By now, in Russia more than 70 000 fuel assemblies (FA) have operated their service life at NPPs with RBMK–1000 (11 units); the discharged FAs have been stored in water of cooling ponds and in spent nuclear fuels (SNF) storage facilities for 25 years. During their operation, more than 3 000 spent fuel assemblies became leaky while claddings of some integer fuel rods were subjected to a substantial wear (fretting-corrosion up to 400μ in depth, nodular corrosion to a depth of 380μ — both kinds of the wear proceeded under spacer grids). In this case, the claddings revealed a higher hydrogen uptake and degradation of mechanical properties. During the wet storage in cans, spent FAs show general corrosion proceeding at the rate of 3–5 μ/ year. The paper points out that as applied to a long-term (up to 50 years) dry storage of RBMK-1000 spent nuclear fuel, consideration is to be given to the following degradation mechanisms proceeding in stainless steel fuel rod claddings and FA components: Oxidation of Zr-1%Nb alloy (fuel rod claddings, central channel) in the atmosphere of storage (air, nitrogen, argon and impurities). • Creep of a fuel rod cladding material. • Delayed hydride cracking (DHC) of a fuel rod cladding. • Corrosion induced cracking of FA components fabricated from stainless steels. The first results obtained in assessing the creep rate of standard fuel claddings effected by the internal pressure are under consideration.

The aim of this paper is to inform of the world’s specialists about the systematical approach to the radioactive waste management in Slovak Republic (Slovakia). In Slovakia are in operation six preassurized light water reactors at NPPs V-1, V-2 in Jaslovske Bohunice and reactors 2×440 MWe in NPP Mochovce at present. The installed elektro-capacity produced from these reactors is 2640 MW e total. The first experimental NPP in the former Czechoslovakia was NPP A-1. It had used natural uranium as a fuel and heavy water as a moderator and CO 2 as a cooling medium. The installed preformance of this NPP A-1 was 150 MW e and was in operation since 1972 to 1977. After two technological incidents former Czechoslovak government have decided to take out of this NPP A-1 from operation. According to former Soviet projects concerning for costructions of the above mentioned NPP in Jaslovske Bohunice Slovakia, to these project didn’t belong the projects for constructions of the technological units for the radioactive waste treatment, expect of the evaporating plant for the contaminated water concetration and storage tanks for the evaporated concentrates storage and storage possibilities for the contaminated solid waste storage. A comparable complicated situation was at NPP A-1, because all type of the radioactive waste streams content gama, beta and alfa radioanuclides as well. The second technological incident at NPP A-1 (with the some portion of the cooling medium CO 2 leaking form the reactor vessel out) caused contamination for the NPP A-1 operating spaces by alfa radionuclides and alfa radionuclides penetrated in all waste streams at storage tanks, pipelines, vessels, etc. as well. During the beginning of the 80’s were realized some states R and D tasks that solved the systematical approach to the radioactive waste treatment in former Czechoslovakia. The main aim of the first etap was solved the all problems connected for the radioactive waste, treatment that are formed during NPPs of the VVER-440 type operation. It was necessary to secure the production of the electro energie. The most seriously priority had the tasks, that solved of the evaporator concentrates and solid waste treatment, especially burnable, pretratment. Simultaneuosly were solved the problems connected with the proposals, designings and realisations of the disposals for the long term storage of the treated RAW in the Czech and Slovak regions. In this presentation I have tried to deliver main information, that give of the survey for RAW treatment and their long term safety storage at disposal system in Mochovec - Slovakia.

The BR3 reactor (10.5 MWe) at the Nuclear Research Center SCK•CEN was the first PWR plant installed in Europe and has been shut down in 1987. The BR3 reactor is from 1989 in a decommissioning stage and most of the spent fuel is presently still stored in the deactivation pool of the BR3 plant and has to be evacuated. The BR3 was used as a test-reactor for new fuel types and assemblies (Mixed Oxide (MOX) fuel, fuel rods containing burnable poison (Gd 2 O 3 ) and other types of fuels). Some fuel rods, having undergone a destructive analysis, are stored in different laboratories at the SCK•CEN. In total, the BR3 spent fuel comprises the equivalent of almost 200 fuel assemblies corresponding to some 5000 fuel rods. Beside the spent BR3 fuel, a limited number of spent fuel rods, with equivalent characteristics as the BR3 fuel but irradiated in research reactors outside Belgium and stored in other buildings at the SCK•CEN nuclear site, were added to the inventory of spent fuel to be evacuated. Various options such as reprocessing and intermediate storage awaiting final disposal were evaluated against criteria as available techniques, safety, waste production and overall costs. Finally the option of an AFR (away-from-reactor) intermediate dry storage of the BR3 and other spent fuel in seven CASTOR BR3 ® casks was adopted. As the SCK•CEN declared this spent fuel as radioactive waste, NIRAS/ONDRAF, the Belgian radioactive waste management agency became directly involved and the decision was taken to construct a small building at the Belgoprocess nuclear site for storing the CASTOR BR3 ® casks. Loading at the SCK•CEN followed by transport to Belgoprocess and storage is scheduled to take place at the end of 2001. The CASTOR BR3 ® cask weighing some 25 tonnes, consists of a monolithic body and has two independent lids with metal seals guaranteeing the long term leak-tightness of the cask. The CASTOR BR3 ® cask is designed for transport and the intermediate storage of at least 50 years. Although a defect of the leaktightness of a CASTOR BR3 ® cask is very unlikely to occur, an intervention scenario had to be developed. As no pool is present at the Belgoprocess nuclear site to unload the fuel, an innovative procedure is developed that consists of transferring the basket, containing the spent fuel, into another CASTOR BR3 ® cask. This operation can be performed in the hot cell of the existing storage building for high level waste at the Belgoprocess site.

The CASTOR ® BR3 cask has been designed and manufactured to accomodate irradiated fuel (U and MOX) from the BR3 test reactor at the nuclear research centre SCK/CEN in Dessel near Mol, Belgium, which is currently being dismantled. The CASTOR ® BR3 is designed as a Type B(U)F package for transport and will be licensed in Belgium. In addition, the CASTOR ® BR3 needs a license as a storage cask to be operated in an interim cask storage facility. To obtain these licenses, the cask design has to observe the international regulations for the safe transport of radioactive material as well as the special requirements for the cask storage. The CASTOR ® BR3 is a member of the CASTOR ® family of spent fuel casks, delivered by the German company GNB. In this way, the cask has such typical features as the following: • monolithic cask body made of ductile cast iron; • double-lid system consisting of primary and secondary lid for long-term interim storage of the fuel. This family of casks has been used for over 20 years for transport and storage of spent fuel. In this paper, the IAEA regulatory requirements for transport casks are summarized and it is shown by selected examples how these requirements have been converted into the cask design and the analyses performed for the cask. Finally, the cask features for an interim storage period of up to 50 years will be spotlighted. Main topics are the evaluation of the long term behaviour of selected cask components and the cask monitoring system for the surveillance of the leak tightness of the cask during the storage period.

Research studies by the Electric Power Research Institute (EPRI) established the technical and operational requirements necessary to enable the onsite cask-to-cask dry transfer of spent nuclear fuel. Use of the dry transfer system has the potential to permit shutdown reactor sites to decommission pools and provide the capability of transferring assemblies from storage casks or small transportation casks to sealed transportable canisters. Following an evaluation by the Department of Energy (DOE) and the National Academy of Sciences, a cooperative program was established between DOE and EPRI, which led to the cost-shared design of a dry transfer system (DTS). EPRI used Transnuclear, Inc., of Hawthorne, New York, to design the DTS in accordance with the technical and quality assurance requirements of the code of Federal Regulations, Title 10, Part 72 (10CFR72). EPRI delivered the final design report to DOE in 1995 and the DTS topical safety analysis report (TSAR) in 1996. DOE submitted the TSAR to the United States Nuclear Regulatory Commission (NRC) for review under 10CFR72 and requested that the NRC staff evaluate the TSAR and issue a Safety Evaluation Report (SER) that could be used and referenced by an applicant seeking a site-specific license for the construction and operation of a DTS. DOE also initiated a cold demonstration of major subsystem prototypes in 1996. After careful assessment, the NRC agreed that the DTS concept has merit. However, because the TSAR was not site-specific and was lacking some detailed information required for a complete review, the NRC decided to issue an Assessment Report (AR) rather than a SER. This was issued in November 2000. Additional information that must be included in a future site-specific Safety Analysis Report for the DTS is identified in the AR. The DTS consists of three major sections: a Preparation Area, a Lower Access Area, and a Transfer Confinement Area. The Preparation Area is a sheet metal building where casks are prepared for loading, unloading, or shipment. The Preparation Area adjoins the Lower Access Area and is separated from the Lower Access Area by a large shielded door. The Lower Access Area and Transfer Confinement Area are contained within concrete walls approximately three feet thick. These are the areas where the casks are located and where the fuel is moved during transfer operations. A floor containing two portals separates the Lower Access Area and the Transfer Confinement Area. The casks are located below the floor, and the fuel transfer operation occurs above the floor. The cold demonstration of the DTS was successfully conducted at the Idaho National Engineering and Environmental Laboratory (INEEL) as a cooperative effort between the DOE and EPRI. The cold demonstration was limited to the fuel handling equipment, the cask lid handling equipment, and the cask interface system. The demonstration included recovery operations associated with loss of power or off-normal events. The demonstration did not include cask receiving and lid handling; cask transport and lifting; vacuum/inerting/leak test; canister welding; decontamination; heating, ventilation, and air conditioning; and radiation monitoring. The demonstration test was designed to deliberately challenge the system and determine whether any specific system operation could adversely impact or jeopardize the operation or safety of any other function or system. All known interlocks were challenged. As in all new systems, there were lessons learned during the operation of the system and a few minor modifications made to ease operations. System modifications were subsequently demonstrated. The demonstration showed that the system operated as expected and provided times for normal fuel transfer operations. The demonstration also showed that recovery could be made from off-normal events.

This paper presents aspects of site decommissioning and clearance of a former fuel fabrication facility (development and production of fuel assemblies for research reactors and HTR) at Hanau (Germany). The main pathways for environmental contamination were deposition on soil surface and topsoil and pollution of deep soil and the aquifer by waste water channel leakage. Soil excavation could be done by classical excavator techniques. An effective removal of material from the saturated zone was possible by using advanced drilling techniques. A large amount of demolished building structure and excavated soil had to be classified. Therefore the use of conveyor detector was necessary. Nearly 100000 Mg of material (excavated soil and demolished building material) were disposed of at an underground mine. A remaining volume of 700 m 3 was classified as radioactive waste. Site clearance started in 2006. Groundwater remediation and monitoring is still ongoing, but has already provided excellent results by reducing the remaining Uranium considerably.

The Nuclear Energy Agency (NEA) Co-operative Programme for the Exchange of Scientific and Technical Information Concerning Nuclear Installation Decommissioning Projects (CPD) is a joint undertaking of a limited number of organisations, mainly from NEA member countries. The objective of the CPD is to acquire and share information from operational experience in decommissioning nuclear installations that is useful for future projects. The information exchange includes biannual meetings of the Technical Advisory Group (TAG) and supporting projects on topics. The TAG has decided to form a Task Group to review nuclear site restoration starting in March 2012 that involves nuclear operators, experts and regulators. The group is supported by the International Atomic Energy Agency (IAEA) that is leading similar work on legacy sites. Within NEA Counties, several nuclear sites are being restored for beneficial reuse. Restoration is normally considered the last activity in a sequence of decommissioning steps but increasingly the value of long-term planning and parallel remediation is being recognised. It is essential that regulators know that liabilities are well understood (well characterised) and there is adequate financial provision to carry on the remediation works. Operators are also learning that early intervention is needed to ensure prevention and minimisation of leaks and spills of radioactive and non-radioactive contaminants in order to reduce groundwater and soil contamination, thus reducing overall liabilities and ensuring protection of the environment. Early intervention needs to be guided by good practices that include adequate site characterisation, reliable conceptual models and defined goals. Currently most nuclear site restoration work takes place at the legacy nuclear sites. This work has emphasised the need for better clarity in terms of the regulatory expectations for site restoration. At other nuclear sites the drivers are less evident and there is a risk that land quality issues are overlooked. The aim of the Task Group is to share information on experiences, approaches and techniques for land quality management at selected nuclear sites to ensure risks to workers and the environment, costs and disruption to decommissioning programmes are minimised. The project will also highlight the successes and lessons to learn from experience of remediation that will be helpful to operational situations on nuclear sites. The paper will report on progress with analysis of national and site level questionnaires and early consideration of case studies. The questionnaires will provide a snapshot of the current status, issues and best practice with site restoration and the case studies will provide in depth illustrations of practice with nuclear site or remediation projects.

Since the 1950’s radioactive wastes from a number of laboratories have been stored below ground at the Hanford site, Washington State, USA, in vertical pipe units (VPUs) made of five 200 litre drums without tops or bottoms, and in caissons, made out of corrugated pipe, or concrete and typically 2,500mm in diameter. The VPU’s are buried of the order of 2,100mm below grade, and the caissons are buried of the order of 6,000mm below grade. The waste contains fuel pieces, fission products, and a range of chemicals used in the laboratory processes. This can include various energetic reactants such as un-reacted sodium potassium (NaK), potassium superoxide (KO 2 ), and picric acid, as well as quantities of other liquids. The integrity of the containments is considered to present unacceptable risks from leakage of radioactivity to the environment. This paper describes the successful development and full scale testing of in-ground augering equipment, grouting systems and removal equipment for remediation and removal of the VPUs, and the initial development work to test the utilization of the same basic augering and grouting techniques for the stabilization, size reduction and removal of caissons.

This study explores the development of a mechanical based system to retrieve single-shell tank waste at the Hanford site, located in Richland, Washington, without the addition of water. Out of 177 tanks at Hanford, there are 149 single-shell tanks (SST) and 28 double-shell tanks (DST). There are currently 67 SSTs that have leaked radioactive waste to the surrounding groundwater in the past or assumed to have questionable tank integrity. Leaking tanks continue to be a major concern as it was recently announced in February 2013 that six SSTs are leaking, even though they were believed to have been stabilized back in 2005. There are also several tanks with significant in-tank obstructions, such as air-lift circulators, where an arm-based retrieval would not be possible. All current methods of waste retrieval deployed involve the addition of significant amounts of water to generate a slurry that can be pumped out of the tank. This water, however, can exacerbate the leak and risk the potential of leaking more waste into the surrounding soil, subsequently contaminating the groundwater. It also reduces available waste storage space in DSTs, and increases the risk of dangerous buoyant displacement gas release events (BDGRE) in the headspace of DSTs. Focusing on leaking tanks, tanks with significant in-tank obstructions, and utilizing existing risers in Hanford’s tanks, this study evaluates commercially available dry technologies, such as augers and high-powered vacuums, for feasibility, safety, and efficiency of waste retrieval at Hanford.

There are approximately 56 million gallons (212km 3 ) of high level waste (HLW) at the U.S. Department of Energy (DOE) Hanford Site. It is scheduled that by the year 2040, the HLW is to be completely transferred to secure double-shell tanks (DST) from the leaking single-tanks (SST) via transfer pipeline system. Blockages have formed inside the pipes during transport because of the variety in composition and characteristics of the waste. These full and partial plugs delay waste transfers and require manual intervention to repair, therefore are extremely expensive, consuming millions of dollars and further threatening the environment. To successfully continue the transfer of waste through the pipelines, DOE site engineers are in need of a technology that can accurately locate the blockages and unplug the pipelines. In this study, the proposed solution to remediate blockages formed in pipelines is the use of a peristaltic crawler: a pneumatically/hydraulically operated device that propels itself in a worm-like motion through sequential fluctuations of pressure in its air cavities. The crawler is also equipped with a high-pressure water nozzle used to clear blockages inside the pipelines. The crawler is now in its third generation. Previous generations showed limitations in its durability, speed, and maneuverability. Latest improvements include an automation of sequence that prevents kickback, a front-mounted inspection camera for visual feedback, and a thinner wall outer bellow for improved maneuverability. Different experimental tests were conducted to evaluate the improvements of crawler relative to its predecessors using a pipeline test-bed assembly. Anchor force tests, unplugging tests, and fatigue testing for both the bellow and rubber rims have yet to be conducted and thus results are not presented in this research. Experiments tested bellow force and response, cornering maneuverability, and straight line navigational speed. The design concept and experimental test results are reported.

Up to the end of this decade, corresponding to the planned date of starting operation in the final disposal KONRAD for non-heat generating waste in Germany, a lot of efforts are needed to condition and package the radioactive waste in containers certified by BfS (Federal Institute for Radiation Protection). This waste is produced by public sector and industry as well as nuclear energy facilities, which result in more than a half of the actually declared quantity growing especially after the phase out decision of nuclear power production in Germany and the subsequent decommissioning of nuclear power plants. BAM (Federal Institute for Materials Research and Testing) acts as responsible authority on behalf of BfS for design testing under consideration of the KONRAD requirements. Within the assessment procedure of containers, BAM has to verify the application documents, including material qualification, container geometry, corrosion protection, leakage rate and operational and accidental loading, and to evaluate quality assurance measures. Besides the previous completed approvals for various containers such as steel sheet and cast iron box-shaped containers and concrete cylindrical ones, BAM is actually carrying assessments for other types such as cast iron cylindrical containers and “old” steel sheet box shaped ones. The so called “old” containers present already loaded containers without any KONRAD certification, currently stored at licensed interim storages. In the benefit of the container assessment, BAM operates design testing facilities for drop and fire tests which are also used for research objectives to improve and expand the evaluation methods such as research project ConDrop. The latter deals with numerical simulations and analyses tools for further precise predictions about unfavorable drop test scenarios, safety margins, and design sensitivities for steel sheet containers. Furthermore, during the assessment procedure, several specifications about the containers, the inventory to be disposed and the required safety level have been intensely debated by authorities and applicants. Based on its experience in qualifying containers, BAM has been commissioned to identify insufficiently specified aspects in the waste acceptance criteria and to propose clearer definitions with regard to the secondary regulations in the planning approval notice. This paper focuses on aspects of the contribution of BAM in certifying KONRAD waste containers. This will include the current state of design testing procedures, evaluation methods that are available or are being developed and the main topics within the KONRAD requirements being actually specified by BAM.

The paper deals with experience and techniques in the application of remotely controlled robotic devices for the dismantling of the A-1 NPP technological equipment during undergoing decommissioning process of the A-1 NPP, which is characterized by high level of radioactivity and contamination. For liquidation of the heavy water evaporator has been applied a mobile robotic system MT 80, which had been developed, designed and constructed as a general-purpose decommissioning equipment. The heavy water evaporator as a part of the NPP heavy water system is located inside the main production unit building in Room No. 220 where the inner surface contamination is from 10 1 Bq/cm 2 to the level of 10 3 Bq/cm 2 , dose rate up to 1.5 mGy/h and the feeding pipeline contained LRAW with high tritium content. The first step was the development of a work procedure with special focus on the elimination of activity and aerosols leaking into the environment. Special tooling was developed for application with the robot, such as hydraulic shears, circular saw, reciprocating saw, circular pipe cutter and a system for quick tool-change without direct intervention of the operators. Then, civil engineering modifications were made to the workplace and new technology was installed, including an efficient exhaust system. After draining-off the remains of LRAW and rinsing the pipeline, fragmentation of the pieces of equipment was started. The fragments are being deposited into drums which are transported in shielded containers to the decontamination facility, where their activity is reduced prior to their storage or further use. All operations are remotely controlled on basis of visual information from four cameras, with consistent radiation protection of the operators. This experience will be exploited in the continuation of work in Room No. 219, where the second identical heavy water evaporator is located. Nuclear Power Plant A-1 is situated in the locality of Jaslovske´ Bohunice. The A-1 NPP decommissioning has been organized according to distinctively licensed phases, which allows collecting input data for next stage during the running decommissioning stage. Since 2009 the Stage II of the A-1 NPP decommissioning has been in process.

The Active Storage Facility for Residuals (Aktives Lager fu¨r Ru¨cksta¨nde – ALfR) — an open-air controlled area — was used for the storage of radioactive waste (liquid and solid waste) at the site of the NPP Rheinsberg in the former GDR. Liquid waste was cemented and dumped for near surface final disposal. During the period of operation of the ALfR some incidents happened. Residues of evaporated cooling water of unknown amount (ca. 1–10 m 3 ), the drop-out from small leakages and the wash-off from contaminated open-air storage barrels contaminated the soil. Since 1997 the buildings of the ALfR have been decommissioned and the site will be remediated. From the beginning of decommissioning up to now the groundwater has been monitored. This paper shows different steps of decommissioning and explains the results of groundwater monitoring. Currently 33 sampling points are being used for sampling groundwater from different layers. The radiological parameters under examination are Cobalt-60, Cesium-137, Strontium-90 and Tritium. Non-radiological parameters like Bore, Potassium, Sodium, Nitrate, pH value and TOC (total organic carbon) are also being investigated. Since 1997 the activity concentration of Cobalt-60 and Tritium in the groundwater has been decreasing. The activity concentration of Strontium-90 had increased from 1997 to 2007. Since 2007 activity concentration of Strontium-90 has been decreasing. 350 m away from the ALfR, in the direction of groundwater flow, the Lake Stechlin is located in a nature protection area. The groundwater beyond the ALfR-site is not used as drinking-water and up to now there has been no detectable contamination of Lake Stechlin. Therefore there is no radiological impact on humans by contaminated groundwater. The hypothetic effective dose of Cobalt-60, Tritium and Strontium-90 resulting from postulated consumption of groundwater was determined using consumption rates according to the German Radiation Protection Ordinance. In case of using the groundwater as drinking-water the annual effective dose is 0,4 mSv/a at a postulated drinking-water well on the ALfR-site and 2,5 μSv/a at a postulated drinking water well in a distance of 142 m. Results of the observations: The observations of direction and velocity of the flow of the ground-water principally confirms the transportation model made at the beginning of decommissioning of ALfR. In vertical direction, the real spread of contamination in the groundwater is smaller than assumed. The dilution of Co-60 is higher than assumed. In contrast to the assumptions of the transportation model, Cesium-137 has a lower mobility.

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

Golder Associates (UK) Ltd, in partnership with Serco Assurance (Serco), undertook targeted cone penetration testing (CPT) of a series of six parallel on-site burial trenches on a nuclear licensed site in the UK. The form and concentration of radioactive and chemical material within the trenches is unknown. CPT was used to confirm the location of the bund walls and to characterise the material within the trenches. The CPT technique involves hydraulically pushing rods fitted with specialist characterisation “cones” into the ground. CPT generates no solid or liquid waste, and allowed rapid investigation of the trenches and bunds while ensuring exposure of radiation and contamination to workers was kept to a minimum, or removed in entirety. As a result of the unknown nature of radiological contamination within the trenches and the potential of introducing contamination into the inside of the CPT truck, a purpose-built extraction rig was constructed to withdraw the CPT equipment from the ground. Extraction of the equipment assumed airborne radioactive contamination was a potential hazard. The CPT locations selected for the investigation were based on non-intrusive geophysical survey work and a radiation survey, which identified the approximate location of the trenches, anomalies within the material (e.g. metallic objects), and radiation hotspots. The results of the geophysical surveys were overlaid with the original as-built drawings of the trenches. During the investigation the following investigation cones were deployed: • Resistance/friction cone, which determines geology through measurement of the friction on the sleeve of the cone and resistance on the tip of the cone; this was used to investigate the geology of the bunds. • Total gamma cone, which was used to obtain total gamma radiation results (in counts per second); • Groundwater sampler (BAT Sampler™), which was used to obtain a water sample from beneath the trenches; • Video cone, which was used to obtain a visual recording of the material within the trenches; and • Conductivity cone, which was used to investigate the presence of and depth to bodies of water below ground level (e.g. perched water, regional groundwater). The investigation collected essential data from an area of the site that had not previously been investigated, while minimising potential radiological exposure to all workers, and producing no investigation-derived waste. The investigation therefore confirmed the efficacy of cone penetration testing as a valid site investigation technique in a high hazard area. The data acquired from the CPT investigation and geophysical investigation also allowed boreholes to be sited in the bund walls between the trenches. Siting of the boreholes was a major risk for the project and presented a significant potential hazard. Golder has successfully used CPT on nuclear sites in the UK: as an innovative site investigation technique to facilitate radiological characterisation of an area with variable ground conditions; to investigate organic solvent plumes; and for the installation of blind tubes as a way of conducting subsurface radiation surveys and as part of a leak detection system (work in progress in partnership with Serco).

CANDU Stations are designed with significant amounts of carbon steel piping in the primary circuit. Although the primary coolant chemistry is such that carbon steel corrosion is minimized, nevertheless magnetite transport from the carbon steel surfaces to the steam generators is a significant issue leading to potential reduction in heat transfer efficiency in the steam generator. There are other contributors to the reduction of heat transfer efficiency such as divider plate leakage whereby some of the coolant short circuits the steam generator tubes and secondary side steam generator tube fouling. CANDU station operators have utilized a number of mitigating measures such as primary and secondary side mechanical and chemical tube cleaning, and divider plate refurbishment to counter these problems but these are all expensive and dose intensive, It is therefore very important to establish the relative contribution of each source to the overall heat transfer degradation problem so the most effective results are obtained. Tube removal and laboratory assessment of the oxide loading is possible and has been utilized but at best it provides an incomplete picture since typically only short lengths of tubes are removed — most often from the hot leg and the tube removal process adversely impacts the primary side oxide integrity. Kinectrics Inc. has developed, qualified and deployed Oxiprobe, a highly mobile non destructive technology able to remove and quantify the deposited oxide loading on the primary surfaces of steam generator tubes. The technology is deployed during shutdown and provides valuable, direct information on: • Primary oxide distribution within the steam generator; • Oxide loading (thickness of oxide) on the primary surfaces of steam generator tubes; • Oxide composition and radiochemical characterization. The End Effector probe can reach either side of the straight section of the steam generator U tube but as currently designed it is unable to be deployed in the U-tube region. The current technology is able to visit 4 tubes simultaneously. The technology is Code classified as a Class 6 fitting by the Canadian Nuclear Safety Commission and registered by the Ontario Technical Standards and Safety Authority as a pressure boundary retaining system. Although the application of the technology to date has been applied to steam generator tubes, in principle it can be applied to any heat exchanger tube, vertical or horizontal. This paper will describe the system, the qualification program for its deployment as well as some actual field results. The applicability of the technology for PWR steam generators is also addressed.

In 2000, Nuvia Limited was contracted to carry out the decommissioning of a former Active Handling Building A59 on the United Kingdom Atomic Energy Authority (UKAEA) site at Winfrith in the UK. This is in support of UKAEA’s mission, which is to carry out environmental restoration of its nuclear sites and to put them to alternative uses wherever possible. Recently UKAEA has been reorganised and responsibility for the site lies with Research Sites Restoration Limited (RSRL) with funding provided by the National Decommissioning Authority (NDA). Following major decommissioning operations the main containment building structure and the two suites of concrete shielded caves were demolished between June 2006 and March 2007 leaving just the base slab for final removal and the site remediation operations undertaken. The base slab contained a quantity of encast, internally contaminated items including more than 100 steel mortuary tubes set up to 6.6m deep into the slab. At the outset it was suspected that some leakage of radioactive contamination had occurred into the ground although the precise location/s of the leakage was unknown. As a result the scope of the work required the underlying soil to be carefully monitored for the presence of radioactive contamination and, if detected, its remediation to an end state suitable for unrestricted use without planning or nuclear regulatory controls. These latter operations form the basis of this paper, which reviews some of the significant tasks undertaken during the process and describes the waste monitoring procedures utilised on the concrete and soil debris. Extensive dewatering was required to support the removal of the deeper mortuary tubes and the impact this had upon the operations and associated excavations will be described. Further, the demolition of an external active effluent tank and excavation and monitoring of the surrounding soils due to the presence of significant local contamination will be a key feature of the paper. A number of significant problems that were encountered during the operations will also be identified with a narrative about how these arose and were subsequently overcome. The use of Nuvia’s Groundhog™ system, a gamma radiation ground surveying and global positioning system, together with a well defined sampling grid enabled the footprint of the base slab to be surveyed and subsequently remediated to an agreed standard by the end of 2008 to allow infilling with non-calcareous soil ahead of final landscaping as the last step for completion of the project. One area of particular significance to the remediation process has been the use of office-based contaminated land assessment tools including ReCLAIM, a Microsoft Excel spreadsheet based tool used to assess current and future impacts of radiological contamination at nuclear licensed sites. This tool is particularly recommended to others working on similar projects.

This study involved challenging nuclear grade high-efficiency particulate air (HEPA) filters under a variety of conditions that can arise in Department of Energy (DOE) applications such as: low or high RH, controlled and uncontrolled challenge, and filters with physically damaged media or seals (i.e., leaks). Reported findings correlate filter function as measured by traditional differential pressure techniques in comparison with simultaneous instrumental determination of up and down stream PM concentrations. Additionally, emission rates and failure signatures will be discussed for filters that have either failed or exceeded their usable lifetime. Significant findings from this effort include the use of thermocouples up and down stream of the filter housing to detect the presence of moisture. Also demonstrated in the moisture challenge series of tests is the effect of repeated wetting of the filter. This produces a phenomenon referred to as transient failure before the tensile strength of the media weakens to the point of physical failure. An evaluation of the effect of particle size distribution of the challenge aerosol on loading capacity of filters is also included. Results for soot and two size distributions of KCl are reported. Loading capacities for filters ranged from approximately 70 g of soot to nearly 900 g for the larger particle size distribution of KCl.

At the Los Alamos National Laboratory Plutonium’s Facility (TA-55), various isotopes of plutonium along with other actinides are handled in a glove box environment. Data collected from 700 unplanned openings in the glovebox environment indicate that 40% of these ruptures are glove failures, openings in the glove caused by degradation of the mechanical properties over time, e.g. exposure to chemicals and nuclear materials. The primary means of minimizing these types of glove failures is through a robust glove inspection program. Through a collaborative effort with Getinge La Calhe`ne, we have beta-tested their In-situ Glove Leak Tester (GLT2) in an active ventilation glovebox environment. This instrument tests glovebox gloves in place and uses pressure decay methodology, to tests up to four gloves at a time, in less than 5 minutes per test, and provides documentation of each test. While the GLT2 has been successful in static pressure enclosures that are possible in the pharmaceutical industry, the uninterruptible active negative pressure gloveboxes used in the nuclear industry present a new set of issues. In the following report, these issues are addressed and the limits of this technology defined, with the end result being that waste generated from glove failures has been minimized.

Sellafield is a former Royal Ordnance Factory used since the 1940’s for the production and reprocessing of fissile materials. Leaks and spills from these plants and their associated waste facilities has led to radioactive contaminated ground legacy of up to 20 million m 3 . Consideration of land contamination at Sellafield began in 1976, following discovery of a major leak from a waste storage silo. Over the past three decades there has been a programme of environmental monitoring and several phases of characterization. The latest phase of characterization is a £10million contract to develop second generation conceptual and numeric models. The Site Licence Company that operates the site has been subject to structural changes due to reorganizations within the British nuclear industry. There has also been a change in emphasis to place an increased importance on accelerated decommissioning. To address these challenges a new contaminated land team and contaminated land and groundwater management plan have been established. Setting and measuring performance against challenging objectives is important. The management plan has to be cognizant of the long timescales (ca. 80 years) for final remediation. Data review, collation, acquisition, analysis, and storage is critical for success. It is equally important to seize opportunities for early environmental gains. It is possible to accelerate the development and delivery of a contaminated land and groundwater management plan by using international experts.