Academic literature on the topic 'Decommissioning'

This PhD project studied the caesium and strontium removal capacities of six nanoporous materials. Indeed, it was linked in with the SIXEP plant present at Sellafield site, UK, which is currently using natural clinoptilolite, originally mined from the Mojave Desert, to remove radioactive caesium and strontium ions from wastewater effluents. At the moment the feeds entering the SIXEP plant correspond to the best working conditions for clinoptilolite. As the feeds are moving toward more acidic pH and complex solutions and as clinoptilolite works at neutral pH, new nanoporous materials are required. This project had been divided into sub-projects, which all had the same purpose of finding the best material for caesium and strontium removal under complex solutions. The first part of the PhD focused on the synthesis of CST, SNT, ETS-4 and ETS-10 materials following literature reviews. The two last materials studied were provided by Fortum Power and Heat Oy for CsTreat material and by Sellafield Ltd for clinoptilolite material. The second part of the PhD consisted in studying the exchange capacities for caesium and strontium ions for each material studied under batch conditions using several water compositions relevant to Sellafield Ltd. Each water composition had different competing ions or pH and were used to classify the materials between them. From that it was found that Clinoptilolite stays the best materials for caesium and strontium removal. The third part of the project was the gamma irradiation of the materials. This irradiation had an aim to check the stability of the structure of the studied materials under gamma irradiation. The materials were exposed to four different gamma doses which were 0.5; 1; 3 and 6 MGy, the last one corresponding to the maximum dose were structural changes could happen. This revealed that all the materials, except ETS-4 material, were stable under gamma irradiation and that no structural changes were observable. The fourth part of the project was an extension of the third part in which the materials were Cs exchanged before being gamma irradiated to a dose of 6 MGy. This showed that the caesium exchange process is not affected by radiation. Finally, the unexchanged materials were alpha irradiated using a 15 MeV He2+ beam at the Dalton Cumbria Facility. Two irradiations were performed on CsTreat, CST, ETS-4, ETS-10 and clinoptilolite material. It was found that all the materials were damaged by alpha irradiation and displayed amorphous framework To resume, it appeared that clinoptilolite still stays one of the best materials for the removal of caesium and strontium ions in a wide range of wastewater composition. Additionally, it was discovered that ETS-10 material could be a good alternative to clinoptilolite material as its structure is stable under gamma irradiation and it displayed good exchange capacities when competing ions were present in the feeds. CST could also be a substitute as it shows good ion exchange properties under a broad range of pH. SNT material was great for strontium removal while CsTreat was working well for caesium removal. Even though ETS- 4 revealed good ion exchange capacities they were lower than clinoptilolite, ETS- 10 or CST materials.

Tritium is a by-product of civil nuclear reactors, military nuclear applications, fusion programmes and radiopharmaceutical production. It commonly occurs, though not exclusively, as tritiated water (HTO) or organically-bound tritium (OBT) in the environment but may exist as other forms in nuclear-related construction and fabrication materials. During the lifetime of nuclear sites (especially those involving heavy water) tritium becomes variably incorporated into the fabric of the buildings. When nuclear decommissioning works and environmental assessments are undertaken it is necessary to accurately evaluate tritium activities in a wide range of materials prior to any waste sentencing. Of the various materials comprising UK radioactive wastes, concrete and metal account for approximately 20% of the total weight of low level waste (LLW) and 12% and 35% of the total weight of intermediate level waste (ILW). Proper sampling and storage of samples are significant factors in achieving accurate tritium activities. The degree of loss of 3H and cross-contamination can be significantly reduced by storing samples in an air/water tight container in a freezer (-18°C). The potential for tritium contamination is dependent on the 3H form. Most 3H loss originates from tritiated water which is easily exchanged with atmospheric hydrogen in the form of water vapour at room temperature. However, the loss of more strongly bound 3H, produced in-situ in materials by neutron activation, is not significant even at room temperature. Such tritium is tightly retained in materials and does not readily exchange with water or diffuse. In nuclear reactor environments tritium may be produced via several neutron-induced reactions, 2H(n,g)3H, 6Li(n,a)3H, 10B(n,2a)3H and ternary fission (fission yield <0.01%). It may also exist as tritiated water (HTO) that is able to migrate readily and can adsorb onto various construction materials such as structural concrete. In such locations it exists as a weakly-bound form that can be lost at ambient temperatures. Bioshield concretes present a special case and systematic analysis of a sequence of sub-samples taken from a bioshield core (from UKAEA Winfrith) has identified a strongly-bound form of 3H in addition to the weakly bound form. The strongly bound 3H in concrete is held more strongly in mineral lattices and requires a temperature of >850°C to achieve quantitative recovery. This more strongly retained tritium originates from neutron capture of trace lithium (6Li and potentially 10B) distributed throughout minerals in the concrete. The highest proportion of strongly bound 3H was observed in the core sections closest to the core. Weakly bound tritium is associated with water loss from hydrated mineral components. Tritium is retained in metals by absorption by free water, hydrated surface oxidation layer, H ingress into bulk metal and also as lattice-bound tritium produced via in-situ neutron activation. Away from the possible influence of neutrons, the main 3H contamination to metals arises from absorption and diffusion via atmospheric exposure to the HTO. Here contamination is mainly confined to the metal surface layer. The tritium penetration rate into metal surfaces is controlled by the metal type and its surface condition. Where metals are exposed to a significant neutron flux and contain 6Li, 7Li and 10B then in situ 3H production will occur which may propagate beyond the surface layer. In such cases tritium may exist in two forms namely a weakly bound HTO form and a non-HTO strongly bound form. The HTO form is readily lost at moderate temperatures (~120°C) whereas the non-HTO requires up to 850°C for complete extraction.

In this report decommissioning of offshore petroleum platforms have been investigated. It treats decommissioning in general, the process of a typical project. A variety of suitable lifting vessels have been presented, and some concepts of removal have been evaluated. Decommissioning is important to go through with because of the environment and the use of the area after the petroleum activities ceases. Other ocean users benefit from the decommissioning because the area can be utilized when it is opened for ordinary traffic. The environment will benefit from the cessation of production because of fewer spills to sea or risk of it and being restored to a state as close to possible to original. Sometimes it is found acceptable to leave facilities behind, partially or wholly, and the marine life in the area can actually benefit from this in some cases. A decommissioning project consists of the elements planning and approval, topside and jacket preparatory work, topside removal, jacket removal, transport, and onshore dismantling and recycling. There are two main concepts of removal, reverse installation and single lift. The elements of the project are the same whichever of the concepts are used. The vessels to be used need large lifting capacities in either concept. Especially designs that lift topsides and jacket in single lifts can improve the efficiency in the projects. Several lifting vessels have been presented and a few concepts were reviewed regarding future requirements. The concepts were platform removal using only a traditional HLV, topside removal using HLV and jacket removal using buoyancy tanks, and removal using the new lifting vessel design Pieter Schelte. It is found that all the concepts can remove fixed platforms, but Pieter Schelte was especially well-suited. When Pieter Schelte is used, personnel will be exposed to less risk because more work is done onshore. Onshore work is in general more inexpensive, faster, cleaner, and safer. This gives a better HSE profile to the project as well as being more cost-efficient. On the downside, the vessel is not available before 2013. In the meantime the alternative with the HLV and buoyancy tanks can gain some more experience. This alternative scored the same as removal with reverse installation and is therefore not so successful in this comparison. There are however uncertainties involved and the solution with the buoyancy tanks are still interesting. The challenges in decommissioning are to make it less expensive, less time-consuming, and safer than today as well as keeping the environmental issues satisfactory. This can probably be reached by making the process more standardised to some extent, making use of new technologies like new vessels, doing less work at the offshore site, carrying out less lifts offshore, and do a lot of the dismantling work onshore. Pieter Schelte can to a great extent be the solution to these challenges.

Ice pigging is a novel technology using thick ice slurry (a two-phase mixture of ice crystals and freezing point depressant solution) to clean the internal surfaces of pipes or ducts; this mixture displays semi-solid characteristics. When pumped through a pipe, the slurry adopts plug flow, forming an 'ice pig'; slip occurs at the interface with the pipe walls generating high shear stresses; thus able to mobilise and remove sediment residing at the pipe wall. Ice pigs are able to navigate demanding topologies such as vertical falls, diameter changes, orifice plates, heat exchangers, and intrusive instrumentation; they provide a method of removing fouling without the need for dis-assembly, reducing valuable down-time, labour intensive pipe work dismantling, and subsequent manual cleaning. Many decades of nuclear activity here in the UK have produced unique and difficult challenges that require solving at Sellafield, the UK's nuclear waste reprocessing site. The drive to produce plutonium for atomic weapons during the 1950's, with very little foresight towards how the wastes and facilities would be dealt with, has brought about significant challenges. As these facilities are nearing the end of their design lives, the time has come to assess methods of treating these wastes and decommissioning the facilities in a safe, controlled, and cost-effective manner. Ice pigging is one of many technologies being assessed for such a task; this thesis details specific areas of application where experimental work has been conducted. Experimental work conducted in this thesis has: developed a method of characterising the ice pig's sediment removal performance compared to simple water flushing, assessed the ice pig's ability to remove representative sediments, assessed the ice pig's suitability for removing sediment from heat exchangers to restore thermal performance, and analysed the rate of percolation of the driving fluid through the ice pig body, such that the suitability of the ice pig for separating fluids can be established.