October 1996, Romania became a country with an operating nuclear power plant. Reactor 2 reached the criticality on May 6th 2007 and it will be put in commercial operation in September 2007. The Ministry of Economy and Finance has decided to proceed with the commissioning of Units 3 and 4 of Cernavoda NPP till 2014. The Strategy for radioactive waste management was elaborated by National Agency for Radioactive Waste (ANDRAD), the jurisdictional authority for definitive disposal and the coordination of nuclear spent fuel and radioactive waste management (Order 844/2004) with attributions established by Governmental Decision (GO) 31/2006. The Strategy specifies the commissioning of the Saligny L/IL Radwaste Repository near Cernavoda NPP in 2014. When designing the L/IL Radwaste Repository, the following prerequisites have been taken into account: 1) Cernavoda NPP will be equipped with 4 CANDU 6 units. 2) National Legislation in radwaste management will be reviewed and/or completed to harmonize with UE standards 3) The selected site is now in process of confirmation after a comprehensive set of interdisciplinary investigations.
Long-Term Storage Facility for Reactor Compartments in Sayda Bay: German Support for Utilization of Nuclear Submarines in Russia
Authors: Wolff Dietmar , Vo¨lzke Holger , Weber Wolfgang , Noack Volker , Ba¨uerle Gu¨nther
Published: June 2009

The German-Russian project that is part of the G8 initiative on Global Partnership Against the Spread of Weapons and Materials of Mass Destruction focuses on the speedy construction of a land-based interim storage facility for nuclear submarine reactor compartments at Sayda Bay near Murmansk. This project includes the required infrastructure facilities for long-term storage of about 150 reactor compartments for a period of about 70 years. The interim storage facility is a precondition for effective activities of decommissioning and dismantlement of almost all nuclear-powered submarines of the Russian Northern Fleet. The project also includes the establishment of a computer-assisted waste monitoring system. In addition, the project involves clearing Sayda Bay of other shipwrecks of the Russian navy. On the German side the project is carried out by the Energiewerke Nord GmbH (EWN) on behalf of the Federal Ministry of Economics and Labour (BMWi). On the Russian side the Kurchatov Institute holds the project management of the long-term interim storage facility in Sayda Bay, whilst the Nerpa Shipyard, which is about 25 km away from the storage facility, is dismantling the submarines and preparing the reactor compartments for long-term interim storage. The technical monitoring of the German part of this project, being implemented by BMWi, is the responsibility of the Federal Institute for Materials Research and Testing (BAM). This paper gives an overview of the German-Russian project and a brief description of solutions for nuclear submarine disposal in other countries. At Nerpa shipyard, being refurbished with logistic and technical support from Germany, the reactor compartments are sealed by welding, provided with biological shielding, subjected to surface treatment and conservation measures. Using floating docks, a tugboat tows the reactor compartments from Nerpa shipyard to the interim storage facility at Sayda Bay where they will be left on the on-shore concrete storage space to allow the radioactivity to decay. For transport of reactor compartments at the shipyard, at the dock and at the storage facility, hydraulic keel blocks, developed and supplied by German subcontractors, are used. In July 2006 the first stage of the reactor compartment storage facility was commissioned and the first seven reactor compartments have been delivered from Nerpa shipyard. Following transports of reactor compartments to the storage facility are expected in 2007.
Estimation of Clearance Potential Index and Hazard Factors of CANDU Fuel Bundle and Its Validation Based on the Measurements of Radioisotopes Inventories From Pickering Reactor Fuel
Authors: Pavelescu Alexandru Octavian , Cepraga Dan Gabriel , Voukelatou Konstantina , Tinti Renato
Published: June 2009

This paper is related to the clearance potential levels, ingestion and inhalation hazard factors of the spent nuclear fuel and radioactive wastes. This study required a complex activity that consisted of more steps such as: the acquisition, setting up, validation and application of procedures, codes and libraries. The paper reflects the validation stage of this study. Its objective was to compare the measured inventories of selected actinide and fission products radionuclides in an element from the Pickering CANDU reactor with the inventories predicted using a recent version of the SCALE 5\ORIGEN-ARP code coupled with the time dependent cross sections library for the CANDU 28 reactor (produced by the sequence SCALE4.4a\SAS2H and SCALE4.4a\ORIGEN-S). In this way, the procedures, the codes and the libraries for the characterization of radioactive material in terms of radioactive inventories, clearance, and biological hazard factors could be qualified and validated, in support of the safety management of the radioactive wastes.
Novel Technology for Hydrothermal Treatment of NPP Evaporator Concentrates
Authors: Avramenko Valentin , Dobrzhansky Vitaly , Marinin Dmitry , Sergienko Valentin , Shmatko Sergey
Published: June 2009

A novel technology was developed for treatment of evaporator concentrates produced as a result of operation of evaporation devices comprising the main component of special water purification systems of nuclear power plants (NPP). The developed technology includes a hydrothermal (T = 250–300°C and P = 80–120 bar) processing of evaporator concentrates in oxidation medium in order to destruct stable organic complexes of cobalt radionuclides and remove these radionuclides by oxide materials formed during such a processing. The cesium radionuclides contained in evaporator concentrates are removed by a conventional method — through application of one of the developed composite sorbents with ferrocyanides of transition metals used as active agents. Extensive laboratory studies of the processes occurring in evaporator concentrates under hydrothermal conditions were performed. It was shown that hydrothermal oxidation of evaporator concentrates has a number of advantages as compared to traditional oxidation methods (ozonation, photocatalytic, electrochemical and plasma oxidation). A laboratory installation was built for the flow-type hydrothermal oxidation of NPP evaporator concentrates. The obtained experimental results showed good prospects for the developed method application. On the basis of the results obtained, a pilot installation of productivity up to 15 l/hour was developed and built in order to work out the technology of evaporator concentrates hydrothermal treatment. The pilot tests of the hydrothermal technology for evaporator concentrates hydrothermal treatment were performed for 6 months in 2006 at the 1st reactor unit of the Novovoronezhskaya NPP (Voronezh Region, Russia). Optimal technological regimes were determined, and estimations of the economic soundness of the technology were made. The advantages of the presented technology in terms of management of concentrated liquid radioactive wastes (LRW) at nuclear cycle facilities, as compared to other methods applicable for this type of LRW, were demonstrated. Application of the hydrothermal technology in the system of NPP LRW management enables one to reduce substantially the volume of solid radioactive waste sent for final disposal.
Local Negotiation on Compensation: Siting of the Spent Nuclear Fuel Repository in Finland
Authors: Kojo Matti
Published: June 2009

The aim of the paper is to analyse the local negotiation process between the Municipality of Eurajoki and the nuclear power company Teollisuuden Voima (TVO) and the nuclear waste management company Posiva Oy. The aim of the negotiations was to find an acceptable form of compensation for siting a spent nuclear fuel repository in Olkiluoto, Finland. The paper includes background information on the siting process in Finland, the local political setting in the Municipality of Eurajoki and a description of the negotiation process. The analysis of the negotiations on compensation is important for better understanding the progress of the Finnish siting process. The paper describes the picture of the contest to host the spent nuclear fuel repository. It also provides more information on the relationship between the Municipality of Eurajoki and the power company TVO. The negotiations on compensation and the roles of various players in the negotiations have not been studied in detail because the minutes of the Vuojoki liaison group were not available before the decision of the Supreme Administrative Court in May 2006.
The Effect of Particle-Particle Interaction Forces on the Flow Properties of Silica Slurries
Authors: Harbottle David , Rhodes Dominic , Fairweather Michael , Biggs Simon
Published: June 2009

Preliminary work has been completed to investigate the effect of particle-particle interaction forces on the flow properties of silica slurries. Classically Hydrotransport studies have focused on the flow of coarse granular material in Newtonian fluids. However, with current economical and environmental pressures, the need to increase solid loadings in pipe flow has lead to studies that examine non-Newtonian fluid dynamics. The flow characteristics of non-Newtonian slurries can be greatly influenced through controlling the solution chemistry. Here we present data on an “ideal” slurry where the particle size and shape is controlled together with the solution chemistry. We have investigated the effect of adsorbed cations on the stability of a suspension, the packing nature of a sediment and the frictional forces to be overcome during reslurrying. A significant change in the criteria assessed was observed as the electrolyte concentration was increased from 0.1mM to 1M. In relation to industrial processes, such delicate control of the slurry chemistry can greatly influence the optimum operating conditions of non-Newtonian pipe flows.
European Network on the Determination of Site End Points for Radiologically Contaminated Land
Authors: Booth Peter , Lennon Chris
Published: June 2009

Nexia Solutions are currently running a small European network entitled “European Network on the Determination of Site End Points for Radiologically Contaminated Land (ENDSEP)”. Other network members include NRG (Netherlands), UKAEA (UK), CEA (France), SOGIN (Italy), Wismut (Germany), Saxon State Agency of Environment and Geology (Germany). The network is focused on the technical and socio-economical issues associated with the determination of end points for sites potentially, or actually, impacted by radiological contamination. Such issues will cover: • Those associated with the run up to establishing a site end point; • Those associated with verifying that the end points have been met; and • Those associated with post closure. The network’s current high level objectives can be summarized as follows: • Share experience and best practice in the key issues running up to determining site end points; • Gain a better understanding of the potential effects of recent and forthcoming EU legislation; • Assess consistency between approaches; • Highlight potential gaps within the remit of site end point determination and management; and • Consider the formulation of research projects with a view to sharing time and expense. The programme of work revolves around the following key tasks: • Share information, experience and existing good practice. • Look to determine sustainable approaches to contaminated land site end point management. • Through site visits, gain first hand experience of determining an appropriate end point strategy, and identifying and resolving end point issues. • Highlight the key data gaps and consider the development of programmes to either close out these gaps or to build confidence in the approaches taken. • Production of position papers on each technical area highlighting how different countries approach/resolve a specific problem.
Landscape Modeling for Dose Calculations in the Safety Assessment of a Repository for Spent Nuclear Fuel
Authors: Lindborg Tobias , Kautsky Ulrik , Brydsten Lars
Published: June 2009

The Swedish Nuclear Fuel and Waste Management Co., (SKB), pursues site investigations for the final repository for spent nuclear fuel at two sites in the south eastern part of Sweden, the Forsmark- and the Laxemar site (figure 1). Data from the two site investigations are used to build site descriptive models of the areas. These models describe the bedrock and surface system properties important for designing the repository, the environmental impact assessment, and the long-term safety, i.e. up to 100,000 years, in a safety assessment. In this paper we discuss the methodology, and the interim results for, the landscape model, used in the safety assessment to populate the Forsmark site in the numerical dose models. The landscape model is built upon ecosystem types, e.g. a lake or a mire, (Biosphere Objects) that are connected in the landscape via surface hydrology. Each of the objects have a unique set of properties derived from the site description. The objects are identified by flow transport modeling, giving discharge points at the surface for all possible flow paths from the hypothetical repository in the bedrock. The landscape development is followed through time by using long-term processes e.g. shoreline displacement and sedimentation. The final landscape model consists of a number of maps for each chosen time period and a table of properties that describe the individual objects which constitutes the landscape. The results show a landscape that change over time during 20,000 years. The time period used in the model equals the present interglacial and can be used as an analogue for a future interglacial. Historically, the model area was covered by sea, and then gradually changes into a coastal area and, in the future, into a terrestrial inland landscape. Different ecosystem types are present during the landscape development, e.g. sea, lakes, agricultural areas, forest and wetlands (mire). The biosphere objects may switch from one ecosystem type to another during the modeled time period, from sea to lake, and from lake to mire and finally, some objects are transformed into agricultural area due to favorable farming characteristics.
Hydrogeological Characterization on Surface-Based Investigation Phase in the Mizunami Underground Research Laboratory Project, in Japan
Authors: Saegusa Hiromitsu , Onoe Hironori , Takeuchi Shinji , Takeuchi Ryuji , Ohyama Takuya
Published: June 2009

The Mizunami Underground Research Laboratory (MIU) project is being carried out by Japan Atomic Energy Agency in the Cretaceous Toki granite in the Tono area, central Japan. The MIU project is a purpose-built generic underground research laboratory project that is planned for a broad scientific study of the deep geological environment as a basis of research and development for geological disposal of nuclear wastes. One of the main goals of the MIU project is to establish comprehensive techniques for investigation, analysis, and assessment of the deep geological environment. The MIU project has three overlapping phases: Surface-based Investigation (Phase I), Construction (Phase II) and Operation (Phase III). Hydrogeological investigations using a stepwise process in Phase I have been carried out in order to obtain information on important properties such as, location of water conducting features, hydraulic conductivity and so on. Hydrogeological modeling and groundwater flow simulations in Phase I have been carried out in order to synthesize these investigation results, to evaluate the uncertainty of the hydrogeological model and to identify the main issues for further investigations. Using the stepwise hydrogeological characterization approach and combining the investigation with modeling and simulation, understanding of the hydrogeological environment has been progressively improved.
Evaluation of Aluminum Pit Corrosion in Oak Ridge Research Reactor Pool by Quantitative Imaging and Thermodynamic Modeling
Authors: Jang Ping-Rey , Arunkumar Rangaswami , Lindner Jeffrey S. , Long Zhiling , Mott Melissa A. , Okhuysen Walter P. , Su Yi , Monts David L. , Kirk Paula G. , Ettien John
Published: June 2009

The Oak Ridge Research Reactor (ORRR) was operated as an isotope production and irradiation facility from March 1958 until March 1987. The US Department of Energy permanently shut down and removed the fuel from the ORRR in 1987. The water level must be maintained in the ORRR pool as shielding for radioactive components still located in the pool. The U.S. Department of Energy’s Office of Environmental Management (DOE EM) needs to decontaminate and demolish the ORRR as part of the Oak Ridge cleanup program. In February 2004, increased pit corrosion was noted in the pool’s 6-mm (1/4")-thick aluminum liner in the section nearest where the radioactive components are stored. If pit corrosion has significantly penetrated the aluminum liner, then DOE EM must accelerate its decontaminating and decommissioning (D&D) efforts or look for alternatives for shielding the irradiated components. The goal of Mississippi State University’s Institute for Clean Energy Technology (ICET) was to provide a determination of the extent and depth of corrosion and to conduct thermodynamic modeling to determine how further corrosion can be inhibited. Results from the work will facilitate ORNL in making reliable disposition decisions. ICET’s inspection approach was to quantitatively estimate the amount of corrosion by using Fouriertransform profilometry (FTP). FTP is a non-contact 3-D shape measurement technique. By projecting a fringe pattern onto a target surface and observing its deformation due to surface irregularities from a different view angle, the system is capable of determining the height (depth) distribution of the target surface, thus reproducing the profile of the target accurately. ICET has previously demonstrated that its FTP system can quantitatively estimate the volume and depth of removed and residual material to high accuracy. The results of our successful initial deployment of a submergible FTP system into the ORRR pool are reported here as are initial thermodynamic modeling results.
Bioavailability and Stability of Mercury Sulfide in Armuchee (USA) Soil
Authors: Han Fengxiang , Shiyab Safwan , Su Yi , Monts David L. , Waggoner Charles A. , Matta Frank B.
Published: June 2009

Because of the adverse effects of elemental mercury and mercury compounds upon human health, the U.S. Department of Energy (DOE) is engaged in an on-going effort to monitor and remediate mercury-contaminated DOE sites. In order to more cost effectively implement those extensive remediation efforts, it is necessary to obtain an improved understanding of the role that mercury and mercury compounds play in the ecosystem. We have conducted pilot scale experiments to study the bioavailability of mercury sulfide in an Armuchee (eastern US) soil. The effects of plants and incubation time on chemical stability and bioavailability of HgS under simulated conditions of the ecosystem have been examined, as has the dynamics of the dissolution of mercury sulfide by various extractants. The results show that mercury sulfide in contaminated Armuchee soil was still to some extent bioavailable to plants. After planting, soil mercury sulfide is more easily dissolved by both 4 M and 12 M nitric acid than pure mercury sulfide reagent. Dissolution kinetics of soil mercury sulfide and pure chemical reagent by nitric acid are different. Mercury release by EDTA from HgS-contaminated soil increased with time of reaction and soil mercury level. Chelating chemicals increase the solubility and bioavailability of mercury in HgS-contaminated soil.
Inductive Cold Crucible Melting and Characterization of Uranium and Thorium Bearing Murataite Ceramics
Authors: Knyazev O. A. , Ptashkin A. G. , Stefanovsky S. V. , Zen’kovskaya M. A. , Yudintsev S. V. , Nikonov B. S. , Stefanovsky O. I.
Published: June 2009

Murataite-based ceramics containing 10 wt.% either UO2 or ThO2 were produced by inductive cold crucible melting (ICCM) at operating frequencies of 1.76 MHz and 5.28 MHz and examined using X-ray diffraction and scanning and transmission electron microscopy. High specific productivity was achieved. Average melting ratio was 8.3 kW×h/kg. We identified three distinct murataite polytypes in the U-bearing ceramics: a five- (5C), eight- (8C), and three-fold (3C) fluorite unit cell composing respectively the core, intermediate and rim zones of the grains. In contrast, in the Th-bearing ceramic one of the polytypes with eight-fold (8C) fluorite unit cell was found to be prevailing over two others (5C and 3C). Computer simulation of the major reflection due to the murataite phase in the Th-bearing sample also exhibits superposition of peaks due to three distinct polytypes but one of them (8C) is predominant. The core zone of the murataite in the U-bearing specimens is characterized by UO2 concentrations as high as 12.1 wt%, which successively diminishes in concentration through the intermediate zone to the rim, the latter of which contains 5.2 wt% UO2. Thorium distribution within the murataite crystals is more uniform. The other phases found in the ceramics are crichtonite, rutile and traces of perovskite, Fe/Mn titanate ilmenite/pyrophanite, zirconolite, and vitreous phase. The difference in phase composition and actinide partitioning in the ceramics is influenced by synthesis conditions. Thus, application of large-scale cold crucibles is a prospective route for the development of industrial-scale process and technology for ceramization of actinide-bearing HLW. The advantage of this method is production of zoned crystals with the highest concentrations of actinides and rare earth elements in the core, effectively isolating these elements from potential leach solutions.
A National Perspective: Establishing and Implementing a Characterization Program in the U.S. for Remote-Handled Transuranic Waste
Authors: Nelson Roger , Harris Alton D.
Published: June 2009

The U.S. Department of Energy (DOE) is responsible for waste management from nuclear weapons production and operates the Waste Isolation Pilot Plant (WIPP) for permanent disposal of defense-generated transuranic waste (TRU), as authorized by Congress in 1979. Radioactive waste in the U.S. has historically been managed in one of two ways depending on its penetrating radiation dose rate. Waste with surface dose rates above 200 millirem/hour (0.002 sievert/hour) and waste that has been managed remotely (remote-handled). In 1992, Congress passed the WIPP Land Withdrawal Act, which created the regulatory framework under which DOE was to operate the facility, and authorized disposal of waste up to 1,000 rems/hour (10 Sievert/hour). Subsequently, DOE submitted applications to the Environmental Protection Agency (EPA), at the Federal level, for certification to operate WIPP, and to the New Mexico Environment Department (NMED), at the State level, for a hazardous waste permit. Both applications described the characterization methods that DOE proposed to use to ensure only compliant waste was shipped to WIPP. No distinction was employed in these methods concerning the surface dose rate from the waste. During the applications review, both regulatory agencies came to the conclusion in their approval that DOE had not demonstrated that remote-handled transuranic (RH-TRU) waste could be adequately characterized. Therefore, WIPP was only granted approval to begin waste disposal operations of waste with surface dose rates less than 200 millirem/hour (0.002 sievert/hour) — or contact-handled transuranic (CH-TRU) waste. Emplacement of CH-TRU waste in WIPP began March 26, 1999. However, WIPP was designed for disposal of both CH- and RH-TRU waste, with the RH-TRU waste in canisters emplaced in the walls of the underground disposal rooms and CH-TRU waste in containers in the associated open drifts. Therefore, as disposal rooms filled with CH-TRU waste, the space along the walls for RH-TRU waste disposal was lost. This made removal of the regulatory prohibition on RH-TRU waste a very high priority, and DOE immediately began an iterative process to change the two regulatory bases for RH-TRU waste disposal. These changes focused on how DOE could rely on CH-TRU characterization methods for adequate characterization of RH-TRU waste. On January 23, 2007, the first shipment of RH-TRU waste was finally received at WIPP. The revised EPA certification and NMED permit now both consider all waste characterization methods to be equally effective when applied to either CH- or RH-TRU waste, as DOE maintained in the original applications over 10 years ago.
Final Cleanup of Buildings Within in Legacy French Research Facilities: Strategy, Tools and Lessons Learned
Authors: Le Goaller C. , Doutreluingne C. , Berton M. A. , Doucet O.
Published: June 2009

This paper describes the methodology followed by the French Atomic Energy Commission (CEA) to decommission the buildings of former research facilities for demolition or possible reuse. It is a well known fact that the French nuclear safety authority has decided not to define any general release level for the decommissioning of nuclear facilities, thus effectively prohibiting radiological measurement-driven decommissioning. The decommissioning procedure therefore requires an intensive in-depth examination of each nuclear plant. This requires a good knowledge of the past history of the plant, and should be initiated as early as possible. The paper first describes the regulatory framework recently unveiled by the French Safety Authority, then, reviews its application to ongoing decommissioning projects. The cornerstone of the strategy is the definition of waste zoning in the buildings to segregate areas producing conventional waste from those generating nuclear waste. After dismantling, suitable measurements are carried out to confirm the conventional state of the remaining walls. This requires low-level measurement methods providing a suitable detection limit within an acceptable measuring time. Although this generally involves particle counting and in-situ low level gamma spectrometry, the paper focuses on γ spectrometry. Finally, the lessons learned from ongoing projects are discussed.
Towards Solving the Problems of the Solid Radioactive Wastes From Decommissioned Soviet-Designed Nuclear Submarines
Authors: Borgaas Bjo̸rn , Amundsen Ingar , Reistadt Ole
Published: June 2009

This paper describes the cooperation and inspection arrangements for verifying the dismantlement of non-strategic Russian submarines, including waste handling, in accordance with the provisions given in the bilateral agreement between the Governments of Russia and Norway and the Norwegian Plan of Action for Nuclear Safety. The main concerns during the project implementation are nuclear safety, the individual safety of workers and the safe and secure handling of radioactive and toxic wastes. Based on data provided by the shipyard on individual exposure, one dismantlement project results in average in an effective dose for shipyard workers between 1,1 and 1,9 mSv, depending on the scope of the work for each worker. The main conclusion is that the present handling of solid radioactive waste is not optimal as such waste now is being put into the reactor compartments without the possibility for adequate control, including the retrieval for repacking or transfer to adequate storage facilities, after the reactor compartments have been placed in the long term storage facility in Saida Bay. Today there is no repository in Northwest-Russia for defence-related radioactive waste.
Dismantling of Alpha Contaminated Obsolete Installations and Glove Boxes on the IRMM Site in Geel (Belgium)
Authors: Cretskens Pieter , Melis Gustaaf , Lenie Koen
Published: June 2009

At the Institute for Reference Materials and Measurements (European commission, Joint Research Centre, IRMM) a dismantling campaign of obsolete installations and glove boxes has been carried out in 2005. There were various reasons for their removal. Some large installations did not meet modern safety standards, other installations were worn out and expected to cause a radioactive contamination risk in the future. The main goal was to create as less waste as possible by extensive contamination checks and by decontamination if necessary. For the glove boxes, decontamination was not possible. A controlled area was set up around the installation to be dismantled in order to prevent spreading of contamination from dust and dirt. This was only possible for the “minor” contaminated installations. The dismantling campaign of the glove boxes was carried out by using tents of two types depending the contamination inside the glove boxes. The most common glove boxes were dismantled in a tent constructed with hard surfaced polycarbonate plates (ventilated cell). For glove boxes with higher contamination, the same principle was used but with a second “glove box tent” inside (ventilated glove tent). The purpose of this project was to learn from the experience of this campaign which gave the ability to make estimates of future radioactive waste or classic waste that could be expected from dismantled installations.
Making the Optimal Decsion in Selecting Protective Clothing
Authors: Price J. Mark
Published: June 2009

Protective Clothing plays a major role in the decommissioning and operation of nuclear facilities. Literally thousands of employee dress-outs occur over the life of a decommissioning project and during outages at operational plants. In order to make the optimal decision on which type of protective clothing is best suited for the decommissioning or maintenance and repair work on radioactive systems, a number of interrelating factors must be considered, including: – Protection; – Personnel Contamination; – Cost; – Radwaste; – Comfort; – Convenience; – Logistics/Rad Material Considerations; – Reject Rate of Laundered Clothing; – Durability; – Security; – Personnel Safety including Heat Stress; – Disposition of Gloves and Booties. In addition, over the last several years there has been a trend of nuclear power plants either running trials or switching to Single Use Protective Clothing (SUPC) from traditional protective clothing. In some cases, after trial usage of SUPC, plants have chosen not to switch. In other cases after switching to SUPC for a period of time, some plants have chosen to switch back to laundering. Based on these observations, this paper reviews the “real” drivers, issues, and interrelating factors regarding the selection and use of protective clothing throughout the nuclear industry.
Study for Radionuclide Transfer Ratio of Aerosols Generated During Heat Cutting
Authors: Iguchi Yukihiro , Baba Tsutomu , Kawakami Hiroto , Kitahara Takashi , Watanabe Atsushi , Kodama Mitsuhiro
Published: June 2009

The metallic elements with a low melting point and high vapor pressure seemed to transfer in aerosols selectively at dismantling reactor internals using heat cutting. Therefore, the arc melting tests of neutron irradiated zirconium alloy were conducted to investigate the radionuclide transfer behavior of aerosols generated during the heat cutting of activated metals. The arc melting test was conducted using a tungsten inert gas welding machine in an inert gas or air atmosphere. The radioactive aerosols were collected by filter and charcoal filter. The test sample was obtained from Zry-2 fuel cladding irradiated in a Japanese boiling water reactor for five fuel cycles. The activity analysis, chemical composition measurement and scanning electron microscope observation of aerosols were carried out. Some radionuclides were enriched in the aerosols generated in an inert gas atmosphere and the radionuclide transfer ratio did not change remarkably by the presence of air. The transfer ratio of Sb-125 was almost the same as that of Co-60. It was expected that Sb-125 was enriched from other elements since Sb is an element with a low melting point and high vapor pressure compared with the base metal (Zr). In the viewpoint of the environmental impact assessment, it became clear that the influence if Sb-125 is comparable to Co-60. The transfer ratio of Mn-54 was one order higher compared with other radionuclides. The results were discussed on the basis of thermal properties and oxide formation energy of the metallic elements.
Hydrogeological Conceptual Model Determined From Baseline and Construction Phase Groundwater Pressure and Surface Tiltmeter Data at the Mizunami Underground Research Laboratory, Japan
Authors: Takeuchi Shinji , Takeuchi Ryuji , Salden Walter , Saegusa Hiromitsu , Arai Takashi , Matsuki Koji
Published: June 2009

A hydrogeological conceptual model has been developed based on pressure responses observed at multilevel pressure monitoring zones in seven boreholes and surface tilt data in and around the Mizunami Underground Research Laboratory site. Pressure changes caused by some earthquakes, cross-hole hydraulic testing, and shaft excavation activities are considered. Surface tilt has been measured from the half way of the shaft excavation phase. The shaft excavation has been commenced from July 2003 with two shafts (Main shaft and Ventilation shaft). By the end of October 2005, discharging of water in the shafts has been halted at the depths of 172m and 191m respectively to allow modifications to be made to the water treatment facility due to an excess of F and B concentration in the water. This results in the recovery of the groundwater levels and filling of the underground workings. Beginning in February 2006 pumping has been resumed and the underground workings have been re-occupied. Continuous groundwater pressure and surface tilt measurements with some numerical analysis during the shaft excavation phase show the existence of the flow barrier fault predicted from the surface-based investigation phase and hydraulic parameter around the shafts.
Development of an Iodine Immobilization Technique by Low Temperature Vitrification With BiPbO2I
Authors: Mukunoki Atsushi , Chiba Tamotsu , Suzuki Yasuhiro , Uehara Seiichiro , Asano Hidekazu , Nishimura Tsutomu
Published: June 2009

This paper describes low temperature vitrification process with BiPbO2I (BPI) as a promising immobilization technique in which Iodine-129 reacts with BiPbO2NO3 (BPN) to form BPI, which is then solidified into a lead-boron-zinc glass matrix (PbO-B2O3-ZnO) using a low temperature vitrification process. Studies with EPMA, STEM-EDS and XRD found that iodine, lead and zinc were homogeneously dispersed in the waste form, and that there were no residual crystalline minerals in the amorphous glass matrix. Leaching tests conducted under typical geological disposal conditions show that iodine dissolves congruently with the BPI glass matrix in simulated seawater, whereas it dissolves incongruently in alkaline Ca(OH)2 solutions. This is due to retention within an altered surface layer.