bluuebunny/crossref_metadata_2025 huggingface dataset:
Conditional Release of Steel From Decommissioning in a Form of Reinforced Concrete
Authors: Pritrsky Jozef , Brodnan Miroslav , Necas Vladimir
Published: August 2012

The paper deals with the conditional release of low-level radioactive steel from decommissioning in a form of reinforced concrete. The main goal was to determine limits for radionuclides concentration and calculate the annual dose for a member of a critical group of public, which should not exceed 10 μSv/year (according to IAEA Safety Guide RS-G-1.7). Corrosion is the principle mechanism of radionuclides release in this case; therefore effort was devoted to assess the time-dependent rate of steel reinforcement corrosion. It was assumed, that concrete is initially highly alkaline (with pH of 12 to 13) because of hydration products such as calcium hydroxide, which keeps the steel surface passive and protected from corrosion. However, carbonic acid resulting from carbon dioxide and water in the atmosphere can react with these products to produce calcium carbonate. This process is referred to as a “carbonation”, and leads after a period of time to significant reduction of the alkalinity (to pH as low as 8.5) followed by destruction of passive layer and starting corrosion of the embedded steel. The analytical principles and a set of input data have been implemented into a mathematical model developed by means of GoldSim software. The paper presents the results of mathematical simulation of corrosion process, which are compared with real measured values.
Taking the UK’s National LLW Progamme From Strategy Development to Implementation
Authors: Rossiter David , O’Donnell Rachel
Published: August 2012

In 2008 UK Nuclear Waste Management Ltd (UKNWM) became the Parent Body Organisation (PBO) at the Low Level Waste Repository (LLWR) in the UK. LLWR is the primary disposal facility for the UK’s LLW, supporting a wide range of industries across the nuclear power generation, reprocessing, defence, healthcare, education, and oil and gas sectors. One of the key tasks following the appointment of the new PBO was to work with the Nuclear Decommissioning Authority (NDA) to develop a national strategy for LLW generated in the UK, predominantly in the NDA estate. The new National Strategy for LLW was required to address the gap between the forecast waste arisings and predicted capacity at LLWR. The National Strategy for LLW Management was published in August 2010 following an 18 month development period. The main focus of the strategy is on three areas: • Application of the waste management hierarchy to extend the life of LLWR and ensure waste is managed in a risk-based, fit-for-purpose manner; • Making best use of existing assets such as transport, packaging, treatment and disposal facilities; • Opening up new fit-for-purpose waste management routes to divert waste away from LLWR. Developing a robust strategy is vital to provide strategic direction to Government, waste producers, regulators, and stakeholders. Once the strategy is developed and approved, the key challenge is then to implement the strategy on a national scale in an efficient and cost-effective manner that delivers maximum value for money to the UK taxpayer. As well as developing the strategy, LLWR has been actively working to develop the enablers to implement the strategy. Since the publication of the strategy in August 2010 LLWR has been re-organised to reflect the shift in focus, from strategy development to implementation and delivery of the strategy. New resources have been brought in with international waste management experience to help integrate delivery with waste producers. This paper covers the changes in focus required from developing the strategy to how this is implemented. This includes the development of metal recycling, incineration, VLLW disposal, characterisation, packaging transport services. These services have been developed to allow the key aims in the strategy to be achieved, and the cultural changes that both LLWR and the customers have had to adapt to, as well as future developments in waste services.
Uranium Workshops Dismantling and Criticality Control
Authors: Lisbonne Pierre , Rosello Julie
Published: August 2012

Within the CEA at Cadarache Research Center located in southern France, the ATUE, standing for Ateliers de Traitement de l’Uranium Enrichi (Enriched Uranium Treatment Workshops), is a nuclear facility started in 1965 and shut down in 1995. Various chemical processes were in operation in the ATUE in order to product enriched uranium oxide intended for fast breeders and naval propulsion reactors. According to the mass of fissile material, the most contaminated process equipment has already been dismantled under criticality risk management procedures. Cleanup of the building and dismantling of the remaining equipment, such as ventilation and electricity network, are now in progress without any constraint regarding criticality, thanks to in-situ measurement and calculations. Criticality management and how we finally proved the risk was over, as well as the strategy under French rules for final decommissioning of the facility leading to “brown field” will be presented.
Characterization of a Deep Radiological Contamination: Integration of Geostatistical Processing and Historical Data
Authors: Desnoyers Yvon , de Moura Patrick
Published: August 2012

The problem of site characterization is quite complex, especially for deep radiological contamination. This article illustrates the added value of the geostatistical processing on a real application case dealing with grounds of facilities partially dismantled at the end of the 1950s in Fontenay-aux-Roses CEA Center (France). 12 years ago, a first exploratory drill-hole confirmed the presence of a deep radiological contamination (more than 4m deep). More recently, 8 additional drill-holes failed to delineate the contamination extension. The integration of the former topography and other geological data led to the realization of 10 additional drillholes. This final stage significantly improved the characterization of the radiological contamination, which impacted the remediation project and the initially estimated volumes.
Treatment of Spent Acidic Decontaminants With a High-Efficiency Cementation Method
Authors: Lin Kou-Ming , Chang Ching-Tu , Wu Ming-Shin , Lee Wen-Chen , Chung Jen-Chieh , Wei Tsong-Yang
Published: August 2012

Metal scrap is a major waste generated from the decommissioning of nuclear facilities. Through a decontamination process, most of the metal scraps can be cleaned to meet the clearance levels, which can then be reused or released according to the regulatory procedure. Usually, chemical processes will be used in the cleaning step. Phosphoric acid and fluoroboric acid are the typical chemicals used for decontamination. Although the decontaminant could be reused multiple times after regeneration, its decontamination efficiency would decrease after 3 to 5 cycles. In addition, the radioactive nuclides such as Cs-137 are not easily removed during the regeneration process; it tends to accumulate slowly in the decontaminant. According to the ALARA principle, decontaminant must be replaced if its radioactive activity exceeds the regulatory levels. As a result, a significant amount of spent strong acid solution would be generated. The conventional way of treatment is to neutralize the acid solution with an alkaline solution. However, such method will produce a large amount of sludge that requires further stabilization, which offsets the advantages of metal decontamination by use of the decontaminant. A high-efficiency solidification method has been developed and used to treat the spent phosphoric acid and fluoroboric acid solution in Institute of Nuclear Energy Research (INER). The self-polymerization nature of highly concentrated phosphoric acid is adopted to immobilize the radioactive nuclides. The volume of solidified form is almost equal to that of the treated acid solution. The waste form demonstrates its quality by compression test and leaching test. This cementation process is an excellent method to minimize the secondary waste, which is generated from chemical decontamination for treating metal waste.
Optimization of Radioactive Waste Cementation for Decommissioning of Salaspils Research Reactor
Authors: Abramenkova G. , Abramenkovs A. , Klavins M.
Published: August 2012

This paper deals with information on the radioactive waste cementation technology for decommissioning of Salaspils Research Reactor (SRR). Dismantled and segmented radioactive materials were cemented in concrete containers using tritiated water-cement mixture. The viscosity of water-cement mortar, mechanical tests of solidified mortar’s samples, change of temperature of the samples during solidification time and long time leakage of 137Cs, 14C, 60Co and 3T radionuclides was studied for different water-cement compositions with additives. The pH and electro conductivity of the solutions during leakage tests were controlled. It was shown, that water/cement ratio significantly influences on water-cement mortar’s viscosity and solidified samples mechanical stability. The role of additives — fly ash and Penetron admix in reduction of solidification temperature is discussed. It was found, that addition of fly ash to the cement-water mortar can reduce the solidification temperature from 81°C up to 62°C. The optimal interval of water ratio in cement mortar is discussed. Radionuclides leakage tests show that the release curves has a complicate structure. The possible radionuclides release mechanisms are discussed. Experimental results indicated that additives can significantly influence on the radionuclides release processes from cemented samples. The optimization of cementation of radioactive wastes in concrete containers was performed using mechanical stability, solidification temperature, radionuclide releases and viscosity of mortar.
Identification and Implementation of End-User Needs During Development of a State-of-the-Art Modeling Toolset
Authors: Seitz Roger , Freshley Mark , Williamson Mark , Dixon Paul , Gerdes Kurt , Collazo Yvette T. , Hubbard Susan
Published: August 2012

The U.S. Department of Energy (US DOE) Office of Environmental Management, Technology Innovation and Development is supporting a multi-National Laboratory effort to develop the Advanced Simulation Capability for Environmental Management (ASCEM). ASCEM is an emerging state-of-the-art scientific approach and software infrastructure for understanding and predicting contaminant fate and transport in natural and engineered systems. These modular and open-source high performance computing tools and user interfaces will facilitate integrated approaches that enable standardized assessments of performance and risk for EM cleanup and closure decisions. The ASCEM team recognized that engaging end-users in the ASCEM development process would lead to enhanced development and implementation of the ASCEM toolsets in the user community. End-user involvement in ASCEM covers a broad spectrum of perspectives, including: performance assessment (PA) and risk assessment practitioners, research scientists, decision-makers, oversight personnel, and regulators engaged in the US DOE cleanup mission. End-users are primarily engaged in ASCEM via the ASCEM User Steering Committee (USC) and the ‘user needs interface’ task. Future plans also include user involvement in demonstrations of the ASCEM tools. This paper will describe the details of how end users have been engaged in the ASCEM program and will demonstrate how this involvement has strengthened both the tool development and community confidence. ASCEM tools requested by end-users specifically target modeling challenges associated with US DOE cleanup activities. The demonstration activities involve application of ASCEM tools and capabilities to representative problems at DOE sites. Selected results from the ASCEM Phase 1 demonstrations are discussed to illustrate how capabilities requested by end-users were implemented in prototype versions of the ASCEM tool.
Development of Nuclear Facilities Piping Cleaning System Using Microbubble
Authors: Jeon Jongseon , Lee Sangchul , Kim Haksoo , Han Byoungsub , Kim Wisoo
Published: August 2012

It removes radioactive sludge and corrosion products deposited on the inner walls of the pipes and valves in replacement or decommission, upon termination of life time, of nuclear power plant or nuclear facility. It lowers a cost of waste treatments taking advantage of a reduction of quantity of radioactive wastes by treating in classification of the radioactive wastes whose activities are lower than legal standards. The cleaning or decontamination methods developed until now have induced a damage on systems while being operated. A decontamination has been restrained if it was difficult to access physically. We are in development of the cleaning technique for pipelines by utilizing micro-bubbles in order to improve an efficiency and to prevent from any damage of systems. It aims to conduct a decontamination for spaces difficult to access there by applying cavitation phenomenon that is generated in collapse of micro-bubbles. In order to improve an efficiency of the micro-bubble device, the experimental conditions suitable to decontamination have been established and the auxiliary equipments have been added. The generation conditions and characteristics of micro-bubbles have been demonstrated by adjusting pHs and temperatures of cleaning solution. A decontamination effect has been confirmed by adding up an electrolytic method and ozone into micro-bubbles.
Gas Transport Properties of Pumice Tuff for Performance Assessment of LLW Disposal Facility
Authors: Yamamoto Shuichi , Suzuki Kenichiro , Kumagai Mamoru , Tawara Yasuhiro , Mori Koji
Published: August 2012

In Japan, some of the radioactive waste with a relatively higher radioactivity concentration from nuclear facilities is to be packaged in rectangle steel containers and disposed of in sub-surface disposal facilities, where normal human intrusion is unlikely to occur. If dissolved oxygen in pore water is consumed by steel corrosion after the closure of the facility, hydrogen gas will be generated from the metallic waste, steel containers and concrete reinforcing bars largely by anaerobic corrosion. If the generated gas accumulates and the gas pressure increases excessively in the facility, the facility’s barrier performance might be degraded by mechanical influences such as fracturing of surrounding rock and cementitious materials or plastic deformation of the bentonite buffer. In this study laboratory experiments for gas and water transport properties of the rock were performed to evaluate gas flow through the rock mass (pumice tuff) around a facility for low level waste disposal. Based on the experimental results two-phase flow properties were evaluated by means of an inverse analysis method. The pumice tuff was subjected to hydraulic conductivity tests, water retention (moisture characteristic) tests, and gas injection tests. Non-linear properties such as relative permeability and water retention curve and hydraulic conductivity as a function of confined stress obtained from these tests are discussed. It was possible to estimate the intrinsic permeability, the relative permeability for gas and water and the water retention curve by applying an inverse analysis method using the multi-phase flow analysis code GETFLOWS and universal sensitivity analysis code UCODE_2005 to the gas injection tests. It was found from this study that gas flow in the pumice tuff is reasonably well described by classical two-phase flow concepts and that the two-phase flow properties can be applied to performance assessment of the facility with regard to the influence of gas generation and migration.
High-Density Gamma Radiation Spectrometry Surveys of Contaminated Land
Authors: Davies Mike , Clark Robert , Adsley Ian
Published: August 2012

The design of gamma radiation surveys of contaminated land has always had to compromise between the density of sampling and the accuracy of measurement. Large-area contamination is readily detected by a variety of measurement techniques, while the detection of small areas or ‘particles’ requires a high density of measurement, generally one measurement per square metre. High Resolution Gamma-radiation Spectrometry (HRGS) can provide accurate qualitative (radionuclide identification) and quantitative (Bq.g−1 of the radionuclide for a stated scenario) assessment of the state of the land, but are not cost effective at high density. In contrast, simple walk-over or ‘scan’ surveys using standard Health Physics instruments can provide a high density of measurement, but cannot provide the qualitative and quantitative accuracy of HRGS. In 1996, Nuvia Limited adopted methods to address some of these issues, by allowing scan surveys to include a degree of qualitative analysis of the gamma radiation detected using Low Resolution Gamma-radiation Spectrometry systems with Sodium Iodide detectors, while maintaining a high density of measurement. While low-resolution systems (including medium-resolution Lanthanum Bromide) have become the de-facto standard for large area land surveys, the use of the technology has changed little. High density scan surveys can be conducted using, for example, simple gross-gamma techniques, doserate or region-of-interest logging. However, if spectra are required for qualitative purposes, they are normally collected at fixed locations and for inconveniently long counting periods. The ideal would be to collect spectra at every measurement location, preferably once per second, and then ‘aggregate’ the spectra using spatially-aware techniques. This would allow a scan survey to be performed rapidly, while losing no spectral information. It would thus be possible to analyse the data in a number of ways, for example, producing justifiable doserate measurements or using neural-network techniques to search the measurements for spectral anomalies. This paper describes the work done by Nuvia to develop a system to collect spectra efficiently and to make the spectra readily available for a number of analyses. The efficiency of spectrum acquisition and compression are discussed, along with methods of managing the potentially large volumes of data. The analysis of the data, using a customised Geographical Information System is described, including spatial aggregation of spectra and semi-automatic analysis of spectral structure for identifying common ‘background’ features. Examples of the use of these facilities in Nuvia’s ‘Groundhog’ site survey service are provided.
New Nuclear Programmes Must Not Neglect Waste Management
Authors: McCombie Charles
Published: August 2012

Many established nuclear power programmes have learned to their dismay that waste management and disposal are not tasks that can be postponed at will if public and political acceptance is a prerequisite for progress. In fact, some programmes that recognised this back in the 1970s and 1980s moved into leading positions in repository development. This happened, for example, in Sweden and Switzerland where already in the 1970s Laws were passed specifying that safe disposal must be demonstarted before new nuclear plants could opersate. In recent years, it has become recognised that, in order to ensure that the radioactive wastes in any country are managed safely, it is necessary to have an established legislative and regulatory framework and also to create the necessary organizations for implementation and for oversight of waste management operations and facility development. Guidance on these issues is given in the Joint Convention and a number of other IAEA documents. The IAEA, and also the EC, have in addition published key overarching advisory documents for new nuclear programmes. These are useful for strategic planning but, when it comes to actual implementation projects, the advice tends to imply that all nuclear programmes, however large or small, should be pressing ahead urgently towards early operation of geological repositories. In practice, however, in small programmes there are neither economic nor technical drivers for early implementation of deep geological repositories. Constructing simpler facilities for the disposal of the larger volume of low-level wstes has higher priority. Nevertheless, in all countries political decisions have to be taken and policies set in place to ensure that geological disposal will implemented without unjustified delay. This paper distils out a set of key messages for new programmes. Amongst the most critical are the following. Even if disposal is far off, planning and organization should begin at the initiation of the programme; this can help with technical and economic optimization and (importantly) also with public and political acceptance. Important lessons can be learned from advanced programmes — but these must be adapted to allow for the different boundary conditions of new and small programmes. The key differences relate to the timescales involved, and the resources available. There is a range of waste management and waste disposal options open to new programmes. It is not necessary to choose definitive solutions at the outset; options can be kept open, but a minimum level of engagement is required for all open options.
Bayesian Statistics Applied to the Activities of Waste Drums
Authors: Bull Richard K. , Adsley Ian
Published: August 2012

Gamma spectrometry is widely used to determine the radioactive content of waste drums. However, the results of such surveys often result in large numbers of limit-of-detection (LOD) results. In a previous paper it was shown how simple statistical methods could be used to make estimates of the mean activity of a set of waste drums even when the individual measurements were dominated by limit-of-detection results. In that approach it was necessary to make assumptions about the statistical distribution (taken to be lognormal) of the activities and the geometric standard deviations of the distributions. In this paper a development of the method is presented which uses two statistical methods — Bayes’ Theorem and marginalization. These allow the probability distribution of the mean activity of the waste drums to be calculated, whilst the number of assumptions is reduced to just one, namely the form of the statistical distribution. The mathematical details are described in the paper and the application to a set of U235-in-drum measurements is presented.
A Matrix-Inversion Method for Gamma-Source Mapping From Gamma-Count Data
Authors: Bull Richard K. , Adsley Ian , Burgess Claire
Published: August 2012

Gamma ray counting is often used to survey the distribution of active waste material in various locations. Ideally the output from such surveys would be a map of the activity of the waste. In this paper a simple matrix-inversion method is presented. This allows an array of gamma-count data to be converted to an array of source activities. For each survey area the response matrix is computed using the gamma-shielding code Microshield [1]. This matrix links the activity array to the count array. The activity array is then obtained via matrix inversion. The method was tested on artificially-created arrays of count-data onto which statistical noise had been added. The method was able to reproduce, quite faithfully, the original activity distribution used to generate the dataset. The method has been applied to a number of practical cases, including the distribution of activated objects in a hot cell and to activated nimonic springs amongst fuel-element debris in vaults at a nuclear plant.
THOR® Steam Reforming Technology for the Treatment of Complex and Problematic Wastes
Authors: Mason J. Brad , Foster Adam L.
Published: August 2012

Many types of complex radioactive wastes exist, that due to their chemical and radiological characteristics, are problematic to treat using traditional methods. Studsvik has developed the THOR® process that has the capability to successfully treat a wide variety of radioactive wastes including those with high organic content, high nitrate and nitrite content, high solids content, and heavy metal bearing. The process produces a dry stable waste form with no liquid effluents and environmentally compliant off-gas emissions, while reducing the overall waste volume. The more than 10 years of successful operations at Studsvik’s facility in Erwin, TN USA provides the foundation for the deployment of the THOR® steam reforming process for the treatment of many complex wastes. The THOR® process has been shown to effectively convert numerous types of U.S. Department of Energy (DOE) liquid radioactive wastes into solid products while providing a final waste form with leach resistance comparable to vitrified glass waste forms. In addition to testing and on-going project work in the U.S., multiple test programs including lab-scale and engineering-scale demonstrations have been successfully completed for several international organizations. This international work has included testing and development work with solid and liquid organic wastes, as well as high nitrate content liquid waste and graphite. In order to better meet the needs of the international market, Studsvik offers a modular form of its THOR® process that is transportable to multiple locations, enabling customers to treat a wide variety of waste streams at multiple sites without the need to design and construct multiple treatment facilities. This paper will provide an up to date overview of the THOR® technology, its current capabilities, and a summary of current THOR® projects.
Computational Study for Inventory Estimation of Se-79, Tc-99, Sn-126, and Cs-135 in High-Level Radioactive Wastes From Spent Nuclear Fuels of Light Water Reactors
Authors: Okumura Keisuke , Asai Shiho , Hanzawa Yukiko , Okamoto Tsutomu , Suzuki Hideya , Toshimitsu Masaaki , Inagawa Jun , Kimura Takaumi , Suzuki Kensuke , Kaneko Satoru
Published: August 2012

Inventory estimation of long-lived fission products (LLFPs) in high-level radioactive wastes (HLW) from spent nuclear fuels of light water reactors is important for a safety assessment of their disposal. In order to develop an inventory estimation method of difficult-to-measure LLFPs (Se-79, Tc-99, Sn-126, and Cs-135), a parametric study was carried out by using a sophisticated burnup calculation code and data. In the parametric study, fuel specifications and irradiation conditions are changed in the conceivable range. The considered parameters are fuel assembly types (PWR / BWR), U-235 enrichment, moderator temperature, void fraction, power density, and so on. From the calculated results, we clarify the burnup characteristics of the target LLFPs and their possible ranges of generations. Finally, candidates of the key nuclide are proposed for the scaling factor method of HLW.
Hydraulic/Mechanical Modeling of Smectitic Materials for HMC Analytical Evaluation of the Long Term Performance of TRU Geological Repository
Authors: Kobayashi Ichizo , Owada Hitoshi , Ishii Tomoko
Published: August 2012

Aiming at evaluation of the long term performance of transuranic (TRU) geological repositories, the hydraulic/mechanical/chemical (HMC) analysis method has been studied. In this phase of research (four years) the hydraulic/mechanical modeling of smectitic materials for HMC analyses has been studied. In this paper, new experimental methods for investigation of the hydraulic/mechanical behavior of smectitic materials were developed. For hydraulic modeling, the measurement method of the specific surface area of compacted smectitic materials was developed using X-ray diffraction (XRD). The results of the method were applied to the Kozeny-Carman law. Since the specific surface area represents the microstructure of smectitic materials such as the degree of swelling, it was found that the Kozeny-Carman law using measured specific surface area of compacted smectitic materials was useful in evaluating the hydraulic performance of smectitic materials. Moreover, since the Kozeny-Carman law can take the alteration of content of pore water into consideration by not only a coefficient of viscosity but also by changes in specific surface area, the Kozeny-Carman law will be more suitable to chemical and mechanical couple analyses than the ordinary Darcy’s law. For the mechanical modeling, the procedure of one dimensional exhausting compression test was developed. The tests gave the dry density and compression stress relation in the state of full-saturation of smectitic materials with varying water content. The relations between the dry density and compression stress in the state of fully-saturation were termed fully saturation lines. The group of isograms of degree of saturation and water content were also given with this test. It was found that the fully-saturation line is consistent with swelling deformation-pressure relation in the equilibrium state. The results indicated that the swelling deformation-pressure relation does not depend on the saturation manner, such as the injecting of water or exhausting of air due to compression. There will be, therefore, the possibility that the swelling deformation-pressure relation will be state functions and independent on the stress history. The fully-saturation line was compared with the equilibrium swelling pressure and dry density relation calculated from chemical potential by Sato (2008). Both were consistent with each other. It was found that the swelling behavior of smectitic materials can be treated as a state function in the mechanical analysis.
Capillary Electrophoresis With Laser-Induced Fluorescent Detection Method Using Highly Emissive Probes for Analysis of Actinides in Radioactive Wastes
Authors: Haraga Tomoko , Nakano Yuta , Shibukawa Masami , Kameo Yutaka , Takahashi Kuniaki , Saito Shingo
Published: August 2012

Actinides are important nuclides for the analysis of radioactive wastes arising from nuclear fuel cycle facilities. In order to achieve simple and rapid analysis of actinides, capillary electrophoresis-laser-induced fluorescent detection method (CE-LIF) is one of the potential candidates. In this study, new emissive probes of actinide ions suitable for CE-LIF were developed for the first time. The detection and separation of americium and neptunium ions as model nuclides were examined using several new emissive complexing probes, each of which possessed a fluorophore and a different chelating moiety. With a pre-capillary complexation technique without addition of the probe to separation buffer electrolyte, the highly sensitive fluorescent detection of Am and Np was successfully achieved using acyclic and macrocyclic multidentate probes. The results suggests that the probe with an acyclic hexadentate chelating moiety is suitable for detection and separation of Am and Np. The detection limit of mid–ppt levels was determined.
Mineralogical Analyses of Old (78 and 98 Years) Concrete
Authors: Ishii Tomoko , Owada Hitoshi , Sakamoto Hiroyuki , Shibata Masahito , Negishi Kumi
Published: August 2012

Because Ca-dissolution from cementitious material is considered a source of long-term alteration of the performance of radioactive waste repositories, much research including dissolution tests has been conducted on this topic. These studies have introduced models such as Atkinson’s model [1] to calculate the leaching of cementitious material. These models have been used to verify that the results of many studies do represent the alteration of cementitious minerals and the composition of the leachate. They have also been used to make numerous estimates of long-term mineralogical alteration in repositories, such as cement-clay interaction in cementitious barrier systems, and to evaluate the change in repository performance. However, immersion tests using bulky cementitious material have often indicated that the actual alteration of cementitious material might be slower than the rates calculated by these models. This difference may be due to a change of mass-transport characteristics, either in the bulky cementitious material or at the interfaces with other materials. In this study, a mineralogical analysis was conducted on two types of old concrete. Drilled cores from the foundations of rotary kilns at two cement factories were collected beneath the groundwater level. Both concrete structures were made from Japanese ordinary Portland cement (OPC), which is similar to European type 1 cement. One structure had been set into a fresh groundwater environment for 78 years (78-F), and the other had set sunk into a saline environment for 98 years (98-S).
Characterization of Bedrock Hydrogeology at the Olkiluoto Site Using Surface Based and Underground Data
Authors: Hartley Lee , Swan Dave , Baxter Steven
Published: August 2012

Posiva Oy is responsible for implementing the program for geological disposal of spent nuclear fuel in Finland. Olkiluoto in Eurajoki has been selected as the primary site for the repository, subject to further detailed characterization which is currently focused on the construction of an underground rock characterization and research facility (the ONKALO). An essential part of the site investigation program is analysis of groundwater flow and solute transport to depth by means of numerical flow modeling as part of the 2011 site descriptive model. Groundwater flow in the crystalline rocks at Olkiluoto takes place predominantly in the void space of the interconnected fractures in the bedrock. Many of the hydraulic characteristics of flow through such a system can be represented by hydrogeological Discrete Fracture Network (hydro-DFN) models, since they capture some of the details of fracture geometry, size, connectivity and openings. The paper describes how field data is used to derive such models, ultimately for use in safety assessment. Surface based site investigations at Olkiluoto include 53 vertical or inclined deep core drilled boreholes and 27 shallower sub-vertical boreholes. The characterization of fractures is a key objective. Fracture geometrical data (fracture positions and orientations) are determined from drill core mapping and/or borehole TV images, while hydraulic data are determined with the Posiva Flow Log (PFL). The PFL method is a geophysical logging device developed to detect continuously flowing fractures in sparsely fractured crystalline bedrock by means of difference flow logging, providing the effective transmissivity of individual fracture intercepts. A methodology developed originally in the Swedish program for constraining hydrogeological DFN models based on the PFL high resolution hydraulic test data has been extended to incorporate data from underground. The approach integrates deterministic models of the main hydro-structures with a stochastic model of the bedrock between. An exploration of conceptual uncertainties in the representation of the very scarce distribution of flow at repository depth, c. 400m, has been performed. This has been greatly informed by the analysis of 15 additional sub-horizontal boreholes drilled ahead of the ONKALO tunnel excavations. Conditions for PFL hydraulic tests performed in these pilot holes provide a lower detection limit and reveal the very low magnitude of groundwater flow that may be expected around the repository depositional volumes.
Modeling of Alteration Behavior on Blended Cementitious Materials
Authors: Owada Hitoshi , Ishii Tomoko , Takazawa Mayumi , Kato Hiroyasu , Sakamoto Hiroyuki , Shibata Masahito
Published: August 2012

A “realistic alteration model” is needed for various cementitious materials. Hypothetical settings of mineral composition calculated based on the chemical composition of cement, such as Atkins’s model, have been used to estimate the alteration of cementitious material. However, model estimates for the concentration of certain elements such as Al and S in leachate have been different from experimental values. In a previous study, we created settings for a mineralogical alteration model by taking the initial chemical composition of cementitious materials from analysis results in experiments and applying their ratios to certain hydrated cement minerals, then added settings for secondary generated minerals in order to account for Ca leaching. This study of alteration estimates for ordinary portland cement (OPC) in groundwater showed that the change in Al and S concentrations in simulated leachate approached values for actual leachate[1]. In the present study, we develop an appropriate mineral alteration model for blended cementitious materials and conduct batch-type leaching experiments that use crushed samples of blast furnace slag cement (BFSC), silica cement (SC), and fly ash cement (FAC). The cement blends in these experiments used OPC blended with blast furnace slag of 70 wt.%, silica cement consisting of an amorphous silica fine powder of 20 wt.%, and fly ash of 30 wt.%. De-ionized water was used as the leaching solution. The solid-liquid ratios in the leaching tests were varied in order to simulate the alteration process of cement hydrates. The compositions of leachate and minerals obtained from leaching tests were compared with those obtained from models using hypothetical settings of mineral composition. We also consider an alteration model that corresponds to the diversity of these materials. As a result of applying the conventional OPC model to blended cementitious materials, the estimated Al concentration in the aqueous solution was significantly different from the measured concentration. We therefore propose an improved model that takes better account of Al behavior by using a more reliable initial mineral model for Al concentration in the solution.
Decommissioning and Demolition of Facilities Used for the Storage and Stabilisation of Water Reactor Sludge
Authors: Brown N. , Cornell R. M. , Staples A.
Published: August 2012

Nuvia Limited was contracted to design, build and operate a waste treatment plant to stabilise a quantity of about 300m3 of active sludge stored in the External Active Storage Tanks (EAST) at the former United Kingdom Atomic Energy Authority (UKAEA) research site at Winfrith, UK. At the end of this process both the old and new plants are to be decommissioned and demolished with the minimisation of waste material volumes. The sludge was produced in the Steam Generating Heavy Water Reactor (SGHWR), which is now in the early stages of decommissioning. As part of the reorganisation of UKAEA, responsibility for the site now lies with RSRL (Research Sites Restoration Limited) with funding provided by the Nuclear Decommissioning Authority (NDA). The process of stabilization of the SGHWR sludge from the EAST tanks within 500 litre stainless steel drums in the newly constructed Winfrith EAST Treatment Plant (WETP) was completed in March 2010. The plant will now enter a full decommissioning phase since no further work on any other waste materials has been identified by the client. This paper describes the development of a decommissioning plan for the mixture of old and new structures such that this next process can be undertaken without delay. This involved the completion of a decommissioning safety case together with supporting documents to demonstrate best practice and close attention to ALARP issues. The potential techniques for decommissioning are discussed leading to the final processes selected for this project. The challenges to be overcome, particularly with the older plant items where local contamination of their structures is suspected, are also identified and addressed. A process of recovery of concrete core samples for radiochemical analysis and their outcomes will be included. In contrast, the means by which the newer WETP facility was shown to be capable of unrestricted demolition as a result of the care taken during the design and construction phase will also be explained in the context of the practicalities, minimisation of the costs and timescale of such operations. Finally, issues concerning the minimisation of waste volumes, particularly where tritium contamination is present, are addressed within a UK-based disposal environment. The link between these outcomes and the planning and execution of the various structure surveillance tests will also be described and potential waste volumes identified. These considerations may be of interest to other groups undertaking similar projects on mixtures of older and more recent plants in varying states of contamination.
Characterization and Assessment of the Groundwater Pathway for the Low Level Waste Repository, UK
Authors: Hartley Lee J. , James Martin , Jackson Peter , Couch Matt , Shevelan John
Published: August 2012

The Low Level Waste Repository (LLWR) is the UK’s principal facility for the disposal of solid low-level radioactive waste and is operated by LLW Repository Limited. Presently, LLWR Ltd is establishing the long-term environmental safety of disposals of solid radioactive waste at the LLWR, through the submission of the 2011 Environmental Safety Case for the LLWR. This Environmental Safety Case addresses the Environment Agency Guidance on Requirements for Authorisation. Aspects of the submission consider improved vault design, closure design, and quantitative assessments. Each of these issues requires an understanding of the movement of water through the facility and the surrounding geology during operations and following facility closure. Groundwater flow modelling has been used extensively in support of the interpretation of field investigations, the development of the engineering design, and an assessment of the groundwater pathway as one of the major pathways by which contaminants may reach the environment. This paper describes these important aspects of the Environmental Safety Case. The geological environment in the region of the LLWR consists of Quaternary age deposits overlying older bedrock. The facility involves shallow excavations into the Quaternary deposits, originally for trenches, with disposals to a vault system beginning in 1988. In the post-closure phase these disposals are covered by a cap and surrounded by a cut-off wall to minimise the water flow around or through the waste. An innovative modelling methodology has been developed to represent the range of scales that have to be considered from the regional groundwater flow patterns over several kilometres, the scale of tens of metres around the immediate site area, and down to about 1 metre for details of flows within the repository itself in three dimensions. Detailed finite-element models of the flow through geological media and the engineered features are used to interpret site data and assess a credible set of post-closure situations and model cases. In the radiological assessment, a more simplified compartment model is used to assess uncertainties in hydrogeological properties and the long-term evolution of the engineered barriers. Together the approach provides flexible tools for understanding and assessing a comprehensive range of aspects including details of flows within the repository, dilution and migration in the external geology, the long-term evolution of the hydrogeological system, the implications of spatial variability and alternative geological models, and effects of uncertainties.
UK Contributions to the Decommissioning of the BN-350 Reactor in Kazakhstan: 2002–2011
Authors: Laws Steven , Wells David , Herrick Andrew
Published: August 2012

Since 2002, the UK’s Global Threat Reduction Programme managed by the Department of Energy and Climate Change (DECC) has provided assistance to the Republic of Kazakhstan with the decommissioning of the BN-350 sodium cooled fast reactor. Assistance has focused on non-proliferation, safety and security projects to ensure the permanent and irreversible shutdown of the reactor and the reduction of security, safety and environmental hazards, particularly those associated with the large inventory of liquid metal coolants (sodium and sodium-potassium alloy) and the presence of spent nuclear fuel (SNF). UK assistance efforts have been co-ordinated with those of the USA and have made use of the UK’s experience in decommissioning its own fast reactor plants at Dounreay. The paper describes work undertaken with UK technical and funding assistance support in the following areas: • Provision of training and technical support in project management and technical topics, including assistance with completion of the BN-350 Decommissioning Plan. • Liquid metal coolant treatment projects, including immobilization of liquid products from the Sodium Processing Facility (SPF) and processing of residual sodium remaining within the drained coolant circuits. • Immobilization of highly active caesium traps, arising from sodium clean-up both during reactor operations and post-shutdown. • Operations to transfer the entire inventory of spent nuclear fuel from the reactor storage pond into dual-use storage and transport casks and then consign these casks to long-term secure storage remote from the reactor site. This activity was part of the major US-Kazakhstan SNF Storage Project. • Surveys of spent fuel route facilities to establish the absence of any significant amount of nuclear material. Key achievements in 2010 were the successful completion of residual sodium processing and completion of the SNF Storage Project. Through 2011, it is intended that the surveys of the fuel storage pond and immobilization of caesium traps will be completed, bringing the current UK assistance activities to an end before March 2012.
Elution Behavior of Heavy Metals From Cement Solidified Products of Incinerated Ash Waste
Authors: Meguro Yoshihiro , Kawato Yoshimi , Nakayama Takuya , Tomioka Osamu , Mitsuda Motoyuki
Published: August 2012

A method, in which incinerated ash is solidified with a cement material, has been developed to dispose radioactive incinerated ash waste. In order to bury the solidified product, it is required that elution of hazardous heavy metals included in the ash from the solidified products is inhibited. In this study, the elution behavior of the heavy metals from the synthetic solidified products, which included Pb(II), Cd(II), and Cr(VI) and were prepared using ordinary portland cement (OPC), blast furnace slag cement (BFS), or a cement material that showed low alkalinity (LA-Cement), was investigated. Several chemicals and materials were added as additive agents to prevent the elution of the heavy metals. When OPC was used, Cd elution was inhibited, but Pb and Cr were not enough even using the additive agent examined. FeSO4 and Na2S additive agents worked effective to inhibit elution of Cr. When BFS was used, the elution of Pb, Cd and Cr was inhibited for the all products prepared. In the case of LA-Cement, the elution of Pb and Cd was inhibited for the all products, but only the product that was added FeSO4 showed good result of the elution of Cr.
A Room Temperature Stabilization and Waste Volume Reduction Method Tested on Belgian Intermediate-Level Homogeneous Bituminized Waste (Eurobitum)
Authors: Impens Nathalie R. E. N. , Adriaensen Lesley , Deckers Jan , Van Iseghem Pierre
Published: August 2012

We developed a room temperature method to recondition intermediate-level homogeneous bituminized waste, called Eurobitum. Using this method, Eurobitum could be separated in an aqueous and an organic effluent, and a solid asphaltene-salt mix. In this paper, we present the results of the room temperature reconditioning method using real radioactive Eurobitum samples. The dissolution and separation processes were optimized. The residual waste could then be pressed towards a material, that might be suitable for geological disposal. The concentration of radionuclides in the two secondary waste effluents was thereafter determined. The aqueous effluent contained radionuclide concentrations below the treatment levels for common radioactive waste water treatment processes, whereas the alpha radioactivity of the organic phase was low enough to consider the possibility to inject it into the CILVA radwaste incinerator where it would undergo a thermal treatment process together with other solid radwaste. Our proposed process could reduce the Eurobitum volume considerably. The most flammable and the least radiation resistant molecules and the nitrates were removed from the original Eurobitum waste. Finally, the residual waste is expected to be more stable in geological disposal conditions.