In the frame of its radwaste disposal research programme, SCK•CEN started the construction of the HADES underground research facility in 1980. Including several extensions and a comprehensive experimental programme, it has provided a lot of experience on monitoring. Monitoring is performed for many reasons: construction follow-up, field characterisation, investigation of phenomena, and model validations — in which the underground lab offers the opportunity for upscaling conventional laboratory set-ups. Construction monitoring has allowed to develop and optimise the underground construction techniques in a previously poorly known environment, resulting in a well-mastered application of mechanised methods for gallery construction with minimal damage to the host formation. Access to this formation also allows its characterisation, both geotechnical, geological and geochemical, and the detailed investigation of phenomena such as fracturing and oxidation. Finally, instrumented set-ups allow to test various numerical models by comparing the observations with the predicted behaviour. The specific conditions of the underground laboratory put particular requirements to the sensors. These conditions include the long-term nature of many set-ups — typically several years to decades, the inaccessibility of many sensors after installation, high mechanical and water pressures, and corrosion. Combined with the fact that many sensors are custom made, obtaining and maintaining a fully functional instrumented set-up can be challenging. A lot of experience has therefore been gained which is very valuable when designing the monitoring of radwaste repositories — and it has allowed us to determine the critical success factors for monitoring. Engineers tend to look at this first from a technical viewpoint — and there are many technical aspects indeed that determine the reliability of monitoring. A first one is the combination of different observations (“redundancy”) which can be implemented by the use of several sensors, different sensor principles, different (coupled) parameters, and the combination of point measurements with geophysical techniques. Cabling is also a critical issue as it is often considered as the primary enemy of barrier integrity. Minimal cabling techniques, such as distributed fibre optic monitoring and wireless signal transmission, therefore get increasing attention. Also the interpretation of the monitoring data — in particular those that are perceived as “wrong” or “unexpected”, needs sufficient attention. The long-term experience has however also shown that the design of a monitoring programme must look beyond the technical part. In particular for long-term applications, issues such as data management and record keeping are vital to guarantee success in this.

The economic, environmental, social and societal constraints have pushed the nuclear industry to develop new dismantling and decommissioning (D&D) techniques in order to meet the current industrial requirements. Since 5 years, AREVA, in partnership with Air Liquide, has been working on a new cryogenic solution that can be used to achieve the main important D&D operations (i.e. cutting, surface decontamination and concrete scabbling). This solution, called NiThrow™, is based on the Nitrocision LLC concept and consists in spraying highly pressurized liquid nitrogen (3500 bar) at a very low temperature (−140°C) onto a surface. The main advantage in using liquid nitrogen instead of water is due to its rapid conversion into gaseous nitrogen that will avoid the generation of liquid effluent. The preliminary trials made in France at Saint Ouen l’Aumône and at the AREVA SICN facility by using NiThrow™ appeared to be very interesting and promising in terms of effectiveness, reduction of the dosimetry (i.e. ALARA principle) and adaptability to the different types of surfaces and materials to be treated. On an other hand, they also highlighted a couple of defects like a lack of reliability that is not compatible with a safety use in a nuclear environment. In 2009, with the aim of fixing all these issues, AREVA decided to start an extensive R&D program. This work has been essentially focused on the lancing tool, the insulation and the flexibility of the liquid nitrogen distribution network, but also on the design and the manufacturing of a waste collection unit and a dedicated carrier. Today, these R&D efforts have been totally completed. and resulted to the publication of 3 patents. Also, the collaboration, with Air Liquide and the American company Conco allows AREVA to propose to its customers a safe and reliable solution for their cleaning operations in France and all over the world.

Ongoing experimental work has been underway at selected nuclear sites in the Russian State Atomic Energy Corporation (ROSATOM) during the past two years to determine the effectiveness, reliability, application and acceptability of high technology polymers for liquid radioactive waste solidification. The long term project is funded by the U.S. Department of Energy’s Initiatives for Proliferation Prevention (IPP) program. IPP was established in 1994 as a non-proliferation program of DOE / National Nuclear Security Administration and receives its funding each year through Congressional appropriation. The objectives of IPP are: • To engage former Soviet nuclear weapons scientists, engineers and technicians, currently or formerly involved with weapons of mass destruction, in peaceful and sustainable commercial activities. • To identify non-military, commercial applications for former Soviet institute technologies through cooperative projects among former Soviet weapons scientists, U.S. national laboratories and U.S. industry. • To create new technology sources and to provide business opportunities for U.S. companies, while offering commercial opportunities and meaningful employment for former weapons scientists. Argonne National Laboratory provides management oversight for this project. More than 60 former weapons scientists are engaged in this project. With the project moving toward its conclusion in 2012, the emphasis is now on expanding the experimental work to include the sub-sites of Seversk (SCC), Zheleznogorsk (MCC) located in Siberia and Gatchyna (KRI) and applying the polymer technology to actual problematic waste streams as well as to evaluate the prospects for new applications, beyond their current use in the nuclear waste treatment field. Work to date includes over the solidification of over 80 waste streams for the purpose of evaluating all aspects of the polymer’s effectiveness with LLW and ILW complex waste. Waste stream compositions include oil, aqueous, acidic and basic solutions with heavy metals, oil sludge, spent extractants, decontamination solutions, salt sludge, TBP and other complex waste streams. Extensive irradiation evaluation (up to 270 million rad), stability and leach studies, evaporation and absorption capacity tests and gas generation experimentation on tri-butyl phosphate (TBP) waste have been examined. The extensive evaluation of the polymer technology by the lead group, V.G. Khlopin Radium Institute, has resulted in significant discussion about its possible use within the ROSATOM network. At present the focus of work is with its application to legacy LLW and ILW waste streams that exist in a variety of sectors that include power plants, research institutes, weapons sites, submarine decommissioning and many others. As is the case in most countries, new waste treatment technologies first must be verified by the waste generator, and secondly, approved for use by the government regulators responsible for final storage. The polymer technology is the first foreign sorbent product to enter Russia for radioactive waste treatment so it must receive ROSATOM certification by undergoing irradiation, fire / safety and health / safety testing. Experimental work to date has validated the effectiveness of the polymer technology and today the project team is evaluating criteria for final acceptance of the waste form by ROSATOM. The paper will illustrate results of the various experiments that include irradiation of actual solidified samples, gas generation of irradiated samples, chemical stability (cesium leach rate) and thermal stability, oil and aqueous waste stream solidification examples, and volume reduction test data that will determine cost benefits to the waste generator. Throughout the course of this work, it is apparent that the polymer technology is selective in nature; however, it can have broad applicability to problematic waste streams. One such application is the separation and selective recovery of trans-plutonium elements and rare earth elements from standard solutions. Another application is the use of polymers at sites where radioactive liquids are accidently emitted from operations, thus causing the risk of environmental contamination.

Material control and accountability (MC&A) operations that track and account for critical assets at nuclear facilities provide a key protection approach for defeating insider adversaries. MC&A activities, from monitoring to inventory measurements, provide critical information about target materials and define security elements that are useful against insider threats. However, these activities have been difficult to characterize in ways that are compatible with the path analysis methods that are used to systematically evaluate the effectiveness of a site’s protection system. The path analysis methodology focuses on a systematic, quantitative evaluation of the physical protection component of the system for potential external threats, and often calculates the probability that the physical protection system (PPS) is effective ( P E ) in defeating an adversary who uses that attack pathway. In previous work, Dawson and Hester observed that many MC&A activities can be considered a type of sensor system with alarm and assessment capabilities that provide reccurring opportunities for “detecting” the status of critical items. This work has extended that characterization of MC&A activities as probabilistic sensors that are interwoven within each protection layer of the PPS. In addition, MC&A activities have similar characteristics to operator tasks performed in a nuclear power plant (NPP) in that the reliability of these activities depends significantly on human performance. Many of the procedures involve human performance in checking for anomalous conditions. Further characterization of MC&A activities as operational procedures that check the status of critical assets provides a basis for applying human reliability analysis (HRA) models and methods to determine probabilities of detection for MC&A protection elements. This paper will discuss the application of HRA methods used in nuclear power plant probabilistic risk assessments to define detection probabilities and to formulate “timely detection” for MC&A operations. This work has enabled the development of an integrated path analysis methodology in which MC&A operations can be combined with traditional sensor data in the calculation of PPS effectiveness. Explicitly incorporating MC&A operations into the existing evaluation methodology provides the basis for an effectiveness measure for insider threats, and the resulting PE calculations will provide an integrated effectiveness measure that addresses both external and insider threats. The extended path analysis methodology is being further investigated as the basis for including the PPS and MC&A activities in an integrated safeguards and security system for advanced fuel cycle facilities.

In 1992, the U.S. Congress passed legislation to discontinue above- and below-ground testing of nuclear weapons. Because of this, the U.S. Department of Energy (DOE) must rely on laboratory experiments and computer-based calculations to verify the reliability of the nuclear stockpile. The Sandia National Laboratories/New Mexico (SNL/NM) Z machine was developed to support the science-based approach for mimicking nuclear explosions and stockpile stewardship. Plutonium (Pu) isotopes with greater than ninety-eight percent enrichment were used in the experiments. In May 2006, SNL/NM received authority that the Z Machine Isentropic Compression Experiments could commence. Los Alamos National Laboratory (LANL) provided the plutonium targets and loaded the target assemblies provided by SNL/NM. Three experiments were conducted from May through July 2006. The residues from each experiment, which weighed up to 913 pounds, were metallic and were packaged into a 55-gallon drum each. SNL/NM conducts the experiments and provides temporary storage for the drums until shipment to LANL for final waste certification for disposal at the Waste Isolation Pilot Plant (WIPP) in southeastern New Mexico. This paper presents a comprehensive approach for documenting generator knowledge for characterization of waste in cooperation with scientists at the two laboratories and addresses a variety of essential topics.

The long-term safety of a geological repository for high-level radioactive waste (HLW) in Japan is achieved using a multi-barrier system, with emphasis on an engineered barrier system (EBS) consisting of bentonite-based buffer material and a steel overpack containing vitrified wastes. Providing a wide variety of technical options for repository operation, which can allow EBS construction under various repository conditions, is valuable for the volunteering approach to siting a HLW repository in Japan. In order to confirm technological applicability of candidate techniques and to evaluate compatibility in an operation system, full-scale tests of various remote-handling and-emplacement techniques for the EBS were conducted. The test results are summarized in the technical menu which is hierarchical structured database. Since the long-term performance of EBS is significantly influenced by remote handling and emplacement techniques for EBS, consistency of the operation technology must be evaluated from the view point of system standing. In this program, evaluation methodology was studied and several indexes, representing requirements for the repository operation and management, were identified. Indices include current technical availability, long-term safety of the resultant EBS, operational safety, engineering reliability, retrievability, etc., with expert assessment complementing a more quantitative evaluation.

RWMC and JGC have been running an all-round R&D program for the period of 2000–2010 to develop the concept of Vertical Emplacement for disposal of vitrified waste. The conceptual design of its basic equipment was worked out in 2000, followed by forming the large-scale bentonite block in 2001–2004. Study has also been conducted on a mechanism to convey and position the large-scale block using a vacuum suction device. Subsequent to these developments, various technologies necessary for designing the Vertical Emplacement equipment have been reviewed, which would enhance engineering feasibility and reliability. Full-scale demonstration program under a joint research program with JAEA (Japan Atomic Energy Agency) started in 2008 with the twin objectives i) supporting of public relations and ii) technical verification. The large-scale bentonite block and part of the full-scale Vertical Emplacement equipment are now on view at the Full-scale demonstration facility in Horonobe, Hokkaido, Japan.

The paper describes the experience gained by the author in teaching decommissioning in the Highlands of Scotland. Initially when asked to teach the subject of decommissioning to students sitting for a BSc degree in “Electrical or Mechanical Engineering with Decommissioning Studies”, the author was taken aback, not having previously taught degree students and there was no precedent since there was no previous material or examples to build on. It was just as difficult for the students since whilst some had progressed from completing HND studies, the majority were employed at the Dounreay site and were mature students with families who were availing themselves of the opportunity for career advancement (CPD). Some of the students were from the UKAEA and its contractors whilst others were from Rolls-Royce working at Vulcan, the Royal Navy’s establishment for testing nuclear reactors for submarines. A number of the students had not been in a formal learning environment for many years. The College which had originally been funded by the UKAEA and the nuclear industry in the 1950’s was anxious to break into the new field of Decommissioning and were keen to promote these courses in order to support the work progressing on site. Many families in Thurso, and in Caithness, have a long tradition of working in the nuclear industry and it was thought at the time that expertise in nuclear decommissioning could be developed and indeed exported elsewhere. In addition the courses being promoted by the College would attract students from other parts so that a centre of excellence could be established. In parallel with formal teaching, online courses were also developed to extend the reach of the College. The material was developed as a mixture of power point presentations and formal notes and was obtained from existing literature, web searches and interactive discussions with people in the industry as well as case studies obtained from actual situations. Assignments were set and examination papers prepared which were validated by internal and external assessors. The first course was started in 2004 (believed to be unique at that time) and attracted eight students. Subsequent courses have been promoted as well as a BEng (Hons) course which also included a course on Safety and Reliability.

The Keeping the Nuclear Option Open (KNOO) research consortium is a four-year research council funded initiative addressing the challenges related to increasing the safety, reliability and sustainability of nuclear power in the UK. Through collaboration between key industrial and governmental stakeholders, and with international partners, KNOO was established to maintain and develop skills relevant to nuclear power generation. Funded by a research grant of £6.1M from the “Towards a Sustainable Energy Economy Programme” of the UK Research Councils, it represents the single largest university-based nuclear research programme in the UK for more than 30 years. The programme is led by Imperial College London, in collaboration with the universities of Manchester, Sheffield, Leeds, Bristol, Cardiff and the Open University. These universities are working with the UK nuclear industry, who contributed a further £0.4M in funding. The industry/government stakeholders include AWE, British Energy, the Department for Environment, Food and Rural Affairs, the Environment Agency, the Health and Safety Executive, Doosan Babcock, the Ministry of Defence, Nirex, AMEC NNC, Rolls-Royce PLC and the UK Atomic Energy Authority. Work Package 3 of this consortium, led by the University of Leeds, concerns “An Integrated Approach to Waste Immobilisation and Management”, and involves Imperial College London, and the Universities of Manchester and Sheffield. The aims of this work package are: to study the re-mobilisation, transport, solid-liquid separation and immobilisation of particulate wastes; to develop predictive models for particle behaviour based on atomic scale, thermodynamic and process scale simulations; to develop a fundamental understanding of selective adsorption of nuclides onto filter systems and their immobilisation; and to consider mechanisms of nuclide leaving and transport. The paper describes highlights from this work in the key areas of multi-scale modeling (using atomic scale, thermodynamic and process scale models), the engineering properties of waste (linking microscopic and macroscopic behaviour, and transport and rheology), and waste reactivity (considering waste hosts and wasteforms, generation IV wastes, and waste interactions).

Scaling factors are calculated based on a database from radiochemical analyses of representative waste samples. Several data points are needed to derive a reliable scaling factor. The more the number of data points, the better is the correlation, but more costly because of number of needed radiochemical analyses. Therefore, optimization of data points should be considered to minimize the cost without compromising reliability and prediction of the scaling factor. Scaling factors for Ni-63, Sr-90, and C-14 were calculated using Co-60 and Cs-137 as Key nuclides based on the published data in EPRI-4037. Correlation coefficient, percent error and relative standard deviation were plotted against the number of data points used in the estimation of scaling factor. The optimum number of data points was obtained to where there was no significant improvement in the statistical uncertainties by using additional samples. Even though non-segregated (all data points) has greater effect in calculating scaling factor because of its more data points, this study showed that even segregated data points can also give good DTM characterization.

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

The United States Department of Energy’s Carlsbad Field Office (CBFO) is responsible for the successful management of transuranic radioactive waste (TRUW) in the United States. TRUW is a long-lived radioactive waste/material. CBFO’s responsibilities includes the operation of the Waste Isolation Pilot Plant (WIPP), which is a deep geologic repository for the safe disposal of U.S. defense-related TRUW and is located 42 kilometers (km) east of Carlsbad, New Mexico. WIPP is the only deep-geological disposal site for long-lived radioactive waste that is operating in the world today. CBFO also manages the National Transuranic Waste Program (NTP), which oversees TRU waste management from generation to disposal. As of August 1, 2003, approximately 1890 shipments of waste have been safely transported to the WIPP, which has been operating since March 1999. Surface and subsurface facilities designed to facilitate the safe handling and disposal of TRU waste are located within the WIPP site. The underground waste disposal area is in a bedded salt formation at a depth of 650 meters (m). Approximately 176,000 m 3 of TRU waste containing up to 17 kilograms of plutonium will be emplaced in disposal rooms 4 m high, 10 m wide and 91 m long. Magnesium oxide (MgO) backfill will be emplaced with the waste to control the actinide solubility and mobility in the disposal areas. Properties of the repository horizon have been investigated in an underground test facility excavated north of the waste disposal area, and in which seals, rock mechanics, hydrology, and simulated waste emplacement tests were conducted. Thus, in some areas of broad international interest, the CBFO has developed a leading expertise through its 25-years WIPP repository and TRU waste characterization activities. The CBFO’s main programmatic responsibilities during the disposal phase are to operate a safe and efficient TRU waste repository at the WIPP, to operate an effective system for management of TRU waste from generation to disposal, and to comply with applicable laws, regulations, and permits. This responsibility requires maintenance and upgrades to the current technologies for TRU waste operations, monitoring, and transportation. This responsibility also requires the maintenance of scientific capabilities for evaluating the performance of the WIPP repository. Every 5 years, WIPP must be recertified for operations by the regulator, the EPA. Currently, the CBFO is preparing for the 2004 recertification. The CBFO/WIPP has been designated by the International Atomic Energy Agency (IAEA) as an International Center of Excellence. The IAEA is working with CBFO, other designated centers of excellence, and other member states in the IAEA to foster collaborative training activities and experiments in order to address major radioactive waste disposal issues. As the only operating deep radioactive waste repository in the world today, CBFO/WIPP is an important participant in this IAEA initiative. In addition to participating in relevant and beneficial experiments, the CBFO is providing the international community convenient access to information by sponsoring and hosting symposia and workshops on relevant topics and by participation in international waste management organizations and topical meetings. The CBFO has agreed to exchange scientific information with foreign radioactive waste management organizations. These activities result in the cost-effective acquisition of scientific information in support of increased WIPP facility operational and post-closure assurance and reliability. It also demonstrates the CBFO’s intent and resolve to honor international commitments and obligations.

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.

The UKAEA owns and operates a suite of cells designed to process and repackage Remote Handleable Intermediate Level Waste (RH-ILW) at the Harwell site in the UK. Cans of RH-ILW are retrieved and assayed using Non Destructive Assay (NDA) prior to repackaging in Nirex approved drums for interim above — ground storage, and eventual disposal in a national repository. The NDA system comprises a combined passive/active neutron interrogator (NI) and also a gamma spectrometer (GS). The results are used in conjunction with an expert system to aid the reconciliation between the declared / historical records and measured fissile mass and β/γ radionuclide inventories. The assay system was originally designed to accommodate the unique diverse range of waste streams, which are stored within B462. To this end, a key feature of the design was the inclusion of state of the art features such as matrix compensation, fissile material location correction, and self-shielding compensation for the active cycle of the neutron interrogator. The systems were originally designed and calibrated in the laboratory during the early 1990’s. Commissioning proceeded through staged acceptance tests and commissioning trials for the equipment. Active commissioning using selected real waste drums has recently been completed, and the UKAEA is now operating the plant under a testing phase of full trial operations. The UKAEA is rapidly gaining experience in both the operation of the equipment, and its performance for various waste streams. This has led to a number of recommendations for significant improvements with respect to both the mode of operation, and the physics algorithms. By modifying the operating procedures, it is possible to improve the level of management control over the interpretation of the assay results, so that a greater degree of confidence can be demonstrated in the conclusions drawn. The approach has been to streamline the use of all the available information (historical records, assay and expert system data), coupled with extensive operator training, in order to ensure that operations staff are in control of the decision–making process. Sentencing decisions are now made using a proceduralised, open framework / methodology using the experience gained from the commissioning trials to provide extensive operator training, thereby eliminating the risks associated with previous “black box” treatment of expert systems. Technical changes include, for example, a simplification of the treatment of self-shielding in lumps of fissile material, leading to more robust data interpretation, and substantial throughput improvements. In establishing the new procedures and system enhancements, the UKAEA is following the new national Good Practice Guide for NDA measurements. In this paper, we describe the approach being followed to identify and implement these changes, as well as describing some of the important changes in some detail. We also describe UKAEA’s plans to undertake a comprehensive system refurbishment, in order to modernise the equipment for increased throughout and minimal plant down–time, using supportable high reliability components. This will also maximise the operational lifetime of the equipment.

Uncertainties associated with spatially variable properties of geologic characteristics are especially important. However, these uncertainties cannot be specified in a deterministic manner because the process of formation or evolution of host rock is random and because the limited number of data is available for the statistic evaluation of this random process. We develop the hybrid model of fuzzy geostatistics for the appropriate treatment of uncertainties associated with spatially variable properties. For the application of the above method, we prepared the virtual field with heterogeneous hydraulic conductivity for the simulation of site characterization. In this simulation, we consider the situation that the limited number of boring data is available. The conventional method can determine only one variogram. In this case, uncertainties due to the ignorance are not appropriately handled and the degree of belief for the determined variogram is not specified. For the appropriate treatment of ignorance, we define the possible range of variogram in terms of fuzzy membership functions on the level of conservatism. A large number of heterogeneous hydraulic conductivity fields are realized by using corresponding variograms that sampled from the possible range of variogram. For each realized field, we compute the groundwater flow model and estimate the average flow velocity. Resulting many average flow velocities are processed by fuzzy rule to obtain the membership function of estimated average flow velocity. This can be used as the input data for the hybrid probabilistic and possibilistic safety assessment. Comparing with the previous work, the process of defining the membership function of average flow velocity becomes transparent and the reliability of safety assessment is enhanced.

As part of a development program for radioactive waste management, two test boreholes have been constructed at the MosNPO “Radon” Site to evaluate the conditions of LILW storage in such facilities, to observe the behaviour of engineered barriers as a function of time, the ability of the system to contain radionuclides and to test the reliability of a monitoring system. To obtain a licence for operation of such boreholes, MosNPO “Radon” specialists have prepared a package of documents, including a “Report on Safety Assessment of LILW Storage in Large Diameter Boreholes” [1]. On the basis of radiation safety standards and requirements being in force in Russian Federation, a range of possible accidents and emergencies during the operational period has been considered in this study. In this paper, the authors present results of a long term safety assessment of LILW disposal in the boreholes, performed within the framework of the above mentioned report.