bluuebunny/crossref_metadata_2025 huggingface dataset:
The Measurement of Radioactive Surface Contamination Using an Inorganic Fluor-Impregnated Membrane
Authors: Seo Bum-Kyoung , Kim Sung-Kyun , Lee Kune-Woo , Park Jin-Ho , Lim Nan-Ju , Han Myeong-Jin
Published: March 2009

Solid scintillation proximity membranes were prepared for measuring surface in laboratories contaminated by the all types of radionuclides, such as H-3, C-14, and Cs-137 etc. Polysulfone scintillation proximity membranes were prepared by impregnating Cerium Activated Yttrium Silicate (CAYS),an inorganic fluor, in a membrane structure. The inorganic fluor-impregnated membranes were applied to detect the radioactive surface contamination directly without the aid of a scintillation cocktail. The preparation of membranes was divided into two processes. A supporting polymer film was made of casting solution consisting of polysulfone and solvent, their cast film being solidified by vacuum evaporation. CAYS-dispersed polymer solutions were cast over the first, solidified polymer films and coagulated either by evaporating solvent in the solution with non-solvent in a coagulation bath. The prepared membranes had two distinguished, but tightly attached, double layers: one is the supporting layer of dense polymer film and the other results revealed that the prepared membranes were eficient to monitor radioactive contamination with reliable counting ability. For enhancement of pick-up and measurement efficiency, the membrane was prepared with the condition of different membrane solidification. The scintillation produced by interaction with radiation and CAYS was measured with photomultiplier tube. The test results of the cocktail-free wipe test showed that the prepared membranes were efficient to monitor radionuclide-contaminated areas with the good counting ability as well as with the decrease of overall production of radioactive waste.
A Review of Power Fluidics for Nuclear Waste Mobilization
Authors: Williams Martin C. , Murray Paul A. , Stairmand John
Published: March 2009

Power Fluidics technology has been used for handling nuclear fluids for over 30 years. The key feature of the technology is the absence of mechanical moving parts in contact with the process fluid, so giving inherent plant safety and reliability. Power fluidic pumps use compressed gas to effect liquor flow, and are capable of pumping without flow dilution and without susceptibility to erosion or blockage. Applications for Power Fluidics technology were initially in the field of sludge pumping in new-build nuclear processing plants, and a number of designs of pumps capable of handling high-viscosity, high-SG materials were developed. More recently, Power Fluidics technology has been further developed for ‘retrofitting’ to achieve mobilization and retrieval of radioactive wastes and sludges from existing tanks and silos. This paper focuses on this application and will describe how Power Fluidics equipment has been developed and implemented to mix, sample and retrieve waste from existing storage tanks or silos. Case studies of several recent successful projects are included.
Radiation Protection Experience in RA Heavy Water Research Reactor Important for Decommissioning
Authors: Ninkovic Marko M.
Published: March 2009

The RA reactor has been in service since the end of 1959. The beginning of the eightieth using of the reactor was interrupted because of some irregular deposits on the fuel elements in the core. Since that time, for about 20 years, the reactor has not been in service and is now being prepared for decommissioning. There have been no serious accidents to the RA reactor during the operation period, but a few small accidents or incidents, which affected the reactor systems and higher exposure of personnel. These incidents were: contamination of the heavy water cooling system by 60Co; flooding of a few hundred liters of heavy water from the heavy water cooling system; the first and single case of fuel element failure in the reactor core. A specific accident called “phantom”; and finally, some irregular deposits on the fuel elements inside the reactor core that were registered. In this paper more detailed data is given on the above mentioned accidents or incidents, their consequences on the reactor systems and personnel and countermeasures taken for reducing danger radioactivity. According to the opinion of the author, these data and lessons learned would be very interesting in general, and especially, for radiological characterization, as one of the first steps in the process of reactor decommissioning.
Successful Team Working in Major Nuclear Decommissioning Project: Progress on Environmental Restoration at UKAEA Winfrith
Authors: Staples A. T. , Latinwo S. , Parkingson S. J. , Appleton J.
Published: March 2009

Following a successful contract relationship between UKAEA and RWE NUKEM Ltd. to provide operational and decommissioning services for Building A59 (a post-irradiation examination facility) at UKAEA Winfrith, additional major tasks under the WOMAD (Winfrith Operations, Maintenance, and Decommissioning) Project were competitively won by RWE NUKEM. The WOMAD Project advances implementation of the overall environmental remediation mission for the UKAEA Winfrith site. This paper describes the four main parts of the WOMAD Project at Winfrith that are being conducted by RWE NUKEM. The WOMAD Project consists of the following tasks: • Care and maintenance of the two large nuclear reactors SGHWR and DRAGON. • Operation of Building A59 for 1 year followed by decommissioning and demolition. • Conditioning and immobilization of ∼350,000 litres of radioactive sludge in 500-litre drums followed by demolition of the sludge storage tanks. • Refurbishment of a storage facility and emplacement of the drummed waste. These tasks are performed by RWE NUKEM and the overall project is managed by UKAEA within the constraints imposed by the site licensee. The work is being implemented via a contract with RWE NUKEM using the New Engineering Contracts (NEC) suite of contracts (Options A and C) as established by the Institution of Civil Engineers. This framework was chosen because it facilitates good project management practices and encourages a team approach. This strategy has been taken to heart within the project and many joint initiatives are being used to develop an effective team that is working toward a common goal. In the relatively short time since contract award (March 2000), decommissioning in Building A59 is currently ahead of schedule and a number of technical hurdles have been overcome. The new facility for processing the external active sludge tank (EAST) wastes has been designed, planning permission has been obtained, and construction has begun. The first phase of safety submittals was also completed, and a Stage 2 Letter of Comfort was obtained from UK Nirex following a series of small-scale active development tasks. This presentation will feature current progress regarding schedule, cost, and quality aspects and will be an update on the position given to the IMechE Decommissioning Conference in 2001 [1] and the IBC Decommissioning Conference in 2002. [2]
Processes for the Treatment of NORM and TENORM
Authors: Vesely Andreas
Published: March 2009

By contract with the Austrian government, the ARC is treating radioactive waste from research institutions and industries. In the last years, one focus was the development of processes for the treatment of NORM and TENORM. Our goal in developing such processes is to recycle valuable compounds for further industrial usage and to concentrate the radioactive elements as far as possible, to save space in the waste storage facilities. Austria is an important producer of tungsten-thoria- and tungsten-molybdenum-thoria-cermets. Scrap is generated during the production process in the form of turnings and grinding sludge and dust. Although big efforts have been undertaken to replace Thorium compounds, waste streams from past production processes are still waiting for treatment. The total amount of this waste stored in Austria may be estimated to be approx. 100 tons. In close co-operation with the tungsten industries, recycling processes were tested and further developed at ARC in laboratory, bench scale and pilot plants. Three different approaches to solve the problem were studied: Dissolution of tungsten in molten iron in an arc or induction furnace, thus producing an Fe-W or Fe-W-Mo alloy. Slag is produced upon the addition of lime and clay. This slag extracts nearly all of the Thorium contained in the metal melt. Selective dissolution of Tungsten in aqueous alkaline medium after oxidation of the metal to the hexavalent state by heating the scrap in air at temperatures of 500°C to 600°C. The resulting oxides are treated with sodium hydroxide solution. Tungsten and Molybdenum oxides are readily dissolved, while Thorium oxide together with silicon and aluminum compounds remain insoluble and are separated by filtration. Sodium tungstate solution is further processed by the usual hydrometallurgical tungsten mill process. Oxidation and dissolution of Tungsten can be achieved in one step by an electrochemical process. Thus, thoriated Tungsten scrap is used as an anode in an electrolysis cell, while sodium hydroxide or ammonia serve as electrolyte. After dissolution of Tungsten, the solids are separated from the liquid by filtration. With the electrochemical process, treatment of Tungsten-Thoria scrap can be achieved with high throughput in rather small reactors at moderate temperatures and ordinary pressure. The Tungsten solution exhibits high purity. Another process which we examined in detail is the separation of radium from rare earth compounds. Radium was separated by co-precipitation with barium sulfate from rare earth chloride solutions. The efficiency of the separation is strongly pH-dependent. Again, the valuable rare earth compound can be reused, and the radioactive elements are concentrated.
Practical Implementation of National Clearance Levels at Dounreay
Authors: McClelland Paul , Dennis Frank , Liddiard Mark
Published: March 2009

Clearance is a very important part of any effective waste management strategy for both operating and decommissioning nuclear facilities. Radioactive waste disposal capacity is becoming an increasingly valuable resource and costs for disposal of radioactive wastes continue to dramatically rise. Considerable cost savings may be realised by efficient segregation of essentially non-radioactive material from radioactive wastes. The release of these materials from licensed nuclear sites for disposal, reuse or recycle without further regulatory controls is commonly referred to by the nuclear industry as “clearance”. Although much effort has been directed at establishing national clearance levels, below which, materials may be released without further regulatory controls, there is little practical guidance regarding implementation into local waste management programmes. Compliance with regulatory clearance limits is a relatively straightforward technical exercise involving appropriate management control and monitoring of the material. Whilst this is sufficient to avoid prosecution for breach of regulatory requirements, it is not sufficient to avoid a myriad of political and public relations land mines. When material is unconditionally released, unless additional attention is given to management of its future destination off-site, it may end up anywhere. The worst nightmare for a waste manager at a nuclear site is headlines in local and national newspapers such as, “RADIOACTIVE WASTE DISPOSED IN LOCAL MUNICIPAL LANDFILL,” or, “RADIOACTIVE WASTE USED AS CONSTRUCTION MATERIAL FOR CHILDRENS PLAYGROUND,” etc. Even if the material were released legally, the cost of recovering from such a situation is potentially very large, and such public relations disasters could threaten to end the clearance programme at the given site, if not nationally. This paper describes how national regulatory clearance levels have been implemented for the decommissioning of the Dounreay nuclear site in the far north of Scotland. It specifically focuses on the management of public relations aspects of clearance in order to limit the exposure to non-regulatory pressures and liabilities associated with clearance programmes from nuclear sites. The issues are put into context for uncontaminated wastes, trace contaminated wastes and management of contaminated land.
Public Involvement in Nevada: The Role of the University
Authors: Hechanova Anthony E.
Published: March 2009

The controversy over siting a high-level radioactive waste repository at Yucca Mountain, Nevada, has engaged the southern Nevada community in recent years. This high-profile project has brought a critical eye on other environmental management activities, especially in the neighboring Nevada Test Site. Although radioactive waste management activities have been ongoing at the Nevada Test Site for over 40 years, there is renewed scrutiny on these activities and there is a growing community interest in pursuing alternative waste management strategies. This paper reviews the active role that the University of Nevada, Las Vegas has played in promoting public awareness, understanding, and participation.
Status of the Salt Waste Processing Facility at the Savannah River Site
Authors: Spears Terrel J. , McCullough James W. , Harmon Harry D. , Leugemors Robert K.
Published: March 2009

The Department of Energy’s (DOE) Savannah River Site (SRS) High-Level Waste (HLW) Program is responsible for storage, treatment, and immobilization of HLW for disposal. The Salt Processing Program (SPP) is the salt (soluble) waste treatment portion of this effort. The overall SPP encompasses the selection, design, construction and operation of treatment technologies to prepare the salt waste feed material for the site’s Saltstone Facility and vitrification facility (Defense Waste Processing Facility). Major constituents that must be removed from the salt waste include actinides, strontium, cesium, and entrained sludge. In fiscal year (FY) 2002, research and development (R&D) on the actinide and strontium removal and Caustic-Side Solvent Extraction (CSSX) processes transitioned from technology development for baseline process selection to providing input for conceptual design of the Salt Waste Processing Facility (SWPF), a key component at the SRS SPP. This work included laboratory studies, bench-scale tests, and prototype equipment development. To implement the salt waste treatment technologies, DOE initiated a competitive procurement process to select Engineering, Procurement, and Construction (EPC) contractors for design of the SWPF. The Department awarded EPC contracts to Parsons Infrastructure & Technology Group, Inc. and Foster Wheeler USA Corporation for preparation of conceptual designs (Phase I) for the SWPF. The two EPC contractors began conceptual design activities in September 2002 and are scheduled to complete this work in January 2004. After evaluation of the conceptual designs, DOE will down select one EPC contractor to continue with final design, construction, and hot commissioning (Phase II). Hot startup of the SWPF is targeted for December 2009.
Using Engineering, Administrative and Personal Protective Equipment Controls to Remediate Hazardous and Radioactive Constituents
Authors: Hylko James M.
Published: March 2009

This paper documents how utilizing available source term information, integrated safety management, and associated engineering, administrative and personal protective equipment (PPE) controls are used in concert to perform work safely. Two field projects consisting of 1) a room containing both hazardous (e.g., hydrofluoric acid) and radioactive constituents and 2) a former reaction vessel containing approximately 568 liters (150 gallons) of lime sludge and technetium-99 (Tc-99) were organized using the Department of Energy’s (DOE’s) Integrated Safety Management System (ISMS). This system allowed the project teams to control work-related decisions based on their knowledge, experience, expertise, and field observations. The information and experience gained from each project stage and rehearsals contributed to modifying subsequent entries, further emphasizing the importance of developing hold points and incorporating lessons learned. Furthermore, selecting the appropriate PPE is based on providing an adequate level of employee protection relative to the task-specific conditions and hazards. PPE is categorized into four ensembles based on the degree of protection afforded, e.g., Levels A (most restrictive), B, C, and D (least restrictive). What is often overlooked in preparing an ensemble is that the PPE itself can create significant worker hazards, i.e., the greater the level of PPE, the greater the associated risks. Furthermore, there is confusion as to whether a more “conservative approach” should always be taken since Level B provides the same level of respiratory protection as Level A but less skin protection. Additional information summarizes the Occupational Safety and Health Administration regulations addressing Level A versus Level B, and provides justification for selecting Level B over Level A without under-protecting the employee. The hazards and the chemical nature of hydrofluoric acid provide qualitative evidence to justify Level A. Once hydrofluoric acid is removed as a source term constituent, PPE performance is evaluated against the remaining chemical inventory. If chemical breakthrough from direct contact is not expected to occur and instrument readings confirm the absence of any hazardous vapors, additional skin protection afforded by wearing a vapor-tight, totally encapsulated suit is not required. Therefore, PPE performance and instrument data provide quantitative evidence to justify Level B. These projects exemplify that using guidance provided by DOE’s ISMS and the Occupational Safety and Health Administration (OSHA) demonstrates how a detailed and thorough planning process integrating safe work practices and commitment to teamwork can result in the safe and effective completion of very complex and highly hazardous projects.
An Interactive Database Tool for Applying Integrated Safety Management in a Streamlined and Consistent Manner
Authors: Potts T. Todd , Hylko James M. , Douglas Terrence A.
Published: March 2009

WESKEM, LLC’s Environmental, Safety and Health (ES&H) Department had previously assessed that a lack of consistency, poor communication and use of antiquated communication tools could result in varying operating practices, as well as a failure to capture and disseminate appropriate Integrated Safety Management (ISM) information. To address these issues, the ES&H Department established an Activity Hazard Review (AHR)/Activity Hazard Analysis (AHA) process for systematically identifying, assessing, and controlling hazards associated with project work activities during work planning and execution. Depending on the scope of a project, information from field walkdowns and table-top meetings are collected on an AHR form. The AHA then documents the potential failure and consequence scenarios for a particular hazard. Also, the AHA recommends whether the type of mitigation appears appropriate or whether additional controls should be implemented. Since the application is web based, the information is captured into a single system and organized according to the ≥200 work activities already recorded in the database. Using the streamlined AHA method improved cycle time from over four hours to an average of one hour, allowing more time to analyze unique hazards and develop appropriate controls. Also, the enhanced configuration control created a readily available AHA library to research and utilize along with standardizing hazard analysis and control selection across four separate work sites located in Kentucky and Tennessee. The AHR/AHA system provides an applied example of how the ISM concept evolved into a standardized field-deployed tool yieling considerable efficiency gains in project planning and resource utilization. Employee safety is preserved through detailed planning that now requires only a portion of the time previously necessary. The available resources can then be applied to implementing appropriate engineering, administrative and personal protective equipment controls in the field.
A Very High Efficiency Neutron Counter for the Measurement of Plutonium in Decommissioning Wastes
Authors: Tolchard A. C. , Looman M. R. , Mason J. A.
Published: March 2009

The design of the ANTECH Model 2203 Very High Efficiency Neutron Counter (V-HENC) is a natural progression from the well-proven ANTECH Series 2200 Passive Neutron Drum Monitor used for measuring plutonium in intermediate and low level waste (LLW) 200 litre drums. ANTECH has had considerable experience in the implementation of the base design, originally licensed to ANTECH from the Joint Research Centre at Ispra, Italy. Three Series 2200 systems have been supplied by ANTECH: two are in operation at AWE Aldermaston for waste monitoring, and a third is implemented at the SMP facility at BNFL Sellafield. Some 15 years cumulative operating experience has been gained, and ANTECH provides technical support as part of continuing maintenance and support arrangements. The design has proven to be inherently reliable, safe, easy to operate and maintain. Recently, both AWE instruments have been fully characterised and calibrated to function in conventional coincidence counting mode with calibrations. ANTECH has used MCNP to optimise the design of the fast detector packages in order to achieve the lower detection levels required to measure Pu at USA TRU/LLW and UK Nirex LLW levels. With the aid of simulations a typical detection efficiency of 36% with Cd filters deployed and up to 45% with the internal Cd liner removed has been achieved. Statistical data filtering is used to decrease the cosmic ray induced neutron background, the latter also being minimised by the absence of steelwork within the drum measurement chamber. An outer shielding of 270 mm thickness of polyethylene is used to shield the system from external neutrons. A total of 16 fast detector modules are used, which consist of eight, 7.5 atmospheres 3He tubes (25.4mm diameter and 1033 mm active length) embedded in high density polyethylene and arranged in a double row. Tubes are connected using HN connectors to a junction box at the top of the vertical modules or on the end of the horizontal modules. The junction boxes are hermetically sealed and contain the high voltage distribution and AMPTEK model A-111 charge sensitive amplifiers. The operation of the V-HENC is based on passive neutron counting of the correlated neutrons from spontaneous fission of the even Pu nuclides, principally 240Pu and is coupled with an ANTECH/Ortec Advanced Multiplicity Shift Register employing the Los Alamos INCC code. Alternatively, the multiple gate ANTECH Time Correlation Analyser (TCA) may be used for enhanced data acquisition for multiplicity counting. The V-HENC can be operated in conventional shift register coincidence counting (Reals) mode (with a calibration function), the absolute multiplicity counting mode (histogram function) or totals counting mode. Plant measured isotopic ratios can be used by the software to convert 240Pueffective mass to total Pu mass.
Update on the Status of the West Valley Demonstration Project
Authors: Orlando Dominick , Glenn Chad , Bradford Anna , Craig Claudia
Published: March 2009

From 1966 to1972, under an Atomic Energy Commission (AEC) license, Nuclear Fuel Services (NFS) reprocessed 640 metric tons of spent fuel at its West Valley, New York, facility, the only commercial spent fuel reprocessing plant in the U.S. The facility shut down in 1972, for modifications to increase its seismic stability and to expand its capacity. In 1976, without restarting the operation, NFS withdrew from the reprocessing business and returned control of the facilities to the site owner, the New York State Energy Research and Development Authority (NYSERDA). The reprocessing activities resulted in about 2.3 million liters (600,000 gallons) of liquid high-level waste (HLW) stored below ground in tanks, other radioactive wastes, and residual radioactive contamination. The West Valley site was licensed by AEC, and then the U.S. Nuclear Regulatory Commission (NRC), until 1981, when the license was suspended to execute the 1980 West Valley Demonstration Project (WVDP) Act. The WVDP Act outlines the responsibilities of the U.S. Department of Energy (DOE), NRC, and NYSERDA at the site, including the NRC’s responsibility to develop decommissioning criteria for the site. The Commission published the final policy statement on decommissioning criteria for the WVDP at the West Valley site after considering comments from interested stakeholders. In that regard, the Commission prescribed the License Termination Rule (LTR) criteria for the WVDP at the West Valley site, reflecting the fact that the applicable decommissioning goal for the entire NRC-licensed site is compliance with the requirements of the LTR. This paper will describe the history of the site, provide an update of the status of the decommissioning of the site and an overview of the technical and policy issues facing Federal and State regulators and other stakeholders as they strive to complete the remediation of the site.
Application of Mechanical Activation for Production of Zirconate Pyrochlore Ceramic Containing Simulated Rare-Earth: Actinide Fraction of HLW
Authors: Stefanovsky S. V. , Chizhevskaya S. V. , Kirjanova O. I. , Mikhailenko N. S. , Nikonov B. S.
Published: March 2009

Samples of zirconate pyrochlore ceramic (Ln)2(Zr,U)2O7 (Ln - lanthanides) containing simulated REE-An fraction of HLW were synthesized by conventional cold pressing of oxide mixtures in pellets under pressure of 200 MPa and sintering of pellets at 1550 °C for 24 hours and using preliminary mechanical activation of oxide powders for 5 or 10 min with a reduced sintering time for 6 hours. One more sample was produced by plasma melting. All the ceramic samples sintered at 1550 °C were monolithic and dense with high mechanically activated powders had higher chemical durability and mechanical properties.
Port Hope Area Initiative
Authors: Brown Peter , McCauley David
Published: March 2009

The Port Hope Area Initiative involves a process that will lead to the cleanup of low-level radioactive wastes in two communities in Southern Ontario and the construction of three new long-term waste management facilities in those communities. The history of the Initiative provides important insights into local participation and the successes and failures of siting efforts. The wastes resulted from the operations of an industrial process in Port Hope that began in the 1930s. Initially, wastes (contaminated with radium, uranium, and arsenic) from radium processing were deposited in a relatively uncontrolled manner at various locations within the town. By the 1940s, uranium processing wastes were deposited at nearby purpose-built radioactive waste management facilities. The problem of contamination was first recognized in 1974 and the worst cases quickly cleaned up. However, large volumes of contamination remained in the community. There were three successive efforts to develop an approach to deal with the area’s contamination. In the early to mid 1980s, a standard approach was employed; i.e. indentifying the most technically appropriate local site for a disposal facility, proceeding to evaluate that site, and communicating the benefits of the chosen approach to the local community. That approach was resoundingly rejected by local citizens and government representatives. The second effort, an innovative and consultative voluntary siting effort carried out during the late-1980s and early to mid-1990s involved the solicitation of other municipalities to volunteer to host a facility for the disposal of the Port Hope areas wastes. That effort resulted in the identification of a single volunteer community. However, negotiations between the federal government and the municipality were unable to reach an acceptable agreement establishing the conditions for the community to host the waste management facility. The third effort, a community-driven approach, was undertaken in the late-1990s and resulted in an agreement in 2001 between the Government of Canada and the local communities that sets in motion a process for the cleanup of the local wastes and long-term management in new local waste management facilities. This paper provides insights into the history of the problem, the efforts of the federal government over the last two decades to deal with the issue, how local participation and decision-making processes affected the successes of the various siting approaches, and lessons learned that might be of interest to others who must deal with environmental remediation situations that involve siting long-term management facilities.
Application of a Simplified Approach to GE Detector System Calibration
Authors: Leicht R. , Simon G. G.
Published: March 2009

Basic tasks in all monitoring projects involving gamma spectroscopy are design optimization and calibration of the Ge gamma spectrometer system which in general consists of one or more Ge detectors, the corresponding collimators, and the object with radioactive material (container, drum, wall, material on a conveyor etc.). A simplified approach to these problems has been developed by RWE NUKEM GmbH, leading to the software tool “ISO-Mod” which provides a straightforward comprehensive approach for the assessment and analysis of the most common cases without any need for other resources. The detector response is measured with calibration sources as a function of gamma energy, distance and polar angle. The effect of collimators is evaluated for rotational symmetry and for rectangular design based on polygon cross sections, with the possibility to have one or more layers of specific materials. The object model allows considering cubic (e. g. box, material on conveyor) or cylindrical (e. g. drum) absorption matrix and activity distributions with additional optional absorption layers. For considering the absorption in the object matrix and in the collimator material, mass attenuation coefficients are available for a wide range of energy. A number of selected gamma energies and matrix densities can be evaluated in a single evaluation run using Point Kernel and Monte Carlo simulation methods. For selected energies and densities, ISO-Mod provides the probability distribution of the calibration data. This allows investigating the sensitivity to potentially non-homogeneous activity distributions, e. g. to assess the improvement achieved when using more than one detector since up to 4 detectors can be considered simultaneously. All design data and results are — hierarchically structured — stored in a particular database. ISO-Mod is currently being used in different monitoring projects involving conveyor belts, containers and drums with radioactive material. Good agreement was found between the calculations and the results of measurements using calibrated point and homogeneous extended volume sources. Particular procedures are applied for locating hot spots in waste packages with multi-detector systems.
TRIGA Soil Decontamination by Electrokinetic Method
Authors: Kim Gye-Nam , Oh Won-Zin , Won Hui-Jun , Kim Min-Gil
Published: March 2009

The experiment for removing cesium and cobalt from the contaminated soil around the TRIGA reactor in Korea using the electrokinetic method were conducted. To increase removal efficiency by the electrokinetic method, a device to restrain the pH increase in the soil column was suggested. An acetate buffer solution (CH3COONa+CH3COOH) was injected into the soil column and acetic acid periodically was injected into the cathode reservoir to restrain any pH increase. Many 137Cs and 60Co were transferred by electromigration rather than electroomosis during the intial remediation period, and no precipitate was formed in the soil column. 25% of cesium radioactivity in the soil column was removed after 10 days, while 94% of cobalt radioactivity was removed. Furthermore, the remaining radioactive concentrations predicted by the developed numerical model were similar to those obtained by experiment.
Experiences of Dealing With Environmental Statements for Nuclear Reactor Decommissioning Projects Under the EIA Directive
Authors: McCready-Shea S. , Taylor F. E. , Batt J.
Published: March 2009

European Council Directive 85/337/EEC, as amended by Council Directive 97/11/EC, sets out a framework for the assessment of the effects of certain public and private projects on the environment. It is known as the Environmental Impact Assessment (EIA) Directive. The Directive is implemented in Great Britain (GB) for the dismantling or decommissioning of nuclear power stations and other nuclear reactor by the Nuclear Reactors (Environmental Impact Assessment for Decommissioning) Regulations 1999 (EIADR99). The Health and Safety Executive (HSE) is the competent authority for EIADR99 in GB, and has carried out public consultations on environmental statements that accompanied applications for consent to carry out decommissioning projects at two nuclear power stations in GB. HSE understands that these applications for consent are some of the first under the revised EIA Directive. HSE has developed a strategy for managing applications for consents under EIADR99. This strategy covers two main areas. The first area is public involvement, including identifying a large number of organisations in addition to the consultation bodies identified in the Regulations, providing information through the internet, and making responses to the consultation process publicly available. The second area is interfaces with legislation and Government policy, including town and country planning legislation, related health, safety and environment legislation, and decommissioning timetables. Experiences of implementing the strategy to deal with the environmental statements are described.
Experimental Research of Am-241 Migration From L/ILW in Soils: Transport Parameters
Authors: Arsene Carmen , Negoiu Dumitru
Published: March 2009

A programme on the safety analysis linked to the geological disposal of radioactive wastes is under study at the Institute for Nuclear Research Pitesti. In relation to the migration of radionuclides in the terrestrial environment following a possible release from the repository, specific experimental studies have been developed in order to provide the necessary input data for the risk assessment models. The paper reports the results of studies performed with 241Am. Laboratory experiments were conducted to determine the transport parameters of 241Am in the geological formations of Saligny site — the potential near-surface repository for low and intermediate radioactive wastes in Romania.
A Temporary Store for Radioactive Waste
Authors: Thomson Malcolm , Sims John
Published: March 2009

To provide an efficient process for the final disposal of radioactive wastes, some nuclear sites are currently building suitable, temporary stores for such material. This paper describes the practical construction programme recently completed at a UK power reactor site under decommissioning. This site has a number of half height ISO containers filled with solid radioactive material that needed a temporary storage area before being consigned for final disposal. As a result, the site, Hunterston A Decommissioning Site, needed to construct some temporary radiation shield walls to contain the half height ISO containers (HHISO) and safeguard staff and public during the storage period. The chosen location of the temporary store was the Charge Machine Maintenance Building (CMMB), which was surplus to requirements during the decommissioning programme, providing a weatherproof environment. Following an in-depth study of the different forms of construction available, the commercially available, hollow, interlocking polyethylene block system was adopted. Because of its modular form, it has the advantage that it can to be dismantled and re-used at other locations, if required. Being hollow, the block could be filled with a variety of materials, such as water, sand, lead or iron shot, depending on the radioactive shielding needs. An important aspect of the modular plastic block was that it could be more easily installed, decontaminated and decommissioned than the cheaper concrete block system, which produces copious quantities of waste for disposal. This paper describes the choice of interlocking block shielding after comparisons with more conventional forms of construction, the features of the HHISO store design, the short duration building programme and the on-site experiences of the construction period.
The Properties of Zeolite A in Molten LiCl Waste
Authors: Kim J. G. , Lee J. H. , Park G. I. , Lee S. H. , Kim J. H.
Published: March 2009

An advanced spent fuel management process based on the Li reduction of oxide spent fuel to a metallic form, under development at Korea Atomic Energy Research Institute (KAERI), will generate a LiCl waste. To develop an immobilization process of the LiCl waste salt using zeolite A, the behavior of zeolite A in molten LiCl salt with CsCl and SrCl2 has been investigated at 923 K. All the types of zeolite A’s have been transformed into zeolite Li-A after contact with molten LiCl salt. The sodalite phase has also been formed for the high LiCl/zeolite ratio, and this sodalite formation has been enhanced by an addition of CsCl or SrCl2. These results seem to be caused by the high temperature and Li concentration during the occlusion step. Ion-exchange properties of Cs and Sr at 923 K were very similar to those of the LiCl-KCl eutectic salt at 725 K: an uptake of Cs reached an equilibrium state very rapidly and kept constant, while that of Sr reached a maximum and decreased slightly.
Laser-Based Cutting of Contaminated Concrete
Authors: Crouse Philip L. , Li Lin , Spencer Julian T.
Published: March 2009

A number of laser-based techniques applicable to nuclear decontamination and decommissioning have emerged from LPRC-UMIST and BNFL research projects. The most recent process is deep-section concrete cutting, which promises to yield cutting depths of 1 m and beyond, and which affords very easy waste containment. This paper reports work in deep-section concrete cutting effected by layer-by-layer laser melting of the concrete with mechanical removal of the vitrified dross between passes. It is particularly suited to radioactive substrates where containment of the waste stream is imperative. Slightly higher power densities than e.g. for laser scabbling are required for this application. A comparison of experimental results using high-power CO2 (10.6 μm) and diode (0.808 + 0.940 μm ) lasers under roughly equivalent experimental conditions, cutting to depths of >100 mm, is presented. A marked improvement in cutting depth per pass is observed for the case of the diode laser. The increased cutting rate is rationalised in terms of the combined effects of coupling efficiency and beam shape. Pertinent features of the effluent are discussed.
Test Sorption on Concrete Samples
Authors: Olteanu Mirela , Bucur Crina
Published: March 2009

The near-field disposal facility will be design to provide effective containment for the radionuclide in the waste. Its design is based on a concept of multi-barrier containment, which makes use of physical (by the steel and concrete packaging) and chemical containment (based on development of uniform, beneficial aqueous condition and the provision of a high sorption capacity across the repository) within the cell of repository and of geological isolation. The achieve of this paper is to evaluate the chemical containment process (sorption capacity of backfill, distribution coefficient) of the potential contaminants which can be releases from radioactive waste conditioned form disposed in this repository, because of the dissolution/leaching processes.
Conditioning Experiments of Organic Solvents Solutions Contaminated With Tritium
Authors: Dianu Magdalena , Deneanu Nicolela
Published: March 2009

The method consists in the conditioning of organic solvents solutions contaminated with tritium (acetone, ethylene alcohol, methanol, trichloromethane, toluene, white spirit) into a cement and proper solidification additives matrix. Portland cement is the most common type of hydraulic cement and is the original agent utilized for the solidification of low-level radioactive waste. The paper contents: • the characterization of conditioning matrix components: waste, cement, additives; • preliminary laboratory tests on waste, cement and additives; • mechanical resistance verifications of conditioning matrix (results, tables); • leaching tests for tritium (results, graphs). This method is very simple, with low cost of working materials and long shelf life of working materials. The operator exposure is negligible because there is not any vapor problem. The values of mechanical resistance are higher than the disposal minimal value. Rate of release of tritium from the waste form as a result of interaction with water is very low. The higher values of mechanical resistant and the low rate of tritium released from the waste form offer security for transportation and final disposal of the waste placed in cement.
Low-Level Waste Disposition in D&D: A Systems Engineered Approach
Authors: Prince Robert E. , Ferrigno Dennis P. , Kirshe Mark
Published: March 2009

When approaching the decontamination and decommissioning (D&D) of a nuclear facility, the key elements of the project are the characterization, dismantlement execution, and the waste management (assay, packaging, transportation, and disposal) tasks. Waste management generally accounts for 40 to 45% of the total cost of the decommissioning project. This paper addresses an effective approach, based on demonstrated experience, to manage the dispostion alternatives for the decommissioning of waste materials. A systems engineered approach will minimize waste production and maximize efficiencies in the decommissioning schedule and offsite waste disposal. A key factor in this approach is to understand the disposal options and the staging criteria permitted within the specific applicable regulatory environment governing the decommissioning. For research reactor, nuclear power reactor, and other nuclear facilities involving complex systems, structures, and components, the key criteria and methodology of this systematic approach are applicable. Duratek has decommissioned numerous power reactor, research reactor, and industrial production facilities involving numerous complex systems and components. This paper focuses on the systems engineered approach, including applicable standards, waste streams, disposition alternatives, waste management methodologies, and lessons learned, for such decommissioning projects.
Conceptual and Practical Aspects of the Rehabilitation of Chernobyl NPP Exclusion Zone
Authors: Ivanov Yuri A. , Los Ivan P. , Arkhipov Andrey N. , Proskura Nickolay I.
Published: March 2009

The problems considered for the substantiation and implementation of a complex rehabilitation territory of the exclusion zone and the zone of absolute resettlement (EZZAR) are directed towards the recovery of its normal economic operation. This process of implementation uses a system of countermeasures. The issues of discussion are normative-legal base, principles, criterion, methods and scenarios of rehabilitation, concept of a full and partial rehabilitation, direction of rehab activity (both without change of the territory status and directed towards the rerun of territories under jurisdiction of local government bodies) and re-evacuation of the population and maintenance of normal conditions of its residents and habitability. State of the art natural and technogenic objects on the territory of EZZAR are discussed concerning preliminary estimations for a capability of practical implementation of different directions of rehabilitation activity on the EZZAR territory. GIS technologies are used to take into account the dynamics of a radioecological situation in natural and semi-natural ecosystems. The chief types of practical activities considered are activities on environmental protection and maintenance in a normal condition of phyto- and zoo-sanitary situation (including forestry activity, activation of landscape-recovery processes, etc.).
Advanced 3D Modeling and Simulation Techniques for Decommissioning of A-1 NPP
Authors: Voza´r Pavol , Sleza´k Vladimi´r , Krava´rik Kamil
Published: March 2009

This paper deals with advanced 3D computer-aided technologies used for modelling and simulation for decommissioning purposes. Within the A-1 NPP decommissioning process a set of activities is needed to perform successful dismantling and decontamination of rooms and equipment. Optimal process of performance of D&D of underground storage tanks and auxiliary rooms were used on the base of simulation outputs. The mockup tests were performed before using remotely controlled manipulators. The human presence during decontamination and dismantling is case by case excluded due to the radiation safety and ALARA approach. Within Bohunice A-1 Decommissioning Project an advanced computer-aided technologies were/are developed and used. Modelling software packages EUCLID and 3Dipsos together with 3D-laser scanner SOISIC are used for creating of 3D models and also for the verification of as-built state of selected systems and facilities. Software IGRIP is used for computer simulations of all D&D tasks. The 3D modelling and simulation of selected rooms and technological equipment of the A-1 NPP are used consequently in the process of decommissioning preparation and implementation. 3D modelling for the verification and simulation of operating steps is presented in the paper and its contribution to avoiding of collisions and non-optimal interventions into the building and technological parts during performing particular works is evaluated. The application of 3D models for the verification and simulation of operating steps significantly contribute to the optimal planning of D&D procedures. Minimisation of occupation doses of realisation personnel is main reason why the 3D modelling and simulations are used. The paper also presented 3D models of rooms chosen to simulate specific operations (decontamination, handling of radioactive wastes and/or dismantling by remote controlled manipulators) without risk accident, high dose rates of personnel etc. Process of selection of optimal operating procedure for decontamination and dismantling is presented as well as achieved experiences and recommendations for further work.
Electrochemical Destruction of Waste Lubricants Using Diamond-Coated Electrodes
Authors: Bates C. J. , Taylor G. T.
Published: March 2009

Electrochemical oxidation with boron-doped, diamond-coated electrodes could potentially be applied to treat radioactively contaminated oils. Accordingly, a feasibility study was performed using two representative, non-radioactive oils (a cutting fluid and a hydraulic oil). The process was shown to remove between 92 and 99% of the organic material in an alkaline suspension of the cutting fluid, with 87% of the added organic carbon being recovered as carbonate. A further test demonstrated that the oxidation could be maintained during periodic addition of oil to the electrochemical cell. It was found that the hydraulic oil would not form a stable emulsion and separated from the electrolyte, effectively removing it from the active area around the electrodes; as a result the hydraulic oil resisted oxidation. This limitation is not insurmountable and the results of these initial trials were considered to be sufficiently encouraging that further development is ongoing at AWE with the aim of developing a process that could be applied to radioactively-contaminated oils within a Controlled Area (Contamination).