Symposium Thermal treatment of radioactive wastes .../file/Thermal... · Thermal treatment of...

Thermal treatment of radioactive wastes: research, development and demonstration

Venue: The Centre, Birchwood Park, Warrington, WA3 6YN (Tel: 01925 282 940)

12 December 2013

Start Duration (h:min)

Programme

09:00 01:00 Arrival / Refreshments

10:00 00:10Neil Hyatt (UoS) & Sean Morgan (Sellafield Ltd)Welcome & Opening remarks

10:10 00:25Jonathan Shatwell (NDA)Disposal of Thermally Treated ILW

10:35 00:25Bill Lee (Imperial College London)Understanding Wasteforms from Thermal Treatment Methods.

11:00 00:25Olivier Drouvot (Areva)R&D Programmes and Policy Within the CEA-Areva Joint Vitrification Laboratory

11:25 00:25Keith Witwer (Kurion)Stabilization and Volume Reduction of ILW using Kurion's GeoMelt® In-Container Vitrification™ Technology

11:50 00:20Maxence Lorencki (IRSN)Waste management: expectations of thermal treatment

12:10 00:20Robert Worth & Alex Theodosiou (University of Manchester)Thermal Treatment of Irradiated Graphite Waste: Treatment and 'From Core to Capture'

12:30 01:15 Poster session & Buffet Lunch

13:45 00:20Neil Hyatt (University of Sheffield)Wasteform science in support of deployment of thermal treatment technologies (or know thy enemy).

14:05 00:20Charlie Scales (NNL)Development and design of the HIP process for the immobilisation of Pu residues

14:25 00:25Jan Deckers (Belgoprocess)The innovative plasma tilting furnace for treatment of radioactive and problematic chemical waste

14:50 00:20 Break & Refreshments

15:10 00:25Tim Johnson (Tetronics International)Tetronics Plasma Vitrification Technology in the Management and Conditioning of Nuclear Waste

15:35 00:25Rob Scott (Energy Solutions)Process optimisation of the Savannah River Defence Waste Processing Facility (DWPF) Joule Heated Melter by bubbler modification

16:00 00:25Nick Gribble (NNL) Benefits achieved by the Vitrification Test Rig in support of WVP

16:25 00:05Sean Morgan (Sellafield Ltd) & Neil Hyatt (UoS)Summary and conclusions, presentation of poster prize.

16:30 Close of Meeting

Symposium

Thermal Treatment Symposium - Dec 2013 - 2 - Abstracts Talks & Posters 1 (15)

Talk 1 at 10h10

Disposal of Thermally Treated ILW

Jonathan Shatwell

RWMD NDA

NDA/RWMD undertakes disposability assessment of packaging proposals for radioactive waste. Wastes are assessed against the requirements of the GDF design, safety cases and waste package specifications. RWMD has assessed ILW glass, HIP and HLW glass for disposability with the results discussed. Glass/ceramic wasteforms may offer an alternative to cementitious Wasteforms, potentially providing advantages to the NDA estate for packaging of some wastes. Glass/ceramic wasteforms, their application and understanding of implications for disposal are developing. RWMD can continue providing disposability advice on thermal treatment proposals allowing risk management for Waste Producers and NDA. New waste packages can be introduced in the plans for a GDF through change control. Currently RWMD change control process is examining the compatibility of ILW glass wasteforms in a cementitious ILW concept.

Talk 2 at 10h35

Understanding Wasteforms from Thermal Treatment Methods.

W. E. Lee

Centre for Nuclear Engineering, Imperial College London

The UK has some difficult wastes, such as those in the Sellafield Legacy Ponds and Silos (LP&S), for which there is no current management option. Novel thermal methods, such as plasma vitrification and Joule heating In-container melting, provide a route for dealing with some of the poorly characterised and heterogeneous LP&S wastes. This presentation will review recent developments in both non-thermal and thermal techniques highlighting knowledge gaps and research needs and opportunities.

Talk 3 at 11h00

R&D Programs and Policy within the CEA-AREVA Joint Vitrification Laboratory (LCV)

O. DROUVOT (1), L. PARADIS (2), E. CHAUVIN (3)

(1) AREVA NC Marcoule LCV, France (2) CEA Marcoule LCV, France(3) AREVA NC Paris La Défense, France

Waste management is a key issue for the reprocessing industry; furthermore, vitrification is considered as the reference for nuclear waste management. In order to further improve and strengthen their historical cooperation in high temperature waste management, the CEA, R&D organization, and AREVA, Industrial Operator, decided, in September 2010, to create a


Joint Vitrification Laboratory within the framework of a strategic partnership. The main objectives of the CEA-AREVA Joint Vitrification Laboratory (LCV) are (i) support AREVA’s activities, notably in its La Hague plants and for new projects, (ii) strengthen the CEA’s lead as a reference laboratory in the field of waste conditioning. The LCV is mandated to provide strong, innovative solutions through the performance of R&D on processes and materials for vitrification, fusion and incineration, for high, intermediate and low level waste. The activities carried out in the LCV include academic research on containment matrices (formulation, long-term behavior), and the improvement of current technologies/development of new ones in lab-scale to full-scale pilot facilities, in non-radioactive and radioactive conditions, including modeling and experimental tools. This presentation will focus on the programs and policy managed within the LCV, as well as the means employed by the CEA and AREVA to meet common short-,mid- and long-term challenges, from a scientific and industrial point of view. Among other things, we discuss the technical support provided for the La Hague vitrification facilities on hot melter and CCIM technologies, the start-up of new processes (decommissioning effluents, UMo FP) with CCIM, the preparation of future processes by means of an assessment of new technologies and containment matrices (improved glasses, ceramics, etc.), as well as incineration/vitrification for organic and metallic mixed waste or metallic fusion. The close relationship between the R&D teams and industrial operators enables the LCV to propose attractive waste management solutions, with appropriate schedules and optimized development costs, making allowance for R&D constraints, engineering requirements and the industrial environment.

Talk 4 at 11h25

Stabilization and Volume Reduction of ILW using Kurion's GeoMelt® In-Container Vitrification™ Technology

K. Witwer and B. Campbell

Kurion Inc, 3240 Richardson Road, Richland, WA 99352, USA

A huge inventory of higher activity waste needs to be addressed in the UK, with more than 20,000 m3 of plutonium contaminated materials (PCM) waste stored on the Sellafield Site alone. The traditional method of cement grout encapsulation is incompatible with a significant portion of these wastes, leaving the UK with a pressing need to identify and implement alternate methods for treating these remaining waste stockpiles. Kurion is actively proposing its GeoMelt In-Container Vitrification, or ICV, Technology to meet this need. The GeoMelt technology has been in existence for three decades, and has treated over 26,000 tonnes of hazardous waste, half of this being radioactive material. Projects have included plutonium cleanup in Australia, heavy metals and persistent organic pollutants in the USA, asbestos and PCB’s in Japan, and many others. ICV provides a simple yet robust and flexible method for treating PCM and other ILW waste. Unlike other thermal treatment technologies, the process has a wide operating envelop due to its inexpensive single use melt vessel/disposal container, and the size and shape of the waste treated is limited only by the size of the customer’s chosen disposal container. Demonstration of the effectiveness of the process has been provided to Sellafield and the NDA on several


occasions over the last four years. In 2009, vitrification of a two part 50wt% PCM and 35wt% SIXEP Magnox Sludge was performed at Kurion’s 5-acre test facility in Richland, Washington, USA. The PCM waste was staged in the vitrification vessel and after being partially vitrified, a wet magnox sludge was pumped into the melt vessel and fully treated, producing a full container of glass product. In a follow-on test, a two part mixture of pile fuel cladding silo waste and SIXEP sand/clino was treated. Mass balance calculations showed average retention factors for radioactive surrogates of greater than 99% for Ce, Sr, Eu, and 99% for Cs in one test and 68% in the other. In 2010, the same demonstration system was moved to a new GeoMelt test facility located at AMEC’s Birchwood offices. This second phase was to demonstrate treatment of problematic Sellafield wastes, namely those ILW wastes contained in a skip/box/drum. These wastes in particular contain a large fraction of metallic constituents, including uranium, magnesium, aluminum, and mild and stainless steel. Testing was performed using a simulated ponds solids skip, with similar dimensional ratios as a typical Sellafield skip. The skip and its materials were completely and successfully vitrified, producing a two-part ferrous metal and vitreous product. As previously seen, Ce, La, Eu were all homogeneously contained with the glass at +99% retention. Cs was retained at 89% within the glass and Ru was retained at nearly 100% within the primarily iron metal phase. Efforts continue on this path towards treatment of Sellafield and other UK higher activity wastes. In 2014, the same GeoMelt Demonstration System is scheduled to be installed at the National Nuclear Laboratory facility located on the Sellafield site. Initially, demonstration of active waste treatment will be performed at the facility, with the end goal of thermally treating a significant portion of PCM and other ILW not suitable for treatment using the traditional cement encapsulation process.

Talk 5 at 11h50

Waste management: expectations of thermal treatment

M. Lorencki (1), V. Wasselin (2)

[email protected] [1], [email protected] [2]

Waste management: expectations of thermal treatment IRSN (the French Institute of Radioprotection and Nuclear Safety) is in charge of reviewing the Long Lived-High Level Wastes (LL-HLW) geological repository project managed by ANDRA (the French National Radioactive Waste Management Agency) in support of the Nuclear Safety Authority (ASN). Amongst other IRSN paies particular attention to the performance of confinement of the nuclear wastes packages, to the release of the radionuclides from the wastes and the associated related uncertainties. Guidelines for the conditioning of French radioactive wastes that are not acceptable for disposal in surface facilities, mainly because of their high content of long-lived isotopes, are established in the safety guide for the deep geological disposal by the nuclear safety authority. The physical transformations of the wastes, required by the production of the nuclear wastes package, have to be designed with a view to achieving a high level of intrinsic safety of the package. In a disposal installation, the package should keep a good level of performance in the various conditions prevailing during and after its emplacement, while the facility is still opened. The retrievability of the package should be guaranteed at all time during this phase. Long-term post-closure safety should


also be taken into account in the design of the packages. Assessment of the long term disposal properties of nuclear wastes requires an understanding of the chemical interaction of the wastes with the surrounding environment. For the disposal phase the design of the nuclear wastes packages will be assessed regarding the following properties: - low chemical reactivity with the package environment (corrosion, oxidation in air, etc.) in order to limit package degradation, - absence of significant inner physical and chemical modification in the conditions of storage in order to preserve its favorable long term confinement properties, - low gas generation (radiolysis, chemical evolution, corrosion) to prevent risk of package degradation through pressure or of explosion hazard, - capacity to reduce dispersion of radioactive matter in incidental or accidental situations (package drop, flooding, etc.), - sufficient mechanical strength (long term stacking, etc.). All these points are directly and strongly impacted by the presence of organic matter in the nuclear wastes. Furthermore, the presence of this organic matter allowed the development of a very rich variety of microorganisms. A good waste processing should involve that this processing let as low organic matter as possible in the final waste package. Some of particular waste contains a lot of plastic matter, in association with alpha emitters (in most of the case: plutonium) and other radionuclides. These waste mainly come from the nuclear fuel production industry. Today, they do not have an identified way of processing. The thermal treatments, as vitrification, of these particular waste represent a very encouraging way to decrease the associated risks. This kind of waste processing especially allowed to turn organic matter into ashes and to immobilize radionuclides into a safe matrix. IRSN is strongly aware of the progress of this particular waste processing."

Talk 6 at 12h10

Characterisation and Thermal Treatment Development of UK Irradiated Graphite Waste Robert Worth1*, Alex Theodosiou1*, Lorraine McDermott1, Greg Black1, Abbie Jones1 and Anthony Wickham2

1Nuclear Graphite Research Group, School of Mechanical, Aerospace and Civil Engineering, University of Manchester, M13 9PL 2Visiting Professor, University of Manchester, M13 9PL

A substantial volume of the UK nuclear waste inventory consists of irradiated nuclear graphite, which is currently designated for a deep geological intermediate level waste repository. Given current difficulties with the MRWS voluntary process for selecting a suitable Geological Disposal Facility site, making informed decisions on how best to dispose of large quantities of graphite waste within the UK remains under review. The quantity and migration of radioisotopes within irradiated graphite and appropriate methods for decontamination are of interest, with a view to better understanding irradiated graphite. The behaviour of short- and long-lived isotopes in irradiated graphite is significant for disposal and treatment options, with particular emphasis on evaluating the mobility of the higher activity radioisotopes such as 3H, 14C and 60Co.

Research has been performed within the EU Euroatom FP7 CARBOWASTE consortium on decontamination of irradiated graphite via thermal treatment, which has been applied to


various sources of irradiated graphite waste. This research utilises data and collaborations from within CARBOWASTE and captures work carried out within an IAEA CRP, Treatment of Irradiated Graphite to Meet Acceptance Criteria for Waste Disposal, to drive forward thermal treatment experiments on UK irradiated graphite. A further Technology Strategy Board programme of research entitled Thermal Treatment of Reactor Graphite: ‘From Core to Capture’ has also been initiated to investigate an alternative means of graphite disposal. These research programmes ensure the evaluation and application of thermal treatment processes to UK irradiated graphite.

Talk 7 at 13h45

Wasteform science in support of thermal treatment strategy

Neil Hyatt

Immobilisation Laboratory, Department of Materials Science and Engineering, University of Sheffield.

Thermal treatment technologies offer the potential to efficiently treat a wide spectrum of legacy radioactive wastes with variable and heterogeneous chemical, physical and radiological inventory. The wasteform products from thermal treatment, including ceramic, slag, glass, and metallic materials, will expand the range of materials consigned to the geological disposal facility. In this presentation, I will discuss the need to understand the fundamental materials science of wasteforms from thermal treatment, with a view to underpinning and optimising the product, treatment and geological disposal strategy, based on an understanding of relevant mechanisms and processes. Three examples will be given from our research, in which fundamental materials science has advanced our understanding and optimisation of the product, treatment and geological disposal strategy:

• Solid solutions based on Gd2Zr2O7 for Pu disposition • Thermal treatment of PCM wastes using blast furnace slag • Dissolution mechanisms of waste glasses at hyperalkaline pH

The importance of developing a basic science programme in parallel with any future thermal treatment strategy will be discussed.

References: Hyatt et al., J. Nucl. Mater, 444 (2014) 186; Corkhill et al., Int. J. Appl. Glass Sci., 6 (2013) 1; Utton et al., J. Nucl. Mat., 435 (2013) 112; Utton et al., J. Nucl. Mat., 442 (2013) 33; Reid et al., J. Solid. State Chem., 191 (2012) 2.


Talk 8 at 14h05

Development and design of the HIP process for the immobilisation of Pu residues

C R Scales

NNL, Workington Laboratory, Havelock Road, Workington, CA14 3YQ

NNL in collaboration with ANSTO and on behalf of Sellafield Ltd have developed a hot isostatic pressing (HIP) based process for the immobilisation of Pu residues currently stored on the Sellafield site. Using HIP, the production of ceramic and metal based wasteforms provides an opportunity to convert these residues into a form suitable for ongoing storage and ultimate disposal.

Talk 9 at 14h25

The innovative plasma tilting furnace for treatment of radioactive and problematic chemical waste

Jan Deckers

Belgoprocess, Gravenstraat 73, B-2480 Dessel

The operation and maintenance of nuclear power plants, the nuclear fuel cycle in general, research laboratories and pharmaceutical, medical and industrial facilities generate low-level radioactive waste which, along with the historical radioactive waste from past nuclear activities, needs to be treated and stored, awaiting final disposal. Plasma technology offers a very effective way of treating this waste with a high volume reduction factor (VRF), free from organics, liquids and moisture, and meets without doubt the acceptance criteria for safe storage and disposal. By means of a plasma beam of approximately 5000°C,the inorganic materials are melted into a glassy slag, containing most of the radioactive isotopes while the organic material is gasified and afterwards oxidized in an afterburner and purified in an off-gas cleaning system.

Talk 10 at 14h25

Tetronics’ Plasma Vitrification Technology in the Management and Conditioning of Nuclear Waste

Tim Johnson, David Deegan

Tetronics International

Tetronics International is the world leader in providing DC plasma arc technology and has over 40 years of experience of providing waste treatment solutions to companies around the world in a wide variety of industries. Tetronics DC plasma arc technology has a natural versatility that makes it an ideal choice to address the specific challenges of nuclear intermediate level waste, whilst also transforming the waste into a glassy form of high stability and the lowest practical volume.


This paper will present an overview of the work Tetronics has undertaken with the nuclear industry over the last ten years through a range of demonstration and R&D activities covering a number of different wastes. Outline details will also be presented of the current TSB-funded programme on the nuclearisation of Tetronics’ DC plasma arc technology, which is being conducted jointly by Tetronics and Costain.

Talk 11 at 15h35

The process optimisation of the Savannah River Defence Waste Processing Facility (DWPF) Joule Heated Melter by bubbler modification

Rob Scott

EnergySolutions EU Ltd Loweswater Pavilion, Westlakes Science and Technology Business Park, Moor Row, Cumbria. UK. CA24 3HX. In the US, Savannah River Remediation (SRR) LLC acquired the liquid waste contract at the Savannah River Site (SRS) in the summer of 2009. In order to achieve the main goal of the contract, the closing of High Level Waste (HLW) tanks, it was necessary to improve the processing rate of sludge wastes through the SRS liquid waste facilities. To attain the increased production rate, four unique specification bubblers were installed in the Defence Waste Processing Facility (DWPF) Joule Heated Melter in September 2010 to agitate the glass pool. The design and fabrication of the four bubblers was accomplished through SRR critical subcontractor EnergySolutions LLC. The waste processing rate has increased as the bubblers have created more convection/mixing within the glass melt pool, thereby increasing the heat transfer to the melt pool cold cap. This increase in heat transfer has provided higher melt rates, a consistent waste product and an increase in vitrified canister production needed to meet the contract goals.

Talk 12 at 16h00

Benefits achieved by the Vitrification Test Rig in support of WVP

Nick Gribble

NNL

The Vitrification Test Rig (VTR) is a full scale replica of the Sellafield highly active Waste Vitrification Plant (WVP). It was built post operation of WVP to provide a means of increasing the waste loading, glass production rate and plant availability and to provide product quality data to support operation of WVP with different feed compositions, a new base glass formulation and wider, less constraining operating envelopes for the calcination and vitrification processes. An overview of the test work and results are discussed and the hard and softer benefits provided by this full scale development facility are outlined


Poster Abstracts

Hot Isostatically Pressed Wasteforms for Future Nuclear Fuel Cycles

Ewan Maddrell

National Nuclear Laboratory, Havelock Road, Workington, Cumbria, CA14 3YQ

The anticipated renaissance of nuclear energy will lead to a massive increase in the global inventory of spent nuclear fuel. A once through fuel cycle may not be a credible option because the demands on global uranium reserves will be unsustainable, making recycling a central requirement. Whatever fuel cycle is specified, from a continuation of conventional PUREX to processes involving multiple separations of minor actinides, there will be waste arisings that require immobilisation. Given that these are the main inter-generational legacy from a fuel cycle, wasteforms with a longevity appropriate for the immobilised wastes should be seen as no less important a product than recycled material with an energy value. This paper compares and contrasts hot isostatically pressed wasteform options for a variety of aqueous based fuel recycling scenarios. Initially the options for immobilisation of the full spectrum of fission products and minor actinides will be presented. This will be extended to include scenarios in which, for example, plutonium which is no longer attractive for reuse can be immobilised. A central factor leading to the choice of advanced fuel cycle will be to make the most efficient use of repository capacity in terms of the tonnage of spent fuel disposed. The primary constraint is likely to be thermal, with the decay heat setting the minimum spacing of waste packages. To decrease this spacing requires either lengthy above surface storage, or the partitioning of heat generating nuclides such as 90Sr and 137Cs. Partitioned nuclides will still require immobilisation but wasteforms will differ from those required for a non-partitioned stream, for example a novel glass formulation or a hollandite/perovskite derivative of Synroc. The effect of further separations and appropriate wasteforms will be covered. The paper will also look at whether streams separated perforce due to the extraction cycle can be recombined as a means of simplifying the immobilisation requirements – for example non-recyclable minor actinides and noble metal fission products could be suitable for co-immobilisation with rare earth element fission products, in another Synroc derivative based on zirconolite. In all cases it is emphasised that a full cost benefit analysis will need to be conducted to assess the value added through the extra costs incurred by extensive separation processes and the immobilisation of multiple streams. The paper will additionally describe potential methods by which gaseous nuclides can be sequestered and converted into stable wasteforms. Specifically 14C and 129I will be considered, with the former being targeted to synthetic carbonate minerals and the latter being immobilised in a sodalite. These options require a rethink of existing head-end off-gas management through the use of dry, solid scrubbers. Chemical and process engineering challenges that must be addressed for successful wasteform fabrication will be outlined. The author’s preference for the use of hot isostatic pressing as a powerful and versatile wasteform consolidation technology will be explained. Most simply, HIP provides the capacity to consolidate all of the most active fuel cycle wastes using a single bank of HIP units.


Laser scabbling of cementitious materials for nuclear decommissioning

B. Peach (1), M. Petkovski (1), D.L. Engelberg (2) and J. Blackburn (3)

(1) University of Sheffield, Department of Structural and Civil Engineering, UK, (2) University of Manchester, School of Materials, UK, (3) TWI Ltd, Cambridge, UK

The contamination depth of concrete structures in nuclear facilities is dependent on a number of parameters, such as concrete composition and curing condition, as well as environmental factors, such as contamination type, exposure temperature and time. In the majority of cases contamination is limited to the top millimetres of the exposed concrete, and by physically removing the surface layer the volume of radioactive waste can be reduced. Mechanical and high-pressure water-jet scabbling techniques have long been used in civil engineering, but large reaction forces and secondary waste make these methods difficult to implement for nuclear decommissioning. The application of high power laser on a concrete surface causes fragments to be explosively ejected from the surface. Laser transmission through optical fibres and absence of any reaction force allows for remote handling. For laser scabbling it was found that aggregate type dominates the effectiveness of the material removal process. The objective of the work reported in this paper is to investigate the effect of concrete composition and moisture content on the laser scabbling process, with a focus on identifying the mechanism behind laser scabbling. A 5 kW Yb-fibre laser (wavelength = 1070+/-10 nm) was used to investigate the effects of concrete composition and moisture content on laser scabbling. The specimen range included hardened cement paste, mortar, concrete and aggregate (rock) specimens pre-exposed at two moisture conditions. All tests were carried out on 100x100x50 mm cubes, with laser interaction times between 0.5 and 40 seconds. Infrared and high-speed photography was used to record the laser scabbling process. Surface temperatures were found to vary depending upon material composition. Temperature profiles obtained were compared against differential thermal and thermo gravimetric analyses (DTA/TGA) to identify thermal reactions potentially responsible for the laser scabbling process. The tests showed that the composition has a significant effect on the depth of material removal. The mechanical properties of coarse aggregates were seen to affect scabbling rather than their chemical properties. The use of pulverized fuel ash as a cement replacement material also had a great effect on the material removal. Water content had little effect on material removal however debris size was reduced with greater moisture content. Results of this study suggest two potential scabbling mechanisms, based on either (1) thermal stresses developed as a result of severe temperature gradients caused by the high heating rates and low thermal conductivity, or (2) rapid pore pressure development as a result of vaporization of free, adsorbed and chemically combined water, as well as gases produced by thermal decomposition.


Dissolution of UK high level waste glass under simulated hyper-alkaline conditions of a co-located geological disposal facility

C. L. Corkhill, N. J. Cassingham, P. G. Heath and N. C. Hyatt

Immobilisation Laboratory, Department of Materials Science and Engineering, University of Sheffield.

The University of Sheffield, Sheffield, S1 3JD, UK The aim of this study was to develop an understanding of the dissolution of UK HLW glass (MW+25%) in a co-located HLW and ILW geological disposal facility, using a model system. The effect of a high pH, Ca(OH)2-saturated solution at 50°C (the predicted temperature of the GDF several thousand years after closure) was investigated over a period of 168 days using powdered, semi-static dissolution experiments. For comparison, the UK HLW glass powder was also dissolved in initially pure water. In order to investigate the effect of surface area to volume ratio (SA/V) effects on the dissolution of UK HLW glass in saturated Ca(OH)2, monolith dissolution experiments were also conducted, including a preliminary analysis of the alteration layer formed. Dissolution of the glass in the presence of high concentrations of Ca (> 200 mg L-1) was an order of magnitude lower than in water. Dissolution of Si did not occur until a Ca:Si ratio of <2 was achieved. The mechanism of dissolution was shown to involve the incorporation of Ca into the hydrated surface (initial, incubation regime), the precipitation of C-S-H phases, including a range of compositions in the C-(N)-(A)-S-H and M-S-H systems (intermediate regime) and the precipitation of C-S-H phases (the residual regime). These phases formed a layer on the surface of the glass ~ 60 microns in thickness. Thermodynamic analysis and consideration of the CaO-SiO2-H2O phase diagram suggested that the rate-limiting step of glass dissolution in Ca-rich solutions is Ca-Si “quasi”-equilibrium, involving the precipitation of C-S-H phases, which change in chemical composition as a function of solution chemistry. In low SA/V ratio experiments, the dissolution progressed only to the initial, incubation regime, resulting from fewer glass surface sites for Ca incorporation. Overall, these results indicate that Ca and Si in solution play an important role in the long-term durability of UK HLW in Ca-rich solutions, which may have important implications for the disposal of nuclear waste glass in, or near, a cementitious geological disposal facility.

Advanced ceramic wasteforms for future fuel cycles

Daniel J. Bailey

Immobilisation Science Laboratory, Department of Materials Science and Engineering , The University of Sheffield, UK

Future expansion of civil nuclear energy will result in a substantial inventory of spent nuclear fuel. Sustainability of uranium resource will require both fuel recycle, using advanced separations methods, and enhanced fuel burn up. A substantial expansion of the civil nuclear energy programme is expected to require a concomitant expansion of the GDF footprint to accommodate the resulting HLW products (which must be well separated due to radiogenic decay heat). It is not clear that such an expansion is sustainable or feasible in technological, social, or political terms. This project will explore options for advanced


wasteforms that support a strategy of decay-storage prior to final disposal, enabling sustainable use of GDF resource coupled to advanced fuel cycles.

The Use of Hot Isostatic Pressing in the Vitrification of Clinoptilolite Ion Exchange Material

P. G. Heath, R. J. Hand, N. C. Hyatt

Immobilisation Science Laboratory, Department of Materials Science and Engineering , The University of Sheffield, UK

The zeolite clinoptilolite is used in the U.K. as an ion exchange material for the removal of aqueous radioactive caesium and strontium. A large inventory of this material, in terms of both activity and volume, exists from the operation of the site ion exchange plant on the Sellafield site. This material is classified as intermediate level radioactive waste. No accepted processing or final disposal pathway for this radioactive waste stream currently exists. This work has studied the potential application of Hot Isostatic Pressing (HIPing), in transforming spent clinoptilolite into a vitreous waste form. This method of processing has several potential advantages, including lowering the melting point of the mineral, eliminating volatile losses, decreasing residual porosity whilst providing a self contained final waste form. The addition of various additives has been studied in relation to product quality and densification. Caesium retention during processing has been studied and PCT glass leaching tests have been used to measure waste form dissolution and caesium loss whilst in contact with aqueous solutions.

Thermoluminescence of household glass as forensic tool for accidental dosimetry.

S. Thornber, A. Gandy, and N. Hyatt.

Department of Materials Science and Engineering, University of Sheffield, Mappin Street, Sheffield, S1 3JD

This poster discusses the methodological development of the thermoluminescence (TL) technique as a forensic tool for measuring the possible radiation damage to household glass that may occur upon an accidental release of radiation to the environment. Thermoluminescence is the act of thermally stimulating structural relaxations within a material. When ionization of a material occurs due to incident radiation, electron-hole pairs are created that become trapped at defect points. For recombination to take place, energy is required. The TL technique provides energy (as heat) to the ionized material, which results in a detectable emission of photons. The resulting spectrum is representative of the absorbed radiation. A typical TL spectrum shows multiple peaks, each representative of different types of intrinsic or extrinsic defects, known as ‘traps’. Thus, this technique can be used to determine the radiation dose applied to a material. The effect of UV-radiation on household glass materials was quantified using the TL methodology, in order to define the residual background signal. Following this, the effect of radiation dose, time and temperature on the TL intensity were investigated. Radiation dose was provided via a 60Co irradiation source, with doses ranging from 0 – 20 Gy, irradiation durations ranged from 0 –


72 hours and temperatures of 75°C and 165°C were investigated. It was found that the residual radiation from UV light accounted for ~5 % of the background signal, which is not usually taken into account in TL experiments [1]. It was shown that complete recombination of unstable traps occurred after 12 hours and stability was achieved after 24 – 48 hours, indicating the earliest point for calibration of the data for dosimetry applications. It was also found that defect stability was dependent on temperature (with greater stability achieved at high temperature), and suggested that these high temperature defects can be used for reliable dosimetry analysis. [1] Clark R.A., Rodertson J.D., Intrinsic dosimetry of glass containers: a potential interrogation tool for nuclear forensics and waste management, Radioanal Nuclear Chemistry, 2012, 296, p. 663-668 "

Extending the Waste Vitrification Plant Melter Lifetime

Michelle Cowley and Carl Steele

Sellafield Ltd, Seascale, Cumbria, CA20 1PG Historically melter operations have been a process bottle neck in the production of high level waste containers in the Waste Vitrification Plant (WVP) at Sellafield. The melters are subject to particularly harsh conditions which stress the material and as a result of this the melters fail after a relatively short period of operation. In 2004 a package of technical and engineering work commenced through the vitrification assistance programme (VAP) in conjunction with AREVA to improve melter performance and WVP operations. An optimised melter design was delivered which has improved melter lifetime by up to 400% with one melter producing a record 231 containers compared to the previous record of 215 containers produced by AREVA. Other business benefits have also been delivered through this equipment improvement work including reduced rebuild frequency and reduced volumes of highly active technical waste.

Durability and Aging Studies of Pu-238 and -239 Containing Calcium Phosphate Ceramic Waste-forms

S. Fong (1), B. Metcalfe (1), R. Scheele (2) and D. Strachan (2)

(1) Atomic Weapons Establishment, Aldermaston, Berkshire RG7 4PR, UK. (2) Pacific Northwest National Laboratories, Richland, WA, USA

A calcium phosphate ceramic waste-form has been developed at AWE for the immobilisation of chloride containing wastes arising from the pyrochemical reprocessing of plutonium. In order to determine the long term durability of the waste-form, aging trials have been carried out at PNNL. Ceramics were prepared using Pu-239 and -238, these were characterised by PXRD at regular intervals and Single Pass Flow Through (SPFT) tests after approximately 5 yrs.


X-ray Chemical Imaging of Nuclear Explosion Glass Residues

D. Crean (1), D. Bailey (1), M. Stennett (1), T. Blackwell (1), C. Corkhill (1), R. Tappero (2), B. Ravel (2) and N. Hyatt (1) (

1) Immobilisation Science Laboratory, Department of Materials Science and Engineering, University of Sheffield, UK. (2) Photon Sciences Directorate, Brookhaven National Laboratory, Upton, New York, USA.

In order to understand the formation mechanisms and performance of heterogeneous glasses, spatially resolved information on their key element chemistry is advantageous. Microfocus X-ray chemical imaging combines X-ray fluorescence and absorption spectroscopy using synchrotron radiation to resolve elemental distribution and speciation at high spatial resolution, with a source size of approximately 7 µm x 12 µm. In this study, we characterise chemical heterogeneity in trinitite, a vitreous material produced by detonation of the first nuclear weapon at the Trinity Test in New Mexico, USA, 1945. Contained within trinitite are natural minerals fused in the extreme heat of the nuclear explosion, with remnants of the nuclear device and test apparatus. By probing the distribution of Ca, Cu, Fe and K alongside Fe and Cu chemical states, we provide new information on the formation mechanisms of this complex glass. Fe was discovered to be distributed heterogeneously throughout the matrix, but with a constant redox state of Fe(II). This is evidence of uniform reduction of Fe(III) present in natural minerals, and suggests that material from the native sand was taken up into the reducing environment of the explosive fireball before coagulating to form the trinitite melt. Trace levels of Cu were found decorating vesicle and external surfaces, indicative of Cu condensation. This puts a lower bound on the melt temperature as greater than 2860 K, and the preservation of Cu features on surfaces is evidence of rapid cooling of the melt. Ca and K were distributed according to differing extents of localised melting of mineral grains, and evidence of diffusive mixing of K was observed. In contrast, the preservation of sharp Ca gradients indicates that these minerals were not mixed with the glass, suggesting late phase incorporation. Our study demonstrates the utility of X-ray spectroscopic imaging for revealing fine scale chemical details in complex heterogeneous amorphous materials, and how this data can be used to provide models of formation processes in extreme environments. Such methodology would be useful for the chemical characterisation of slags and complex waste forms to understand performance and underlying formation mechanisms which may govern the distribution and chemistry of key radionuclides.


Thermal Treatment of Plutonium Contaminated Materials

L.Boast

University of Sheffield, Department of Materials Science and Engineering, Mappin St Sheffield, S1 3JD "The projected UK Plutonium contaminated material (PCM) waste volume is >30000m3 with 70% arising at Sellafield. The current baseline treatment is supercompaction / cement encapsulation. However, a BPEO study for Sellafield Ltd. highlighted “concerns regarding the composition of the conditioned wasteform and… stability during long term storage”. Thermal treatment, i.e. in-container or plasma vitrification, was identified as “the main alternative” technology with “advantages in terms of… stability of final product and improved volume reduction”. Although proof of concept studies by Sheffield University and others have demonstrated PCM compatibility with thermal processes, a fundamental understanding of waste incorporation reactions and the impact of waste inventory on product quality remains to be established. This generic understanding is clearly critical to successful technology deployment. The aim of the project is to develop an understanding of waste / matrix interactions during thermal treatment of PCM wastes. The specific objectives are to understand: • The sequence of reactions leading to waste digestion in thermal treatment • Control of Pu surrogate partitioning between off gas and wasteform slag /metal components (using Ce, U and/or Th as a Pu surrogate). • Selection of glass forming additives to optimise waste digestion and wasteform quality More broadly, the project will contribute to accelerating the acquisition of knowledge and experience required to support NDA in deploying thermal technologies as a national asset for ILW treatment, this will lead to a scientifically underpinned product and process, with cost savings of >£360 M for disposal and several £100M for waste treatment."

Plasma Vitrification of Intermediate Level Waste

Elizabeth Wright (1), Sashini Premathilaka (2)

(1) Costain Ltd; (2) Tetronics International

Costain and Tetronics are working together to develop a treatment method for Intermediate Level Waste using a DC Plasma system. Although DC plasma treatment is established for hazardous waste treatment (as outlined in the Keynote Speech by Tetronics) it is not yet in commercial use for ILW. Supported by a grant from the Technology Strategy Board, we will design, build and demonstrate operation of a ‘nuclear aligned’ DC plasma vitrification furnace for processing ILW. Currently the project is addressing the challenges of nuclearizing the furnace itself and envisaging the plant required to support the furnace. The poster will give a basic overview of the furnace, show some of the key challenges faced by the project and highlight some of the benefits to the industry from the project.


Thermal recovery of calcium niobate pyrochlore system CaLnFe0.5Nb1.5O7 (Ln= La, Nd, Sm, Gd, Ho, Yb) under In-situ irradiation

L.D.Casey (1), M. Stennett (1), N.C.Hyatt (1), E. Maddrell (2), N.J. Zaluzec (3) and K.R. Whittle (1)

(1) Immobilisation Science Laboratory, Department of Materials Science and Engineering , The University of Sheffield, UK (2) National Nuclear Laboratory Limited. Registered office: 5th Floor, Chadwick House, Warrington Road, Birchwood Park, Warrington, WA3 6AE (3) Materials Science Division, Argonne National Laboratory, 9700 South Cass Avenue, Argonne, IL 60439, USA

Crystalline ceramics offer advantages for the immobilisation of highly active wastes. This is mainly due to location of radioactive species being mapped and locked within a complex structure. The calcium niobate pyrochlore structure, Fd3m, structure is of interest due to the fact that rock formations, which naturally contain up to 30 wt % of uranium and thorium1, often adopt this crystal structure. However, material with this structure can become amorphous from the radioactive decay of radioactive species retained, dominated by alpha recoil damage. Thus it is important to look at model ceramic systems and their behaviour when subjected to irradiation. This work looks at the system CaLnFe0.5Nb1.5O7, Ln= La, Nd, Sm, Gd, Ho, Yb which adopts the pyrochlore structure. This has been confirmed through characterisation using; X-ray diffraction, scanning electron microscopy, semi-quantitative energy dispersive X-ray analysis and transmission electron microscopy at the University of Sheffield. Along with lattice parameter determined using Rietveld refinement of neutron diffraction data collected using the high-resolution powder diffractometer (HRPD) at ISIS, Rutherford Appleton Laboratory. In-situ irradiation, to investigate a materials tolerance to radiation damage, has been completed at the IVEM-Tandem facility, Argonne National Laboratory. This in-situ transmission electron microscope, operating at 300keV, is coupled with a Danfysik ion implanter implanting 1 MeV Kr2+ ions with a flux of 6.125 x 1011 ions cm-2 sec-1, into a sample placed on a variable temperature stage. The use of variable temperature allows the kinetics of a materials recovery from damage to be examined, allowing the temperature at which the rate of recovery in the material equals the rate of damage, Tc, to be determined. For the samples Ca(La,Sm,Ho,Yb)Fe0.5Nb1.5O7 the critical temperatures have been calculated as 682, 755, 1102 and 1269 K, respectively. Ideally, for a waste form, the structure will not become amorphous under any radiation dose allowing the exact location of species to be known. This could be met through a radiation resistant material or utilising the storage temperature to ensure the recovery rate is always greater than the damage rate. For these samples it can be seen that the temperature required is elevated above that which is desired for geological disposal facilities, at approximately 373K2, so these would not be a suitable candidates for the safe disposal of high level waste. However, this information can be used to support a wider field of work looking at the chemical dependence of the critical temperature of ceramic waste forms. Future work will look at what is responsible within these structures for what appears to be a systematic change in the response to irradiation. 1 Lumpkin, G.R. and Ewing, R.C. (1988) Phys. Chem. Minerals (1988) t 6: 2-20 2 NDA Technical Note no. 11261814

Participants

Thermal Treatment Symposium - Dec 2013 Final Master 1 (3)

Organisation First Name Last Name Role

Sellafield Limited Clifford William Addison Technical directorateSheffield Daniel Bailey PhD StudentPerma-Fix Paula Barry ConsultantNDA RMWD Rebecca Beard Research ManagerNNL Phil Bearman Strategic Business

Development ManagerSellafield Ltd Richard Blackham Technical Development

manager (Effluent & Encapsulation Plant)

Sheffield Luke Boast PhD StudentUCLan Gary Bond Associate DeanGeoRoc Ltd Dan Brew Technical ManagerSheffield Gordon Brown Research Group ManagerMinistry of Defence Christine Bruce Resins dispolsal Project

ManagerWarwick Susan Burrows Senior Research FellowKurion Brett Campbell Vice President

Environment Agency Colin Campbell Nuclear Waste Assessor

Sheffield Laura Casey PhD Student

CEA Marcoule - France

Patrice Charvin Engineer

AREVA Eric Chauvin R&D Project ManagerSellafield Rav Chunilal Support to Higher Active

Waste on Sellafield SiteSheffield Nick Collier Postdoctoral Research

AssociateSellafield Ltd Nick Connelly Process EngineerSheffield Claire Corkhill Research Fellow

Sellafield Ltd Michelle Cowley Vitrification Technical Support

AMEC Jamie Culshaw Project EngineerBelgoprocess Jan Deckers Technical Business Manager

Huddersfield Steve Donnelly Researcher, Professor and Dean

AREVA Olivier Drouvot vice-director LCV (Joint Vitrification Lab)

NNL Martin Dutton Interim Packaging Assessment Manager

UCLAN Harry Eccles Professor of Nuclear Materials

UCLan John Emmot PhD StudentHSE-ONR Chris Fisher HM Inspector - ONRAWE Shirley Fong Programme ManagerNuclear FiRST DTC Robert Frost PhD StudentSheffield Amy Gandy Research Associate

Sheffield Laura Gardner PhD Student

Nuvia Limited David Garlick Principal Consultant

Participants



Nuclear Technologies

Vicky Gaskin Principal Consultant

Sheffield Fergus Gibb ResearcherAMEC Andrew Green Senior Consultant, Applied

Chemistry and Waste Processing

Environmmmment Agency

John Green Nuclear Regulator

NNL Nick Gribble Technical Lead VitrificationNDA-RWMD Richard Guppy Senior Packaging

Assessment ManagerNNL Mike Harrison Technical Specialist

Sheffield Paul Heath PhD Student

Sellafield Ltd Grace Hudson Process EngineerLLWR Richard Hunter Waste Characterisation

Services ManagerImperial College Charlie Hutchison PhD StudentSheffield Neil Hyatt Professor of Nuclear

Materials ChemistryRAE/NDA Research Chair

Sellafield Ltd Mike James Head of TechnicalTetronics Tim Johnson Technical Director

Manchester Abbie Jones Lecturer in Nuclear Engineering decommission

Galson Sciences Ltd Jenny Kent Senior Consultant

Imperial College Bill Lee Director, Centre for Nuclear Engineering

Costain Bryony Livesy Head of Research and Technology

IRSN Maxence Lorencki Nuclear safety expertNNL Ewan Maddrell Technical SpecialistSellafield Ltd Neil Marsland Site Strategy DevelopmentFiberstone Products Ltd

Lee Marston Technical Manager

Sellafield Ltd Charles Mason Technical and Engineering Manager

Sellafield Ltd Andrew Milsom RWMD Interface Technical Manager

Sellafield Ltd Sean Morgan Technical ManagerNDA Darrell Morris Technical Decommissioning

ManagetSellafield Paul Mort Decommissioning - Technical

ManagerSellafield Ltd Kevin Newland Programme Technical

ManagerStudsvik UK Ltd Chloe Newton Waste Management

ConsultantNDA-RWMD Christiano Padovani Senior Research ManagerSheffield Ben Peach PhD StudentSellafield Ltd Andrew Pearson Technical EngineerGen II / nucleargraduates

Alexis Petrides EC&I Engineer

Participants



Imperial College Dimitri Pletser PhD StudentTetronics International

Sashini Premathilaka Process Engineer

Sheffield John Provis ProfessorEcologica Resources Ltd

Julian Ravest Director

Nuvia Gareth Read ConsultantNSG James Rudd Business Development

ManagerNNL Charlie Scales Technical LeadEnergySolutions INS Rob Scott Director of Projects

NDA-RWMD Jonathan Shatwell Packaging Assessment Manager

NNL Rick Short Vitrification Manager

AWE Andy Spurgeon Integrated Waste StrategySellafield Ltd Jon Squire Technical SpecialistSellafield Ltd Carl Steele Technology ManagerSheffield Martin Stennett University Teacher

RSRL Winfrith Joseph Stephenson Waste Policy and Compliance

Manchester Alex Theodosiou PDRANNL Olivia Thompson Technology ManagerSheffield Stephanie Thornber PhD Student

Sellafield Ltd Matt Towers Graduate Trainee PhysicistQueen's University Belfast

Gareth Tribello Lecturer in Atomistic Simulation

HSE-ONR Bill Turner Nuclear Safety InspectorSellafield Ltd Philip Vesey process engineerUCLan Stephen Walker PhD ResearcherSheffield Sam Walling PhD Student

JFNL -James Fisher nuclear

Stewart Watkins Operations Manager

Magnox Bill Westall Section HeadPhoenixsolutions John Williams Plasma Application Specialist

Kurion Keith Witwer Technical DirectorManchester Robert Worth PhD Student

Costain Elizabeth Wright Senior Mechanical Engineer

Sellafield Ltd Tina Wylie Head of Technical Strategy for WEDD on Sellafieldsites

Thermal treatment of radioactive wastes: research, development and demonstration

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