bluuebunny/crossref_metadata_2025 huggingface dataset:
Accelerator Waste: A New Challenge for Radioanalytics
Authors: Schumann Dorothea , Wohlmuther Michael , Neuhausen Jo¨rg
Published: June 2009

The Paul Scherrer Institute (PSI) at Villigen (Switzerland) operates the most powerful accelerator facility in Europe. Due to the increasing quantities of accelerator waste with almost unknown radionuclide inventory, the development of new radio-analytical methods is an urgent task. Besides the characterization by γ-measurements and dose rate determinations, also the investigation of long-lived radionuclides, being probably essential for a final disposal, is required from Swiss authorities. Chemical separation is necessary for the determination of the majority of these isotopes. As a representive example for such studies, the analytics of a beam dump assembly is introduced. Samples were taken from the target E beam dump station from the 590 MeV proton accelerator facility. The content of several radionuclides with half-lives between 2 and 107 years was determined by γ-spectrometry and, after chemical separation, by Liquid Scintillation Counting (LSC) as well as Accelerator Mass Spectrometry (AMS). The results are compared with theoretical predictions. Long-term object of these studies is the elaboration of nuclide vectors, which allow the estimation of nuclide inventories by simple calculations.
Turbulent Impinging Flow Simulation for High-Level Waste Storage and Processing Applications
Authors: Rhea S. , Fairweather M.
Published: June 2009

The efficient storage and processing of high-level nuclear waste could be improved by a better understanding of the behaviour of the particle-laden fluid flows involved. This work reports a mathematical modelling study of impinging single- and two-phase turbulent jets that is of relevance to the flows used industrially to prevent the settling of solid particles in storage tanks, and to re-suspend particles that form a bed. A computational fluid dynamic model, that embodies a Lagrangian particle tracking technique, is applied to the prediction of these flows. Predictions in the free flow and wall regions, and along the stagnation line, of the single-phase flow are in reasonable accord with data, although the addition of particles results in less satisfactory agreement between predictions and measurements. The influence of particles is, however, reproduced qualitatively by the mathematical model, with quantitative differences attributable to a lack of particle drag in the simulations. Uncertainties in experimental parameters may be responsible for some of the differences between predictions and data, and examination of the data used casts doubts on its reliability. Further work is required in terms of the use of more advanced turbulence modelling techniques, and the provision of detailed and reliable data sets.
The Rheology of Oxide Dispersions and the Role of Concentrated Electrolyte Solutions
Authors: Biggs Simon , Tindley Amy
Published: June 2009

Stability control of particulate dispersions is critical to a wide range of industrial processes. In the UK nuclear industry, significant volumes of waste materials arising from the corrosion products of Magnox fuel rods currently require treatment and storage. The majority of this waste is present as aqueous dispersions of oxide particulates. Treatment of these dispersions will require a variety of unit operations including mobilisation, transport and solid-liquid separation. Typically these processes must operate across a narrow optimal range of pH and the dispersions are, almost without exception, found in complex electrolyte conditions of high overall concentration. Knowledge of the behaviour of oxides in various electrolyte conditions and over a large pH range is essential for the efficient design and control of any waste processing approach. The transport properties of particle dispersions are characterised by the rheological properties. It is well known that particle dispersion rheology is strongly influenced by particle-particle interaction forces, and that particle-particle interactions are strongly influenced by adsorbed ions on the particle surfaces. Here we correlate measurements of the shear yield stress and the particle zeta potentials to provide insight as to the role of ions in moderating particle interactions. The zeta potential of model TiO2 suspensions were determined (Colloidal Dynamics ZetaProbe) over a range of pH for a series of alkali metal halides and quaternary ammonium halides at a range of solution concentrations (0.001M–1M). The results show some surprising co-ion effects at high electrolyte concentrations (>0.5M) and indicate that even ions generally considered to be indifferent induce a shift in iso-electric point (i.e.p.) which is inferred as being due to specific adsorption of ions. The shear yield stress values of concentrated titania dispersions were measured using a Bohlin C-VOR stress controlled rheometer. The shear yield stress of a material is defined as the minimum applied shear stress required to induce flow. The yield stress vs. pH curves obtained reflected the shifts in i.e.p. seen in the zeta potential results. Interestingly, specific ion adsorption results in an unexplained increase in the value of the yield stress over that expected for simple systems with no such interfacial ion adsorption. Possible reasons for this effect such as ion-ion correlation effects are discussed. The importance of this increased attraction for the mobilisation of settled solids in an aqueous environment and especially the likely effects on the treatment of Magnox fuel waste materials is discussed.
Development of a Computer Simulation Technique for Predicting Heat Transfer in Multiphase Liquid-Particle Flow Systems
Authors: Malone Kevin F. , Xu Bao H. , Fairweather Michael
Published: June 2009

Many of the highly active waste liquors that result from the reprocessing of spent nuclear fuel contain particulate solids of various materials. Operations for safe processing, handling and intermediate storage of these wastes often pose significant technical challenges due to the need for effective cooling systems to remove the heat generated by the radioactive solids. The multiscale complexity of liquid-particle flow systems is such that investigation and prediction of their heat transfer characteristics based on experimental studies is a difficult task. Fortunately, the increasing availability of cheap computing power means that predictive simulation tools may be able to provide a means to investigate these systems without the need for expensive pilot studies. In this work we describe the development of a Combined Continuum and Discrete Model (CCDM) for predicting the heat transfer behaviour of systems of particles suspended in liquids.
The HRA-SOLARIUM Project: Processing of Historical Waste on the Belgoprocess Site (Belgium)—Project Description and Lessons Learned After 3 Years Operations
Authors: Cuchet J. M. , Luycx P. , Wathion M. , Willems M. , De Goeyse A. , Braeckeveldt M.
Published: June 2009

At the end of the 80’s, the Belgian State ordered an inventory of the liabilities of the Belgian nuclear programme, to be fully or partially financed by them. ONDRAF/NIRAS (National Agency for Radioactive Waste and Enriched Fissile Materials) was entrusted with the management of the waste and the development of a programme for the clearance of the identified liabilities. One of these liabilities is the treatment and conditioning of some 200 m3 of widely varying high- and medium level waste. The gross volume of primary and secondary packages amounts to 2,600 m3. As the waste is stored in vaults or in concrete shielding containers and no appropriate treating and conditioning facilities are in operation, the HRA/SOLARIUM project was launched. The bulk of these wastes, of which 95% are solids, the remainder consisting of mainly solidified liquids, have been produced between 1967 and 1988. They originate from various research programmes and reactor operation at the Belgian nuclear energy research centre SCK·CEN, isotope production, decontamination and dismantling operations. About 4,800 packages of various types are concerned and must be treated (standard steel barrels, special containers, shielded overpacks,…); they contain medium-active wastes (solid or liquid), radium bearing or not, β/γ or α/β/γ, and special wastes (Al, spent resins, Na/Nak, …). The new HRA/SOLARIUM facilities, located on site 2 of Belgoprocess in Mol, have been commissioned in the 2nd semester 2003. The paper describes the project itself and focuses on the lessons learned from first operation years.
Investigation of Particle-Laden Flow in a Straight Duct Using Large Eddy Simulation
Authors: Fairweather M. , Yao J.
Published: June 2009

A particle-laden turbulent flow in a square duct is predicted using large eddy simulation (LES). The simulation is performed for a Reynolds number of 35,500, and correctly predicts the existence of secondary flows and their effects on the mean flow. The results are also in good qualitative agreement with experimental data obtained at different Reynolds numbers. One-way coupling is assumed between solid particles and the fluid, and a particle equation of motion, including Stokes drag, lift, buoyancy and gravity force terms, solved using a Lagrangian particle tracking technique. Three sizes of particle (1, 50 and 100 μm) are considered, and results demonstrate that size has a significant effect on particle dispersion and deposition in the duct flow. As particle size increases, therefore, they tend to settle on the floor of the duct, with less dispersion in the fluid phase. The study demonstrates the usefulness of LES for nuclear waste processing applications since secondary flows occur in many practically-relevant flows, and since it is desirable that the two-phase waste mixture is kept as homogeneous as possible to prevent, or at least discourage, the settling out of solid particles to form a bed which can promote pipe blockages.
Geological Repository for Nuclear High Level Waste in France From Feasibility to Design Within a Legal Framework
Authors: Voizard Patrice , Mayer Stefan , Ouzounian Gerald
Published: June 2009

Over the past 15 years, the French program on deep geologic disposal of high level and long-lived radioactive waste has benefited from a clear legal framework as the result of the December 30, 1991 French Waste Act. To fulfil its obligations stipulated in this law, Andra has submitted the “Dossier 2005 Argile” (clay) and “Dossier 2005 Granite” to the French Government. The first of those reports presents a concept for the underground disposal of nuclear waste at a specific clay site and focuses on a feasibility study. Knowledge of the host rock characteristics is based on the investigations carried out at the Meuse/Haute Marne Underground Research Laboratory. The repository concept addresses various issues, the most important of which relates to the large amount of waste, the clay host rock and the reversibility requirement. This phase has ended upon review and evaluation of the “Dossier 2005” made by different organisations including the National Review Board, the National Safety Authority and the NEA International Review Team. By passing the “new”, June 28, 2006 Planning Act on the sustainable management of radioactive materials and waste, the French parliament has further defined a clear legal framework for future work. This June 28 Planning Act thus sets a schedule and defines the objectives for the next phase of repository design in requesting the submission of a construction authorization application by 2015. The law calls for the repository program to be in a position to commission disposal installations by 2025.
Melting of Low-Level Radioactive Non-Ferrous Metal for Release
Authors: Quade Ulrich , Kluth Thomas , Kreh Rainer
Published: June 2009

Siempelkamp Nukleartechnik GmbH has gained lots of experience from melting ferrous metals for recycling in the nuclear cycle as well as for release to general reuse. Due to the fact that the world market prices for non-ferrous metals like copper, aluminium or lead raised up in the past and will remain on a high level, recycling of low-level contaminated or activated metallic residues from nuclear decommissioning becomes more important. Based on the established technology for melting of ferrous metals in a medium frequency induction furnace, different melt treatment procedures for each kind of non-ferrous metals were developed and successfully commercially converted. Beside different procedures also different melting techniques such as crucibles, gas burners, ladles etc. are used. Approximately 340 Mg of aluminium, a large part of it with a uranium contamination, have been molten successfully and have met the release criteria of the German Radiation Protection Ordinance. The experience in copper and brass melting is based on a total mass of 200 Mg. Lead melting in a special ladle by using a gas heater results in a total of 420 Mg which could be released. The main goal of melting of non-ferrous metals is release for industrial reuse after treatment. Especially for lead, a cooperation with a German lead manufacturer also for recycling of non releasable lead is being planned.
Estimates of Mean Consequences and Confidence Bounds on the Mean Associated With Low-Probability Seismic Events in Total System Performance Assessments
Authors: Pensado Osvaldo , Mancillas James
Published: June 2009

An approach is described to estimate mean consequences and confidence bounds on the mean of seismic events with low probability of breaching components of the engineered barrier system. The approach is aimed at complementing total system performance assessment models used to understand consequences of scenarios leading to radionuclide releases in geologic nuclear waste repository systems. The objective is to develop an efficient approach to estimate mean consequences associated with seismic events of low probability, employing data from a performance assessment model with a modest number of Monte Carlo realizations. The derived equations and formulas were tested with results from a specific performance assessment model. The derived equations appear to be one method to estimate mean consequences without having to use a large number of realizations.
The AVR (HTGR) Decommissioning Project With New Strategy
Authors: Sterner Hakan , Rittscher Dieter
Published: June 2009

The 15-MWel prototype pilot reactor AVR is a pebble bed HTGR. It was designed in the late 50s and was connected to the grid end of 1967. After 21y of successful operation the reactor was shut down end of 1988. In 1994 the first decommissioning license was granted and work with defueling, dismantling and preparation of a Safe Enclosure started. The primary system is contaminated with the fission products Sr90 and Cs137 and the activation products are Co60, C14 and H3. Due to the large amounts of Sr and Cs bound to graphite dust, the dismantling of systems connected to the pressure vessel is very tedious. In 2003 the AVR company was restructured and the strategy of the decommissioning was changed from safe enclosure to green field, i.e. the complete direct dismantling of all facilities and clean up of the site. The highlight during the dismantling is the removal of the reactor vessel (diameter ca. 7.6m and length ca. 26m) in one piece. Before handling the reactor vessel it will be filled with low density cellular concrete. Subsequently the reactor building will be cut open and the reactor vessel (total weight ca. 2100Mg) lifted out and transported to an interim store.
Pre-Decommissioning Radiological Characterization of Concrete
Authors: Boden Sven , Cantrel Eric
Published: June 2009

The decommissioning of the BR3 (Belgian Reactor 3) approaches its final phase, in which the building structures are being decontaminated and either denuclearized for possible reuse or demolished. Apart from the presence of naturally occurring radionuclides in building materials, other radionuclides might be present due to contamination or activation. The overall process of the BR3 building structure D&D (Decontamination & Decommissioning) consists of the following steps: • make a complete inventory and preliminary categorize all elements based on historical data; • characterize and determine the contamination or activation depth; • determine the decontamination method; • perform the decontamination and clean up; • a possible intermediate characterization followed by an additional decontamination step; and • characterize for clearance. A good knowledge of the contamination and activation depth (second step) is fundamental in view of cost minimization. Currently, the method commonly used for the determination of the depth is based on core drilling and destructive analysis. Recently, we have introduced a complementary non destructive assay based on in-situ gamma spectroscopy. Field tests at BR3, both for contamination and activation, showed promising results.
Synthesis of Crystalline Ceramics for Actinide Immobilisation
Authors: Burakov B. , Gribova V. , Kitsay A. , Ojovan M. , Hyatt N. C. , Stennett M. C.
Published: June 2009

Methods for the synthesis of ceramic wasteforms for the immobilization of actinides are common to those for non-radioactive ceramics: hot uniaxial pressing (HUP); hot isostatic pressing (HIP); cold pressing followed by sintering; melting (for some specific ceramics, such as garnet/perovskite composites). Synthesis of ceramics doped with radionuclides is characterized with some important considerations: all the radionuclides should be incorporated into crystalline structure of durable host-phases in the form of solid solutions and no separate phases of radionuclides should be present in the matrix of final ceramic wasteform; all procedures of starting precursor preparation and ceramic synthesis should follow safety requirements of nuclear industry. Synthesis methods that avoid the use of very high temperatures and pressures and are easily accomplished within the environment of a glove-box or hot cell are preferable. Knowledge transfer between the V. G. Khlopin Radium Institute (KRI, Russia) and Immobilisation Science Laboratory (ISL, UK) was facilitated in the framework of a joint project supported by UK Royal Society. In order to introduce methods of precursor preparation and ceramic synthesis we selected well-known procedures readily deployable in radiochemical processing plants. We accounted that training should include main types of ceramic wasteforms which are currently discussed for industrial applications.
Performance-Based Acquisition: A Tool to Reduce Costs and Improve Performance at US Army Environmental Remediation Sites
Authors: Kosko Nancy , Gilman Janet , White Debbie
Published: June 2009

The US Army, like most US federal and state environmental organizations, is faced with limited resources to conduct environmental work, an increasing workload, and challenges in achieving closeout of its environmental cleanup programs. In 2001, in an effort to incorporate proven private sector tools into federal cleanup programs, the Department of Defense (DoD) Business Initiative Council (BIC), initiated the use of Performance-Based Acquisition (PBA) for environmental cleanup. Since fiscal year 2000, the US Army Environmental Command (USAEC) has successfully awarded more than 55 performance-based contracts for environmental remediation. These contracts range in size from $500,000 to $52.4 million, and include closing properties (Base Realignment and Closure (BRAC)) and some of the US Army’s most complex active installations. The contracts address a range of activities including investigation through monitoring and site completion, as well as various technical challenges including dense non-aqueous phase liquids (DNAPL) in ground water, karst systems, munitions and explosives of concern, and biological agents. The contracts are most often firm-fixed price, and 50 percent of the contracts required contractors to purchase environmental insurance in the form of remediation stop loss insurance (also known as cleanup cost cap insurance). The USAEC has conducted continuous process improvement since inception of the initiative. This paper presents results of two studies that were conducted in 2005–2006 to determine what lessons learned can be applied to future activities and to measure performance of contractors currently executing work under the performance based contracts.
Experience of Gamma-Locator System Using for Radiation Monitoring During Rehabilitation Works at Temporal Radwaste Storage Area of Kurchatov Institute
Authors: Ivanov O. P. , Potapov V. N. , Ignatov S. M. , Smirnov S. V. , Stepanov V. E. , Volkovich A. G. , Volkov V. G.
Published: June 2009

Remote monitoring of radiological conditions on large areas is important task during large-scale activity with radioactive contamination/materials. We present results of application of an automatic system for remote measurements of radiological conditions at territory of rehabilitation activity. The system is scanning collimated spectrometric detector, its construction and main performance characteristics are shortly described. System, including two different measurement heads was used during three years for radiological monitoring of area rehabilitation works at temporal radwaste storage area of RRC Kurchatov Institute. The results of its application for exposure dose rate monitoring at particular control points from separate strong sources located in the territory of activity are presented and special cases are studies in details. The method of EDR calculation in 3-D space around territory of activity and analysis of relative input of main gamma-sources into EDR are presented. The results of application of the system during rehabilitation activity at territory of temporal storage of radioactive wastes of RRC Kurchatov Institute have shown high efficiency of this system for such operations.
Application of FEPs Analysis to Identify Research Priorities Relevant to the Safety Case for an Australian Radioactive Waste Facility
Authors: Payne T. E. , McGlinn P. J.
Published: June 2009

The Australian Nuclear Science and Technology Organisation (ANSTO) has established a project to undertake research relevant to the safety case for the proposed Australian radioactive waste facility. This facility will comprise a store for intermediate level radioactive waste, and either a store or a near-surface repository for low-level waste. In order to identify the research priorities for this project, a structured analysis of the features, events and processes (FEPs) relevant to the performance of the facility was undertaken. This analysis was based on the list of 137 FEPs developed by the IAEA project on ‘Safety Assessment Methodologies for Near Surface Disposal Facilities’ (ISAM). A number of key research issues were identified, and some factors which differ in significance for the store, compared to the repository concept, were highlighted. For example, FEPs related to long-term ground-water transport of radionuclides are considered to be of less significance for a store than a repository. On the other hand, structural damage from severe weather, accident or human interference is more likely for a store. The FEPs analysis has enabled the scientific research skills required for the interdisciplinary project team to be specified. The outcomes of the research will eventually be utilised in developing the design, and assessing the performance, of the future facility. It is anticipated that a more detailed application of the FEPs methodology will be undertaken to develop the safety case for the proposed radioactive waste management facility.
Modelling Approach to LILW-SL Repository Safety Evaluation for Different Waste Packing Options
Authors: Perko Janez , Mallants Dirk , Volckaert Geert , Towler George , Egan Mike , Virsˇek Sandi , Hertl Bojan
Published: June 2009

The key objective of the work described here was to support the identification of a preferred disposal concept and packaging option for low and short-lived intermediate level waste (LILW-SL). The emphasis of the assessment, conducted on behalf of the Slovenian radioactive waste management agency (ARAO), was the consideration of several waste treatment and packaging options in an attempt to identify optimised containment characteristics that would result in safe disposal, taking into account the cost-benefit of alternative safety measures. Waste streams for which alternative treatment and packaging solutions were developed and evaluated include decommissioning waste and NPP operational wastes, including drums with unconditioned ion exchange resins in overpacked tube type containers (TTCs). For decommissioning wastes, the disposal options under consideration were either direct disposal of loose pieces grouted into a vault or use of high integrity containers (HIC). In relation to operational wastes, three main options were foreseen. The first is overpacking of resin containing TTCs grouted into high integrity containers, the second option is complete treatment with hydration, neutralization, and cementation of the dry resins into drums grouted into high integrity containers and the third is direct disposal of TTCs into high integrity containers without additional treatment. The long-term safety of radioactive waste repositories is usually demonstrated with the support of a safety assessment. This normally includes modelling of radionuclide release from a multi-barrier near-surface or deep repository to the geosphere and biosphere. For the current work, performance assessment models were developed for each combination of siting option, repository design and waste packaging option. Modelling of releases from the engineered containment system (the ‘near-field’) was undertaken using the AMBER code [1]. Detailed unsaturated water flow modelling was undertaken using the HYDRUS code [2], where the degree of engineered barrier degradation with time is accounted for in each packaging option. Water fluxes relating to each degradation level were then incorporated into the AMBER models for further radionuclide transport calculations appropriate to each packing solution. The approach proved to be highly flexible, transparent and effective in terms of calculation time. Results demonstrate that all waste streams could be accepted at the preferred site with the surface repository option, under the condition that all decommissioning waste would be grouted into high integrity containers. The use of high integrity containers is also recommended for all other waste streams. Results from the detailed analysis further showed that in-drum-dried ion exchange resins in TTCs would be acceptable when grouted into high integrity containers, thereby avoiding the need for complicated processing and repackaging.
The Influence of Zeta Potential and Yield Stress on the Filtration Characteristics of a Magnesium Hydroxide Simulant
Authors: Biggs Simon , Nabi Rafiq , Poole Colin , Patel Ashok
Published: June 2009

In the UK, irradiated fuels from Magnox reactors are often stored in water-filled ponds under alkaline conditions, so as to minimise corrosion of fuel cladding. This is important to prevent or reduce leakage of soluble fission products and actinides to the pond water. A variety of intermediate level wastes derived from Magnox materials are stored at power stations. Under these alkaline conditions, various species of magnesium are formed, of which magnesium hydroxide is the dominant material. The particle-fluid interactions are significant for the design and operation of facilities for hydraulic retrieval, filtration, dewatering and ion exchange treatment of fuel storage pond water and stored wet Magnox wastes. Here we describe a study of particulate properties and filtration characteristics of oxide particle simulants under laboratory conditions. Cake and medium resistance data were correlated across a range of pH conditions with electroacoustic zeta potential and shear yield stress measurements, as a function of particle volume fractions. The influence of zeta potential on filtration properties arises directly from the interaction of particles within the sediment cake.
Developing a Fuzzy Rule Based Cognitive Map for Total System Safety Assessment
Authors: de Lemos Francisco Luiz , Sullivan Terry
Published: June 2009

Total System Performance Assessment, TSPA, for radioactive waste disposal is a multi and interdisciplinary task that is characterized by complex interactions between parameters and processes; lack of data; and ignorance regarding natural processes and conditions. The vagueness in the determination of ranges of values of parameters and identification of interacting processes pose further difficulties to the analysts with regard to the establishment of the relations between processes and parameters. More specifically the vagueness makes uncertainty propagation and sensitivity analysis challenging to analyze. To cope with these difficulties experts often use simplifications and linguistic terms to express their state of knowledge about a certain situation. For example, experts use terms such as “low pH”, “very unlikely”, etc to describe their perception about natural processes or conditions. In this work we propose the use of Fuzzy Cognitive Maps, FCM, for representation of interrelation between processes and parameters as well as to promote a better understanding of the system performance. Fuzzy cognitive maps are suited for the case where the causal relations are not clearly defined and, therefore, can not be represented by crisp values. In other words, instead of representing the quality of the interactions by crisp values, they are assigned degrees of truth. For example, we can assign values to the effect of one process on another such that (+) 1 corresponds to positive, (−) 1 to negative and 0 to neutral effects respectively. In this case the effect of a process A, on a process, B, can be depicted as function of the membership to the fuzzy set “causal effect” of the cause process to the target one. One of the main advantages of this methodology would be that it allows one to aggregate the linguistic expressions as descriptions of processes. For example, a process can be known to have a “very strong” positive effect on another one, or using fuzzy sets terminology the effect is “around (+) 1” with degree of membership μ = 0.9. As another example, a “moderate” negative effect can be represented as “around (−) 1” with degree of membership μ = 0.6 to the set “(−) 1”. Such a methodology can be an important tool for enhancing transparency in the TSPA process by allowing discussions between experts from different fields of research, for example by adding new “what if” analysis, and, therefore, for confidence building.
Evaluation of the Effects of Media Velocity on HEPA Filter Performance
Authors: Alderman Steven , Parsons Michael , Hogancamp Kristina , Norton O. Perry , Waggoner Charles
Published: June 2009

Section FC of the ASME AG-1 Code addresses glass fiber HEPA filters and restricts the media velocity to a maximum of 2.54 cm/s (5 ft/min). Advances in filter media technology allow glass fiber HEPA filters to function at significantly higher velocities and still achieve HEPA performance. However, diffusional capture of particles < 100 nm is reduced at higher media velocities due to shorter residence times within the media matrix. Therefore, it is unlikely that higher media velocities for HEPA filters will be allowed without data to demonstrate the effect of media velocity on removal of particles in the smaller size classes. In order to address this issue, static testing has been conducted to generate performance related data and a range of dynamic testing has provided data regarding filter lifetimes, loading characteristics, changes in filter efficiency and the most penetrating particle size over time. Testing was conducted using 31 cm × 31 cm × 29 cm deep pleat HEPA filters supplied from two manufacturers. Testing was conducted at media velocities ranging from 2.0–4.5 cm/s with a solid aerosol challenge composed of potassium chloride. Two set of media velocity data were obtained for each filter type. In one set of evaluations, the maximum aerosol challenge particle size was limited to 3 μm, while particles above 3 μm were not constrained in the second set. This provided for considerable variability in the challenge mass mean diameter and overall mass loading rate. Results of this testing will be provided to the ASME AG-1 FC Committee for consideration in future versions of the HEPA standard. In general, the initial filter efficiency decreased with increasing media velocity. However, initial filter efficiencies were generally good in all cases. Filter efficiency values averaged over the first ten minute of the loading cycle ranged from 99.970 to 99.996%. Additionally, the most penetrating particle size was observed to decrease with increasing media velocity, with initial values ranging from 194 to 134 nm.
Advances in Development of the Fission Product Extraction Process for the Separation of Cesium and Strontium From Spent Nuclear Fuel
Authors: Law Jack , Peterman Dean , Riddle Cathy , Meikrantz David , Todd Terry
Published: June 2009

The Fission Product Extraction (FPEX) Process is being developed as part of the United States Department of Energy Advanced Fuel Cycle Initiative for the simultaneous separation of cesium (Cs) and strontium (Sr) from spent light water reactor (LWR) fuel. Separation of the Cs and Sr will reduce the short-term heat load in a geological repository, and when combined with the separation of americium (Am) and curium (Cm), could increase the capacity of the geological repository by a factor of approximately 100. The FPEX process is based on two highly specific extractants: 4,4',(5')-Di-(t-butyldicyclo-hexano)-18-crown-6 (DtBuCH18C6) and Calix[4]arene-bis-(tertoctylbenzo-crown-6) (BOBCalixC6). The DtBuCH18C6 extractant is selective for strontium and the BOBCalixC6 extractant is selective for cesium. Results of flowsheet testing of the FPEX process with a simulated feed solution in 3.3-cm centrifugal contactors are detailed. Removal efficiencies, distribution coefficient data, coextraction of metals, and process hydrodynamic performance are discussed along with recommendations for future flowsheet testing with actual spent nuclear fuel.
Green Remediation: Enhanced Reductive Dechlorination Using Recycled Rinsewater as Bioremediation Substrate
Authors: Dawson Gaynor , McKeon Tom
Published: June 2009

Enhanced reductive dechlorination (ERD) has rapidly become a remedy of choice for use on chlorinated solvent contamination when site conditions allow. With this approach, solutions of an organic substrate are injected into the affected aquifer to stimulate biological growth and the resultant production of reducing conditions in the target zone. Under the reducing conditions, hydrogen is produced and ultimately replaces chlorine atoms on the contaminant molecule causing sequential dechlorination. Under suitable conditions the process continues until the parent hydrocarbon precursor is produced, such as the complete dechlorination of trichloroethylene (TCE) to ethene. The process is optimized by use of a substrate that maximizes hydrogen production per unit cost. When natural biota are not present to promote the desired degradation, inoculates can be added with the substrate. The in-situ method both reduces cost and accelerates cleanup. Successful applications have been extended from the most common chlorinated compounds perchloroethylene (PCE) and TCE and related products of degradation, to perchlorate, and even explosives such as RDX and trinitrotoluene on which nitrates are attacked in lieu of chloride. In recent work, the process has been further improved through use of beverage industry wastewaters that are available at little or no cost. With material cost removed from the equation, applications can maximize the substrate loading without significantly increasing total cost. The extra substrate loading both accelerates reaction rates and extends the period of time over which reducing conditions are maintained. In some cases, the presence of other organic matter in addition to simple sugars provides for longer performance times of individual injections, thereby working in a fashion similar to emulsified vegetable oil. The paper discusses results of applications at three different sites contaminated with chlorinated ethylenes. The applications have included wastewaters of both natural fruit juices and corn syrup solutions from carbonated beverages. Cost implications include both the reduced cost of substrate and the cost avoidance of needing to pay for treatment of the wastewater.
Evaluation of a Method for Designing Sealing Plugs for HLW Geological Disposal Facilities, Taking Into Account the Heterogeneous Characteristics of the Geological Environment
Authors: Suyama Yasuhiro , Yanagizawa Koichi , Toida Masaru
Published: June 2009

In order to ensure that a repository for the geological disposal of HLW is isolated from the human environment, underground excavations, including pits and tunnels, must be properly sealed. Effective sealing requires that these excavations are backfilled, and that the Excavation Damage or Disturbed Zone (EDZ), which includes preferential flowpaths, must be intersected by sealing plugs. Methods for constructing a full-scale sealing plug and their influence on plug performance were evaluated and confirmed by a Tunnel Sealing Experiment (TSX). This experiment was carried out by an international partnership of the Japan Nuclear Cycle Development Institute (JNC) and Atomic Energy of Canada Limited (AECL). However certain specific roles of the sealing plugs at the scale of the whole repository were not studied. There remain issues to be clarified, notably the effectiveness of sealing plugs in a geological environment with heterogeneous characteristics and the resulting influences of the heterogeneities in performance assessment. Focusing on a geological environment with spatially heterogeneous characteristics, the authors have developed a method for designing the sealing plugs, based on a concept of “primarily design for closure, secondarily design for construction”. Though the proposed method for designing sealing plugs has presently been developed only at a conceptual level, it indicates the possibility of establishing a repository even in a strongly heterogeneous geological environment that may have been considered previously to be inappropriate for a repository.
Preliminary Safety Analysis of the Baita Bihor Radioactive Waste Repository, Romania
Authors: Little Richard , Dragolici Felicia , Bond Alex , Matyasi Ludovic , Matyasi Sandor , Naum Mihaela , Niculae Ortenzia , Thorne Mike , Watson Sarah
Published: June 2009

A project funded under the European Commission’s Phare Programme 2002 has undertaken an in-depth analysis of the operational and post-closure safety of the Baita Bihor repository. The repository has accepted low- and some intermediate-level radioactive waste from industry, medical establishments and research activities since 1985 and the current estimate is that disposals might continue for around another 20 to 35 years. The analysis of the operational and post-closure safety of the Baita Bihor repository was carried out in two iterations, with the second iteration resulting in reduced uncertainties, largely as a result taking into account new information on the hydrology and hydrogeology of the area, collected as part of the project. Impacts were evaluated for the maximum potential inventory that might be available for disposal to Baita Bihor for a number of operational and post-closure scenarios and associated conceptual models. The results showed that calculated impacts were below the relevant regulatory criteria. In light of the assessment, a number of recommendations relating to repository operation, optimisation of repository engineering and waste disposals, and environmental monitoring were made.
The Belgian Approach and Status on the Radiological Surveillance of Radioactive Substances in Metal Scrap and Non-Radioactive Waste and the Financing of Orphan Sources
Authors: Braeckeveldt Marnix , De Preter Peter , Michiels Jan , Pepin Ste´phane , Schrauben Manfred , Wertelaers An
Published: June 2009

Numerous facilities in the non-nuclear sector in Belgium (e.g. in the non-radioactive waste processing and management sector and in the metal recycling sector) have been equipped with measuring ports for detecting radioactive substances. These measuring ports prevent radioactive sources or radioactive contamination from ending up in the material fluxes treated by the sectors concerned. They thus play an important part in the protection of the workers and the people living in the neighbourhood of the facilities, as well as in the protection of the population and the environment in general. In 2006, Belgium’s federal nuclear control agency (FANC/AFCN) drew up guidelines for the operators of non-nuclear facilities with a measuring port for detecting radioactive substances. These guidelines describe the steps to be followed by the operators when the port’s alarm goes off. Following the publication of the European guideline 2003/122/Euratom of 22 December 2003 on the control of high-activity sealed radioactive sources and orphan sources, a procedure has been drawn up by FANC/AFCN and ONDRAF/NIRAS, the Belgian National Agency for Radioactive Waste and Enriched Fissile Materials, to identify the responsible to cover the costs relating to the further management of detected sealed sources and if not found to declare the sealed source as an orphan source. In this latter case and from mid-2006 the insolvency fund managed by ONDRAF/NIRAS covers the cost of radioactive waste management. At the request of the Belgian government, a financing proposal for the management of unsealed orphan sources as radioactive waste was also established by FANC/AFCN and ONDRAF/NIRAS. This proposal applies the same approach as for sealed sources and thus the financing of unsealed orphan sources will also be covered by the insolvency fund.
Current Status of the Radioactive Waste Management Programme in Spain
Authors: Lang-Lenton Leo´n Jorge , Neri Emilio Garcia
Published: June 2009

Since 1984, ENRESA is responsible of the radioactive waste management and the decommissioning of nuclear installations in Spain. The major recent challenge has been the approval of the Sixth General Radioactive Waste Plan (GRWP) as “master plan” of the activities to be performed by ENRESA. Regarding the LILW programme, the El Cabril LILW disposal facility will be described highlighting the most relevant events especially focused on optimizing the existing capacity and the start-up of a purpose–built disposal area for VLLW. Concerning the HLW programme, two aspects may be distinguished in the direct management of spent fuel: temporary storage and long-term management. In this regards, a major challenge has been the decision adopted by the Spanish Government to set up a Interministerial Committee for the establishment of the criteria that must be met by the site of the Centralized Intermediate Storage (CTS) facility as the first and necessary step for the process. Also the developments of the long-term management programme will be presented in the frame of the ENRESA’s R&D programme. Finally, in the field of decommissioning they will be presented the PIMIC project at the CIEMAT centre and the activities in course for the decommissioning of Jose´ Cabrera NPP.
European Concepts for Shared Storage and Disposal Facilities for Radioactive Wastes?
Authors: Verhoef Ewoud , Code´e Hans , Sˇtefula Vladan , McCombie Charles
Published: June 2009

Geological disposal is an essential component of the long-term management of spent fuel and high-level radioactive waste. Implementation of a suitable deep repository may, however, be difficult or impossible in some (especially small) countries because of challenging geological conditions or restricted siting options, or because of the high costs involved. For these countries, shared regional or international storage and disposal facilities are a necessity. The European Parliament and the EC have both expressed support for concepts that could lead to regional shared facilities being implemented in the EU. The EC, therefore, funded two projects that form the first two steps of a staged process towards the implementation of shared regional or international storage and disposal facilities. In the period 2003 to 2005, the EC funded SAPIERR I, a project devoted to pilot studies on the feasibility of shared regional storage facilities and geological repositories, for use by European countries. The studies showed that shared regional repositories are feasible, but also that, if they are to be implemented, even some decades ahead, efforts must already be increased now. The first step would be to establish a structured framework for the work on regional repositories. This is the goal of SAPIERR II (2006–2008): to develop possible practical implementation strategies and organisational structures. These will enable a formalised, structured European Development Organisation (EDO) to be established in 2008 or afterwards for working on shared EU radioactive waste storage and disposal activities. The EDO can work in parallel with national waste programmes. Participating EU Member States will be able to use the structures developed as, when and if needed for the furtherance of their individual national policies.
On the Role of H2 as an Inhibitor of UO2 Matrix Dissolution
Authors: Merino Juan , Gaona Xavier , Duro Lara , Bruno Jordi , Marti´nez-Esparza Aurora
Published: June 2009

The study of spent fuel behaviour under disposal conditions is usually based on conservative approaches assuming oxidising conditions produced by water radiolysis at the fuel/water interface. However, the presence of H2 from container corrosion can inhibit the dissolution of the UO2 matrix and enhance its long-term stability. Several studies have confirmed the decrease in dissolution rates when H2 is present in the system, although the exact mechanisms of interaction have not been fully established. This paper deals with a radiolytic modelling exercise to explore the consequences of the interaction of H2 with radicals generated by radiolysis in the homogeneous phase. The main conclusion is that in all the modelled cases the presence of H2 in the system leads to a decrease in matrix dissolution. The extent of the inhibition, and the threshold partial pressure for the inhibition to take place, both depend in a complex way on the chemical composition of the water and the type of radiation present in the system.
Evaluation and Analysis of the Residual Radioactivity for the 15UD Pelletron Accelerator Facility
Authors: Sonkawade R. G.
Published: June 2009

For the assessment of radiological impact of the accelerators, it will be better to have the documented information on activation of metal parts of the accelerator components. It is very much essential to get reliable data on these subjects. During acceleration of light ion, the residual radioactivity in the accelerator facility was found near the Analyzing Magnet, single slit, Beam Profile Monitors (BPM), Faraday Cups (FC), bellows, beginning of switching magnet bellows, at the target and the ladder. Study with HPGE detector gives an insight of the formation of the short or long lived radionuclides. The different targets used in the light ion experiment were also monitored and proper decommissioning and decontamination steps were followed. This paper presents the data of residual radioactivity in the 15UD Pelletron accelerator infrastructure.