bluuebunny/crossref_metadata_2025 huggingface dataset:
Information Overload: Processing Consultation Responses From Experts, Stakeholder Groups and Communities
Authors: Collier David , Orr Paula
Published: March 2009

Contemporary approaches to the design and implementation of consultation programmes have come a long way from the days of ‘decide-announce-defend’. However success in expanding involvement and combining different participation methods, especially within a complex multi-phase programme, still poses major challenges. This paper discusses one of them: processing the much increased volume and variety of contributions received and then combining them with other forms of input in a rigorous, balanced, auditable and transparent manner. It is no longer sufficient, if indeed it every really was, for someone simply to work their way through a stack of comments with a highlighter, altering a draft text where he or she judged that clarification was required. We illustrate the nature of the challenge and the response of the Environment Agency of England and Wales by considering some specific issues, including the following. • Applicability of systematic approaches to text and transcript analysis; • Use of qualitative data management systems; • Transparency vs confidentiality; • Inconsistencies between stakeholder/public concerns and regulatory boundaries; • Integration of different types of input; • Deciding what to take into account; judging validity and weight; • Implications for quality management, documentation and resources; • Approaches to programme evaluation.
Application of Optical and Imaging Techniques to Inspection of Off-Line Joule-Heated Melter at the West Valley Demonstration Project
Authors: Plodinec M. John , Jang Ping-Rey , Long Zhiling , Monts David L. , Okhuysen Walter P. , Philip Thomas , Su Yi
Published: March 2009

The West Valley melter has been taken out of service. Its design is the direct ancestor of the current melter design for the Hanford Waste Treatment Plant. Over its eight years of service, the West Valley melter has endured many of the same challenges that the Hanford melters will encounter with feeds that are similar to many of the Hanford double shell tank wastes. Thus, inspection of the West Valley melter prior to its disposal could provide valuable — even crucial — information to the designers of the melters to be used at the Hanford Site, particularly if quantitative information can be obtained. The objective of Mississippi State University’s Diagnostic Instrumentation and Analysis Laboratory’s (DIAL) efforts is to develop, fabricate, and deploy inspection tools for the West Valley melter that will (i.) be remotely operable in the West Valley process cell; (ii.) provide quantitative information on melter refractory wear and deposits on the refractory; and (iii.) indicate areas of heterogeneity of deposits, requiring more detailed characterization. A collaborative arrangement has been established with the West Valley Demonstration Project (WVDP) to inspect their melter.
Data Collection and Tracking of Radioactive Waste at the Los Alamos National Laboratory Plutonium Facility
Authors: Balkey J. J. , Dodge R. L. , Martinez B. T. , Wieneke R. E.
Published: March 2009

The Los Alamos National Laboratory (LANL) is one of two design laboratories in the United States Department of Energy’s weapons complex, with over 60 years of experience in handling radioactive materials, and, consequently, in managing radioactive waste. Actinide research and development is conducted at the Plutonium Facility, which has been in operation since 1978 and is the major source of radioactive waste at LANL. The Nuclear Materials Technology (NMT) Division is responsible for operating the Plutonium Facility and has a dedicated group of personnel who manage radioactive and hazardous waste, and address environmental regulations. The NMT Division also oversees analytical support operations in the Chemistry and Metallurgy Research Facility. Operations at these two nonreactor nuclear facilities generate a wide variety of waste that may be generally classified as sanitary, hazardous, radioactive, and mixed (both radioactive and hazardous). The expedient, cost-effective, and regulatory-compliant management of radioactive waste has been a challenging task, given the propensity for rapid change in the regulatory environment. One major asset is the availability of information on waste generation and characteristics in electronic form. To do so, the Waste Inventory Tracking system (WITS) was developed 6 years ago to collect and store this information. To record waste information in the field, technicians use handheld Palm Pilots®. These units are then docked with personal computers to transfer the data to WITS. The primary use of WITS is the automated generation of waste package data reports, which are used to demonstrate compliance with waste acceptance criteria and gain acceptance for waste disposal. The WITS data are also used to evaluate various aspects of waste generation and handling, and to track performance indicators. The WITS is a fundamental part of waste management in the NMT Division.
Integrated Quality Assurance Program for Transuranic Waste Management at the Los Alamos National Laboratory Plutonium Facility
Authors: Balkey J. J. , Sanchez L. J. , Wieneke R. E.
Published: March 2009

Los Alamos National Laboratory (LANL) is one of two design laboratories in the United States Department of Energy’s (DOE) weapons complex. It has over 60 years of experience in handling radioactive materials and, consequently, in radioactive waste management. The current focus at LANL for actinide research and development is the Plutonium Facility, which has been in operation since 1978 and is the major source of transuranic (TRU) waste. The Nuclear Materials Technology (NMT) Division is responsible for operating the Plutonium Facility. It has a dedicated group of personnel who manage radioactive and hazardous waste, and address environmental regulations. The TRU Waste Certification Program has prescriptive requirements that must be met for waste to be certified by the DOE’s Carlsbad Field Office, which provides oversight to the final waste repository, the Waste Isolation Pilot Plant (WIPP) in Carlsbad, New Mexico. Quality assurance expectations are also well defined, from top-tier documents such as 10 Code of Federal Regulations 830.120. Quality Assurance Requirements, which carry the force of law, through CAO-94-1012, Quality Assurance Program Document, from WIPP, to LANL internal working documents. Internal and external audits are conducted regularly to verify the adequacy of the program for meeting these requirements. To ensure compliance with quality requirements in waste operations, the NMT-7 Waste Management and Environmental Compliance Group has two full-time quality assurance (QA) specialists. They are responsible for maintaining the Quality Assurance Program (QAP). They work directly with waste management personnel, and they are frequently in the field, working beside waste management technicians. They are responsible for ensuring that applicable QAP elements are implemented as required, and that waste operations are effective. They review waste management program documents and waste operations for compliance with requirements, and they observe selected waste operations regularly to ensure that these operations are being conducted in accordance with established procedures. A yearly surveillance schedule is established to guide assessment activities, but it has the flexibility to allow the QA specialists to address any problem areas they may encounter. The QA specialists track performance indicators and evaluate them for systemic issues that may affect quality, including tracking program corrective actions to completion. Monthly reports on QA activities are submitted to group managers. The QA specialists are also asked, on occasion, to lead quality-related investigations and to work with operations personnel to propose solutions. As a direct result of their efforts, the waste management group won Pin˜on quality assurance recognition from Quality New Mexico for the group’s commitment to quality.
Characterisation of Geopolymers for the Immobilisation of Intermediate Level Waste
Authors: Perera D. S. , Vance E. R. , Aly Z. , Finnie K. S. , Hanna J. V. , Nicholson C. L. , Trautman R. L. , Stewart M. W. A.
Published: March 2009

Geopolymers are made by adding aluminosilicates to concentrated alkali solutions for dissolution and subsequent polymerisation to take place. Their physical behaviour is similar to that of Portland cement and they have been considered as a possible improvement on cement for several applications including as media for the encapsulation of hazardous or low/intermediate level radioactive waste. We studied in detail a commercial geopolymer to try to get a better understanding of geopolymers in order to enhance its leach resistance for immobilisation of intermediate level radioactive waste. We also briefly investigated two types of experimental geopolymers, one made with a metakaolinite and another from fly ash as the aluminosilicate source. The commercial geopolymer paste had an apparent porosity of 26% and it was possible to reduce it to 17% by adding ∼ 30 weight % foundry sand. The apparent porosities of the geopolymer made from metakaolinite and from fly ash were 13% and 26% respectively. X-ray powder diffraction showed in the three geopolymers, an amorphous phase (deduced by the presence of a very broad diffuse peak centred at a d-spacing of ∼ 0.32 nm), quartz and other minor phases. The energy dispersive spectroscopic analysis under the scanning electron microscope confirmed these. Magic angle spinning nuclear magnetic resonance data from the samples showed Al to be mainly in 4-fold coordination and Si sites varying from Q0 to Q4 coordination as also found by other researchers. 23Na spectra indicated that the Na was mainly in the pore water. The 133 Cs spectra showed a strong possibility of Cs being mainly bound in the structure while a small amount could still be in the pore water. The initial leach tests showed alkalis were leached out at rates of several orders of magnitude more than the Al and Si network ions. The most likely reason is that a significant alkali inventory is in the pore water. To remove pore water and incorporate simulated radionuclides such as Cs in the network the commercial geopolymer was heated up to 1200°C. Differential thermal/thermogravimetric analysis showed the loss of water occurs in three stages and most had been lost by 700°C. These results are in broad agreement with the Infra red spectra obtained for samples heated over the temperature range 30–900°C. The broad water band intensity in the range 2600-30-900°C. The broad water band intensity in the range 2600–3800 cm −1 decreased steadily with temperature although a small fraction remained even after heating to 500°C. The silanol band had disappeared at 800°C, and the 3619 cm−1 band (due to OH) virtually disappeared by 900°C.
Successful Implementation of a Behavioral Safety Program at the Los Alamos National Laboratory Plutonium Facility
Authors: Balkey J. J. , Kleinsteuber J. F. , Wieneke R. E.
Published: March 2009

Los Alamos National Laboratory (LANL) is one of two design laboratories in the United States Department of Energy (DOE) weapons complex, with over 60 years of experience in handling hazardous and radioactive materials. Actinide research and development are performed in two nuclear facilities. The Plutonium Facility has been in operation since 1978, and the Chemistry and Metallurgy Research (CMR) Facility was built in 1952. The Nuclear Materials Technology (NMT) Division is responsible for operating both facilities in a safe and environmentally sound manner. At these NMT Division nuclear facilities, the primary hazards that are associated with chemicals and radioactive materials are well controlled, with minimal risk to the workforce and the public. Most workforce injuries are physical or ergonomic in nature. In an effort to increase safety awareness and to decrease accidents and incidents, a program focusing on identifying and eliminating unsafe behaviors was initiated. This process is named ATOMICS (for Allowing Timely Observations Measures Increased Commitment to Safety). Workers are trained on how to conduct safety observations of one another and given guidance on specific behaviors to note. Observations are structured to have minimal impact on the workload and are shared by the entire workforce. This program has effectively decreased an already low accident rate and will make long-term sustainability possible. The behavior-based safety process engages the workforce in the implementation and utilization of their own safety initative. The process is based on the simple act of having workers observe other workers and provide feedback on safe and at-risk behaviors. Observations typically take 10 to 15 minutes. Observations are strictly conducted under the conditions that no names are used and no blame is placed. A prospective observer takes 2 days of training on how to perform observations. To date, this behavioral safety program has been effective in further reducing the low accident/injury rate for the division. Most of the division’s workforce has taken observer training, and about half of those trained perform approximately 2500 observations per year on work conducted in NMT Division nuclear facilities.
Testing Safety and Performance Indicators in an Assessment of the Long-Term Performance of the Geological Disposal of Spent Fuel in Boom Clay
Authors: Sillen Xavier , Marivoet Jan , Cool Wim , de Preter Peter
Published: March 2009

The classical numerical output, or indicator, from assessments of the long-term safety of geological disposal systems for high-level radioactive waste is the individual effective dose rate. This indicator is an estimate of the possible individual health detriment and it is commonly compared to regulatory limits for assessing the safety of other nuclear activities as well, such as medical and industrial activities. As a safety indicator, the individual dose rate provides an estimate of the overall safety of the disposal system. However, because of the time frames involved in safety assessments of geological disposal systems, the need arises of complementary safety indicators that could be less affected by uncertainties like those associated with future human behaviour or the effects of climate change on the biosphere and the aquifers. Such alternative safety indicators can be, for example, radionuclide concentrations in the groundwater or fluxes to the biosphere due to a repository. Safety indicators only tell how globally safe a disposal system is. For confidence building, performance indicators can be used in addition to tell how the system works. In particular, performance indicators such as fluxes, activities or activity concentrations of selected radionuclides can show how the different components of the system fulfil their safety functions and contribute to the overall safety. The SPIN project of the European Commission assessed the usefulness of seven safety indicators and fourteen performance indicators by testing them in four actual assessments of disposal systems in granite formations. In this paper, indicators calculated from an assessment of the disposal of spent fuel in the poorly indurated Boom Clay formation are presented. Conclusions from the SPIN project that hold for repositories in clays are highlighted, as well as results that illustrate differences between the granite and clay disposal options. Finally, various performance and safety indicators are combined into a logical sequence to comprehensively present, and explain, the results of a safety assessment.
Modelling 226Ra, 222Rn, and 210Pb Migration in a Proposed Surface Repository of Very Low-Level Long-Lived Radioactive Waste
Authors: Mallants Dirk , Jacques Diederik , Zeevaert Theo
Published: March 2009

A generic concept for disposal of very low-level long-lived radioactive waste is currently being evaluated for radium bearing wastes that originated from a historical radium extraction plant at Olen, Belgium. A total volume of approximately 217 000 m3 of waste with an average radium content of 7 Bq/g has to be disposed of. Upon request by the Belgian National Agency for Management of Radioactive Waste and Nuclear Fuels, NIRAS/ONDRAF, a generic disposal concept was evaluated for the purpose of identifying the minimum disposal concept which guarantees long term safety. Such an analysis would provide useful input to the final design of the disposal concept, as the contribution of the different engineered barriers to the overall safety will have been assessed. The analysis focussed on the migration of 226Ra, 222Rn, and 210Pb to groundwater owing to infiltration of rainwater and the diffusive radon transport via the gas phase to the atmosphere. The generic design considered a waste dump with the contaminated material completely enclosed by a clay barrier. To protect the low-permeability clay from degradation by water and wind erosion, frost and desiccation, burrowing animals, and plant roots, among others, a multi-layer cap was designed to meet these requirements. In the performance assessment calculations, various cases were considered in which one component of the disposal concept was changed at the time. Cases considered included absence of clay layers, effects of hot spots, lower adsorption capacity of various materials, and the effect of separating the radium contaminated material from nearly uncontaminated material. Unsaturated flow calculations were done first to estimate the steady-state water content profile. Knowledge on the degree of water saturation is of paramount importance for radon transport through the gas phase. Based on the steady-state water content profile, advective dispersive transport calculations were done considering the decay chain reaction of 226Ra. In addition to the radon gas flux to the atmosphere, fluxes of 226Ra, 222Rn, and 210Pb to groundwater were also produced. Groundwater flow and transport calculations yielded radionuclide concentrations in a hypothetical well nearby the planned disposal site, whereas biosphere modelling provided the annual doses to the public considering the groundwater pathway and direct inhalation of radon in case of the atmospheric pathway. On the basis of the calculated radionuclide fluxes and doses the importance of the various model parameters and concept components will be evaluated and discussed.
Achievements and Current Issues in Decommissioning of Research Reactors
Authors: Laraia Michele
Published: March 2009

Although the state-of-the-art technology for decommissioning nuclear research reactors is probably adequate to cope with most difficulties associated with the dismantling of such facilities, it is generally imperative to improve, adapt or optimise technologies for the specific needs of the reactor to be dismantled. Learning from others rather than re-inventing the wheel makes sense in today’s globalisation context. This approach is expected to match the needs of many developing countries. The objective of the IAEA’s Co-ordinated Research Project (CRP) on Decommissioning Techniques for Research Reactors was to promote the exchange of information on the practical experience gained by Member States in decommissioning or operation, maintenance, and refurbishment activities which would be eventually related to the decommissioning of research reactors. Special emphasis was given to the development/adaptation of methods and approaches for optimisation of the decommissioning process. Fourteen institutions from thirteen Member States of the IAEA took part in this CRP that lasted from 1997 to 2002. It is felt that the IAEA project succeeded in transferring information and know-how from active decommissioning projects to those planning for decommissioning. It is also expected that this project will draw Member States’ attention to the need for timely planning for and implementation of decommissioning. In some Member States there are research reactors that are kept in an extended state of shutdown, pending decisions on continued operation, extensive refurbishment or decommissioning. This situation — which frequently lasts for many years — weighs heavily on staff morale and motivation, state resources, entails deterioration of structures and components, and may in the longer term have very serious safety implications. The IAEA project is expected to offer the Member States the opportunity of considering financial and other impacts of decommissioning research reactors, so that decommissioning actions can be initiated without undue delay. Aspects such as fuel and waste management and provisions for other technical, administrative and financial resources require timely preparation.
Web-Based Data Management System for Long-Term Performance Monitoring and Stewardship of a Low-Level Radioactive Waste Disposal Facility
Authors: Benson Craig H. , Tipton Robert J. , Kumthekar Uday , Chiou J. D.
Published: March 2009

A web-based data management system is described that is being used for long-term performance monitoring and stewardship of a low-level radioactive waste disposal facility (LLRWDF) at the Fernald Closure Project (FCP) in Fernald, Ohio, USA. The system provides integrated data collection and management for monitoring performance of the liner and cover systems in the LLRWDF. Integrated performance monitoring data (meteorological, hydrological, geophysical, ecological, leachate flow, and leachate quality data) are presented and compared with performance criteria and action levels in near real-time. A status summary is also included for the site manager, regulators, and other stakeholders to provide an early warning regarding the need for corrective action.
Automated Analysis of Trichloroethene and Chloroform
Authors: Burge Scott R.
Published: March 2009

Chloroform and trichloroethene (TCE) are two organic contaminants commonly encountered in ground water. TCE, formerly a common cleaning solvent, is usually associated with contaminated aquifers. Chloroform is usually associated with the chlorination of municipal water. The remediation level for TCE in aquifers is typically 5 ppb, therefore, the analytical method employed for monitoring these analytes must be capable of detecting and quantifying the analytes in the low ppb concentration range. The most common analytical methodology for the determination of TCE or chloroform in water is a purge and trap technique for sample introduction into a gas chromatographic system equipped with electroconductivity or mass spectroscopy detector. The instrumentation has a method limit of detection (LOD) of less than 0.5 ppb for TCE and chloroform, however, the expense, size and complexity of the gas chromatographic techniques limit its use outside the laboratory environment. An alternative to the gas chromatographic method for the analysis of select volatile chlorinated compounds in the low concentration range is an analytical instrument based on a halocarbon-specific optrode. The principle of detection is a quantitative, irreversible chemical reaction (modified Fujiwara reaction) that forms visible light-absorbing products. The operational basis of the optrode is the measure of the time history of the development of the colored (red) product formed by the reaction of the target analytes. The optrode has the selectivity and sensitivity for monitoring TCE and chloroform at the low ppb concentration range in the presence of other volatile chlorinated contaminants. The low-power requirements and simplicity of design make it a good choice for remote operations. This paper presents the analytical results (January 2002, to December 2002) of a panel-mounted instrument used to monitor the influent and effluent water of a TCE treatment facility located in Scottsdale, Arizona, and the analytical results of a well-mounted instrument used to monitor ground water (May 2002, to August 2002) at Edwards Air Force Base, California.
Experience From Safety Assessment of the Existing Storage Facility for Solid Radwaste at Ignalina NPP
Authors: Poskas P. , Ragaisis V. , Smaizys A.
Published: March 2009

There is only one nuclear power plant in Lithuania — Ignalina NPP (INPP). The INPP operates two similar units with present power level of about 1250 MW(e) each. They were commissioned (first grid connection) in 1983 and 1987 respectively and provide approximately 70–80 per cent of the electricity produced in Lithuania. From the beginning of INPP operation all power plant generated solid radwaste is collected and stored in the solid radwaste facility that is located in INPP sanitary protected area. The storage facility for solid radwaste at INPP consists of four above ground buildings, namely buildings No. 155, No. 155/1, No. 157 and No. 157/1. This facility is Soviet type facility designed for the interim storage of low and intermediate level radwaste arising as a consequence of operation of NPP. The institutional waste from industry, hospitals etc., including sealed sources as well, has been stored in this facility from 1990. At present this radioactive waste storage facility is still not licensed but SAR is prepared and licensing process is going on. The paper provides information on INPP solid radwaste storage facility and discusses problems met during the safety assessment of this facility: lack of the information on the inventory data, lack of design data for the buildings etc. Modeling results of the structural analysis, radiological data, shielding analysis and proposals to improve the safety of the facility also are discussed.
RTS - 1 - Galilei Decommissioning Project
Authors: Boschi A. , Cimini E. , Pagni F. , Parracone L. , Pocai M. , Russo M. , Spano M.
Published: March 2009

The RTS-1 “Galileo Galilei” is an open pool research reactor light water moderated and cooled. It had a maximum thermal output of 5 MWth and an average thermal flux of 5 E+13 n/cm2sec. It became critical for the first time on April 1963 and it was definitely shutdown in March 1980. The reactor is situated at CISAM (Joint Centre of Studies and Military Application - Italian Ministry of Defence), S. Piero a Grado, Pisa, Italy, and its decommissioning is in progress. In this paper the strategy adopted to achieve the green status of the reactor site is discussed, with particular attention on the different steps to be done according to the national laws. Emphasis is placed on the characteristics of two different conditions required, namely Passive Protective Custody, which is a step necessary to allow the decay of the radioactive materials present into the plant to decrease the radiological risk to operate safely, and Unconditioned Release, in which all the materials can be released without radiological restrictions. Another aspect discussed in this paper is the effort spent on the determination of the radioisotopic abundance of the reactor components, the personal dose evaluation due to the necessary activities to achieve two different status of “Passive Protective Custody” and “Unconditioned Release” and the waste characterisation. The necessary authorisations to start decommissioning has been obtained as far as concern the removal of spent fuel and the dismantling of some experimental equipments.
Real-Time Support for Precision Excavation of Radionuclide Contaminated Soils
Authors: Durham L. A. , Johnson R. L. , Miller D. S.
Published: March 2009

Traditional approaches to excavation design for sites with radionuclide-contaminated soils generally involve work plans with fixed excavation footprints and post-excavation verification sampling. Recent advancements in real-time data collection systems, combined with data loggers, location control systems, and secure project-support Web sites, allow for an alternative, more precise approach to excavation design and implementation. In this mode, traditional work plans are replaced with dynamic work plans. Excavation efforts are organized by lift, with real-time dig face screening performed and the data for each lift analyzed, before excavation continues. Rather than specifying excavation footprints, dynamic work plans identify the decision logic that will be used to determine footprints based on real-time data collection. The level of investment in excavation support data collection can be balanced against potential cost savings realized through waste stream minimization. Secure project support Web sites ensure that data sets generated during the remediation process are readily accessible to all who need to see them, whether they are on site or not. These include regulators, program managers, and technical support staff. This type of Web site allows for quick problem resolution, increased transparency in field decision making, and more efficient allocation of expensive technical staff time.
Development of Actinide-Containing Waste Immobilization Process
Authors: Stefanovsky S. V. , Ptashkin A. G. , Kuliako Y. M. , Perevalov S. A. , Yudintsev S. V. , Chekmarev A. M. , Ochkin A. V. , Chemarev A. M.
Published: March 2009

Actinide wastes involve actinide or rare earth–actinide fractions of high level waste (HLW), Pu-contaminated materials, including incinerator ashes, excess weapons plutonium, and some wastes formed during plutonium conversion in MOX fuel and nuclear accidents. SIA Radon in cooperation with Vernadsky Institute of Geochemistry, Institute of Geology of Ore Deposits, and D. Mendeleev University of Chemical Technology deals with development and testing of actinide waste forms and preparation methods. Zirconolite, pyrochlore, and murataite are considered as host phases for plutonium and other actinides. Two-phase ceramics based on zirconolite-perovskite, pyrochlore-perovskite, perovskite–cubic zirconia-based solid solution, murataite-perovskite, and zirconolite-murataite assemblages were designed for incorporation of actinide and rare earth–actinide fractions of HLW. Glass-ceramics containing apatite-britholite phases have been proposed for incinerator ash fixation. All these matrices have high chemical durability and radiation stability. The most promising method for production of these waste forms is an inductive melting in a cold crucible. Cold pressing and sintering technology is considered as alternative route. Mechanical activation intensifies ceramization process and reduces sintering temperature. Some new methods such as selfsustaining synthesis and plasma melting are being also examined.
Canada’s New Nuclear Waste Act
Authors: Brown Peter A.
Published: March 2009

The Nuclear Fuel Waste (NEW) Act is a stand-alone piece of legislation with some 30 articles and without regulations. The Act deals essentially with social, financial and socioeconomic considerations of the long-term management of nuclear fuel waste. It complements the health, environment, safety and security requirements under the Nuclear Safety and Control, Act. The NFW Act provides for 1) the nuclear industry to set up a waste management organization to manage the full long-term waste management activities and to establish trust funds to finance long-term waste management responsibilities; and 2) the waste management organization to submit, for government decision, long-term waste management options within three years of coming into force.
Progress With Multinational Storage and Disposal Concepts
Authors: McCombie Charles , Chapman Neil
Published: March 2009

The growing importance of finding shared solutions for countries with small arisings of long-lived wastes from nuclear power generation has been increasingly recognised over the last few years. In this paper, we take a systematic look at recent developments focusing in turn on: • international initiatives (IAEA); • regional initiatives the European Union (EU); • national positions for in 3 categories of countries: • those participating in the Arius association that was founded explicitly to promote the multinational concept; • those with specific interests in shared solutions; • those that might consider hosting a repository; • those with laws or policies requiring national disposal. Multinational concepts continue to face challenges that are more difficult than for national programmes. Nevertheless, progress to date indicates that the nuclear communities of the world may well be on the way to optimising radioactive waste management on a global scale.
Saxton Nuclear Experimental Corporation, Containment Vessel (CV) Concrete Removal: Decommissioning in a Flood Plan
Authors: Endsley Stephen L.
Published: March 2009

The Saxton Nuclear Experimental Corporation (SNEC) constructed and operated a nuclear reactor in rural Pennsylvania as an early demonstration project. The 23.5 MWth pressurized water reactor (PWR) operated from 1962 through 1972, and completed operations with failed fuel experiments. The facility was placed into SAFSTOR and decommissioning activities began in the mid 1990’s. The reactor facility was sited on the banks of a Pennsylvania river within the immediate one hundred year floodplain. Because of the elevated groundwater levels at the site, completion of the decommissioning process would prove to be challenging. The containment was constructed as a vertical steel cylinder of 11/16” (1.75 cm) thick carbon steel, one hundred and nine feet (33 meters) tall with a diameter of fifty feet (15.25 meters). The bottom of the containment vessel was torispherical steel located approximately fifty feet below grade in the flood plain of Central Pennsylvania’s Juniata River. Construction of the vessel included an internal concrete structure that was designed as the ballast to prevent flotation of the vessel, and provided mechanical structure for the reactor and equipment installation. The steel cylinder provided the forms for the internal concrete installation. The initial phase of decommissioning for the reactor containment consisted of the complete removal of the reactor and associated components, and was completed in the late 1990’s without regard to stabilization requirements required to maintain the stability of containment. Due to concerns of possible radioactive material between the concrete structure and the steel shell, it was decided that the internal concrete would be removed. Removal of the concrete constituted removal of the ballast material and possibly degraded the structural integrity of the CV. Therefore, stabilization of the steel cylinder was required to prevent uplift (flotation) and/or deformation of the steel shell. This stabilization required the installation of a bedrock anchoring system, internal and external anti-buckling steel beam stiffener rings, and a complete site dewatering system. TLG Services completed concrete removal in October of 2002, and this paper will provide information and details of the complex concrete removal project, project challenges, and lessons learned.
Gamma Spectrometry Systems for the Assay of Uranium Residues and Potentially Contaminated Low Level Wastes at the Capenhurst Site
Authors: Clark P. A. , Parvin D. F. , Powrie C. Y. , Orr C. H. , Mottershead G. , Forbes A.
Published: March 2009

BNFL has produced and operates a wide range of DrumScan® gamma measurement systems for monitoring packages, drums and boxed wastes arising from nuclear power plant reprocessing, fuel fabrication and decommissioning operations. The challenges associated with decommissioning operations are met by employing a range of technologies predominantly High Resolution and Low Resolution spectrometry (HRGS & LRGS). This paper describes how BNFL Instruments’ LRGS and HRGS DrumScan® gamma measurement systems have been used for the assay of uranium resides and potentially contaminated low level wastes by Capenhurst Integrated Decommissioning Project (IDP) in the UK. A description of the two Capenhurst segmented HRGS systems is included. Whilst Segmented Gamma Scanning is a well established technique for the non-destructive assay of gamma emitting radioisotopes in drummed waste, these systems utlise unique features to address the specific measurement requirements. The first system is configured for the accurate measurement of both small sized containers of uranium residues arising from recovery operations and low level wastes potentially contaminated with uranium contained in 200 litre drums. To achieve a high level of accuracy, this system uses a novel mechanical arrangement to overcome the wide variety of container sizes, and the unique “TransWeight” and “Transmission” matrix correction techniques which provide significant improvements over conventional Segmented Gamma Scanner matrix correction techniques. The second system is configured for Nuclear Safety purposes to provide an upper limit of the 235U present in 200 litre drums of potentially contaminated waste prior to the opening of the drums for sorting and uranium recovery operations. This system is configured to report an appropriately pessimistic upper estimate of the 235U present. A brief description of the LRGS systems used by Capenhurst is also provided. These systems have served to quantify the 235U content within a variety of potentially contaminated waste items ranging from 200 litre drums to 1m3 boxed waste.
Thermochemical Conversion of Asbestos Contaminated With Radionuclides and/or Other Hazardous Materials
Authors: Timmons Dale M. , Cahill James H.
Published: March 2009

Waste asbestos from abatement activities at Department of Energy (DOE) facilities is typically (as is most asbestos waste in the United States) disposed of in landfills. However, some of the asbestos from DOE facilities is contaminated with radionuclides, PCBs, metals regulated under the Resource Conservation Recovery Act (RCRA) and perhaps other regulated components that may require treatment instead of landfill disposal. Land disposal of waste is becoming less desirable to the public and does nothing to reduce the toxicity or the continued liability associated with these wastes. Methods for permanent destruction of these wastes are becoming more attractive as a final solution. One of the methods available for the destruction of asbestos-containing wastes is thermochemical conversion technology. ARI Technologies, Inc. was contracted by the National Energy Technology Laboratory (NETL) to conduct a technology deployment of its thermochemical conversion process. The purpose of the project was to: 1. “Destroy 10,000 lb. of asbestos-containing material (ACM), defined as asbestos fibers and binder by feeding it through an EPA-permitted asbestos destruction technology, such that the resultant materials are no longer considered to be asbestos in accordance with 40 CFR 61.155, Standard for Operation that Convert Asbestos-Containing Waste Materials Into Non-asbestos, and 2. Collect and analyse performance data for the deployed asbestos destruction technology.” In addition to the mandatory objectives, ARI conducted tests on the asbestos that were designed to evaluate the effectiveness of the technology for immobilization of toxic metals and surrogate radionuclides that are known to be present in DOE asbestos waste. This full-scale technology deployment demonstrated economical asbestos destruction and effective immobilization of lead, cadmium, barium and arsenic. Cerium oxide and non-radioactive cesium were also immobilized. Leach testing using EPA and DOE methods showed that leach performance surpassed regulatory criteria by at least one order of magnitude.
Conditioning of Radioactive Waste Resulted From the Treatment of Liquid Waste From the Romanian Nuclear Power Plant
Authors: Arsene Carmen , Negoiu Dumitru
Published: March 2009

For the conditioning of spent resins contaminated with radionuclides, such as: 137Cs, 134Cs, 60Co, 58Co, 57Co, 54Mn, etc., techniques of direct immobilization in cement, bitumen and organic polymers have been tested. The selected process was the bituminization using industrial bitumen, I 60–70, made in Romania, which had very good immobilizing characteristics. The paper presents stages of the research project, technical conditions for the process and advantages of the bituminization of spent resins.
German “Decommissioning Guideline”
Authors: Rehs Bernd , Warnecke Ernst
Published: March 2009

In Germany the legal and regulatory framework is designed to be applicable to all types of nuclear facilities / activities including the decommissioning of nuclear facilities. This generic approach has been developed for regulating nuclear activities over the many years of nuclear experience. Provisions applicable to an individual subject area, such as decommissioning of nuclear facilities, are “scattered” in the various documents of the national legal and regulatory framework. In order to provide assistance to those who are dealing with decommissioning of nuclear facilities, a “Decommissioning Guideline” was published in 1996 [1], which compiled all regulations relevant for licensing and supervising the decommissioning of nuclear facilities and also helps to harmonize the overall licensing process. A great number of amendments were made to the legal and regulatory provisions in recent years, in particular on the phase out of nuclear energy generation; on the environmental impact assessment; and on the release of buildings, materials and sites from nuclear regulatory control. As a consequence, the “Decommissioning Guideline” needed an update which is scheduled to be finished by end 2003. Information on the overall legal and regulatory situation, the revision of the “Decommissioning Guideline” and, in particular, on new legal and regulatory requirements is being provided.
Industrial Complex for Solid Radwaste Management at Chernobyl Nuclear Power Plant
Authors: Rausch Johannes , Ahner Stefan
Published: March 2009

In the framework of the preparation for the decommissioning of the Chernobyl Nuclear Power Plant (ChNPP) an Industrial Complex for Solid Radwaste Management (ICSRM) is going to be built under the EC TACIS Program in the vicinity of ChNPP, comprising: • a solid waste retrieval facility (LOT 1), • a solid waste processing plant (LOT 2), and • a repository for the disposal of short-lived waste (LOT 3).
Development and Implementation of Technology for the Treatment and Encapsulation of Operational Intermediate Level Wastes
Authors: Butcher Ed , Caldwell Rob , Godfrey Hugh , Hayes Martin , Miller Eric
Published: March 2009

Since 1982, BNFL Research and Technology has run an extensive research, development and implementation programme to define suitable materials and processes for the long-term containment of a wide variety of Intermediate Level Waste (ILW) streams arising from nuclear operations. This has provided both the empirical and underpinning data required to allow regulatory approval for the construction, commissioning and operation of four waste stream specific, cement-based encapsulation plants at the Sellafield site. The research and development programme has also succeeding in producing a team of highly skilled and experienced specialist researchers and technical staff equipped with a detailed database of knowledge concerning encapsulation methods applicable for a wide variety of ILW streams of concern within the UK and abroad. The objectives of this paper are to: • review the systematic, phased approach taken to assess potential encapsulants and develop suitable wasteforms and processes for ILW treatment that lead to the selection of a cement-based matrix as the preferred option for current ILW arisings, • discuss the key issues associated with the development of successful waste treatment systems and wasteforms, which have been encountered and addressed, • introduce ongoing work to confirm the longer term properties of archived full scale cemented wasteforms and provide additional information on the fundamental mechanisms which dictate waste-encapsulant interactions and wasteform behaviour.
Underground Research Laboratories in Japan: What Are the Important Factors for Facilities Design?
Authors: Sato T. , Mikake S. , Sakamaki M. , Aoki K. , Yamasaki S. , Shigeta N. , Takeda S.
Published: March 2009

This paper describes the current status of two Japanese off-site Underground Research Laboratories (URLs) Projects, one for crystalline rock and the other for sedimentary rock. This paper is focused on mechanical stability and ventilation, important factors relevant to the design and construction of deep underground facilities. High-pressure inflow, another important factor, will be included in the URL project for crystalline rock. The site of the URL project for crystalline rock is located in Mizunami, Gifu, in the central part of the main island of Japan. The regional geology consists of the Tertiary and Quaternary sedimentary rocks overlying Cretaceous granitic basement. Surface-based investigations, including geological mapping, a seismic refraction survey and shallow borehole investigations, and site preparation at the MIU (Mizunami Underground Research Laboratory) Project site have started in 2002. Numerical analysis is carried out to understand mechanical stability around the openings. The ventilation system design is based on numerical analysis using a ventilation network model. Grouting against the high-pressure inflow is planned. Conceptual design for the MIU at present is as follows: • Two 1,000 m shafts, a Main Shaft (6.5m φ) and a Ventilation shaft (4.5m φ); • Two experimental levels, the Main Stage at 1,000 m and the Middle Stage, at 500 m depths. The site of the URL project for sedimentary rock is located in Horonobe, Hokkaido, north of the main island of Japan. The geology consists of Tertiary sedimentary rocks. Surface-based investigation phase started in 2001. Numerical analysis is carried out to understand mechanical stability of the openings, and to design support. The numerical analysis using ventilation network model is carried out to design the ventilation system and disaster prevention method. Conceptual design for the Hnb-URL at present is as follows: • Two 500 m shafts and a Ventilation shaft; • Two experimental levels, the Main Stage at 500 m and the Middle Stage at 250 m depths.
Ion Irradiation Damage in Zirconate and Titanate Ceramics for Pu Disposition
Authors: Stewart M. W. A. , Begg D. , Finnie K. , Colella M. , Li H. , McLeod T. I. , Smith K. L. , Zhang Z. , Weber W. J. , Thevuthasan S.
Published: March 2009

A wasteform must meet several key requirements: (i) aqueous durability as related to a repository environment, the measurement of which must also allow for long-term processes such as radiation damage and annealing; (ii) the compositional flexibility to cope with real waste streams, which are often variable and not well characterised; (iii) the wasteform must be easy to process and the process parameters must be large enough to cope with changes; and, (iv) the waste loading must be high enough to make the wasteform economically viable. In this paper, we discuss the effect of ion irradiation on pyrochlore-rich titanate and defect-fluorite zirconate ceramics designed for plutonium immobilisation. Samples, with Ce as an analogue for Pu, were made via oxide routes and consolidated by cold-pressing and sintering. Ion irradiation damage was carried out with 2 MeV Au2+ ions to a fluence of 5 ions nm−2 in the accelerator facilities within the Environmental Molecular Sciences Laboratory at Pacific Northwest National Laboratory. Irradiated and non-irradiated samples were examined by x-ray diffraction, scanning and transmission electron microscopy, x-ray photoelectron and infra-red spectroscopy, and spectroscopic ellipsometry. Samples underwent accelerated leach testing at pH 1.75 (nitric acid) at 90°C for 28 days. The zirconate samples were more ion-irradiation damage resistant than the titanate samples, showing little change after ion-irradiation whereas the titanate samples formed an amorphous surface layer ∼ 500 nm thick. While all samples had high aqueous durability, the titanate leach rate was ∼ 5 times that of the zirconate. The ion-irradiation increased the leach rate of the titanate without impurities by ∼ 5 times. The difference in the leach rates between irradiated and unirradiated zirconate samples is small. However, the zirconates were less able to incorporate impurities than the titanate ceramics and required higher sintering temperatures, ∼ 1500°C compared to 1350°C for the titanates.