Composition Changes and Future Challenges for the Sellafield Waste Vitrification Plant

2009 ◽  
Vol 1193 ◽  
Author(s):  
A. Riley ◽  
S. Walker ◽  
Nick R. Gribble
Keyword(s):  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Feed Composition ◽  
Power Stations ◽  
Highly Active ◽  
Long Term Storage ◽  
Waste Storage ◽  
Waste Vitrification ◽  
Future Challenges ◽  
Future Work

AbstractThe Sellafield Waste Vitrification Plant (WVP) immobilises highly active liquors produced during reprocessing of spent nuclear fuel by bonding the fission products as metal oxides into a borosilicate glass matrix. This provides a stable and durable waste form suitable for safe long term storage and ultimate disposal.WVP was commissioned with feed from reprocessing of Magnox reactor fuel. This material is relatively low in fission product content per tonne of fuel, but contains significant Al and Mg from fuel cladding. WVP also routinely treats a blended feed made from a mixture of Magnox and Oxide reprocessing products. The Oxide fuel from Light Water Reactor (LWR) and Advanced Gas Cooled (AGR) power stations is of higher burnup and contains more fission products per tonne of fuel, also Gd and other process additives. Blending allows 25% incorporation of waste oxides by weight in glass to be achieved routinely.Recent programmes of development work in WVP have been aimed at increasing incorporation rates for these feeds, to reduce the number of waste containers produced for disposal. Work has also focussed on increasing the throughput of WVP, to more rapidly treat current stocks of liquid reprocessing waste, both by increasing the feed rate and by improving the lifetime of key components to improve plant availability.Future challenges for WVP include flowsheet changes to treat historic stocks of reprocessing wastes containing high U, Fe and Cr. Washout of solids from the base of waste storage tanks in preparation for decommissioning is also likely to give high Mo feeds. Development of flowsheet and glass formulation to accept these changes in feed composition will be a key objective of future work.

Initial Investigation Into the Vitrification of High Molybdenum Solids in Borosilicate Glass

2009 ◽  
Vol 1193 ◽  
Author(s):  
Barbara F. Dunnett ◽  
Nick R. Gribble ◽  
Andrew D. Riley ◽  
Carl J. Steele
Keyword(s):  
Phase Separation ◽  
Borosilicate Glass ◽  
Spent Nuclear Fuel ◽  
Storage Tanks ◽  
Waste Storage ◽  
Waste Vitrification ◽  
Durability Testing ◽  
Glass Formulation ◽  
Selection Of ◽  
Scanning Electron

AbstractSellafield Ltd operates a Waste Vitrification Plant (WVP) to immobilise the arisings from the reprocessing of spent nuclear fuel. Washout of solids from the base of waste storage tanks in preparation for decommissioning is likely to produce feeds enriched in molybdenum to the WVP. Vitrification of such feeds in the borosilicate glass formulation currently used by the WVP for vitrification of reprocessing waste has been investigated to determine the maximum achievable loading of MoO3.The vitrification of molybdenum in the absence and presence of reprocessing waste was studied. A number of glasses were manufactured in the laboratory containing various waste loadings. The resultant glasses were examined both visually and under the scanning electron microscope for the presence of any phase separation. Additional aluminium was added to the glasses manufactured in the absence of reprocessing waste to improve the durability of the glass. In borosilicate glass containing 3.5 wt% Al2O3 the onset of a molybdenum phase separation was observed in glasses containing 2.6 wt% MoO3. In the presence of Magnox reprocessing waste, phase separation was observed when the product contained >3.8 wt% MoO3. Soxhlet durability testing of a selection of the glasses manufactured was carried out. The Soxhlet durability of glasses in the absence of phase separation was good.

Summary of Results From the Series 2 and Series 3 NNWSI Bare Fuel Dissolution Tests

1987 ◽  
Vol 112 ◽  
Author(s):  
C. N. Wilson
Keyword(s):  
Spent Nuclear Fuel ◽  
Fine Particles ◽  
Fission Products ◽  
Test Site ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Nuclear Waste Storage ◽  
Waste Storage ◽  
Rate Of Increase ◽  
In Series

AbstractThe Nevada Nuclear Waste Storage Investigations (NNWSI) Project is studying dissolution and radionuclide release behavior of spent nuclear fuel in Nevada Test Site groundwater. Specimens prepared from pressurized water reactor (PWR) fuel rod segments were tested for multiple cycles in J-13 well water. The Series 2 tests were run in unsealed silica vessels under ambient hot cell air (25°C) for five cycles for a total of 34 months. The Series 3 tests were run in sealed stainless steel vessels at 25°C and 85°C for three cycles for a total of 15 months. Selected summary results from Series 2 and Series 3 tests with bare fuel specimens are reported.Actinide concentrations tended to saturate and then often decreased during test cycles. Uranium concentrations in later test cycles ranged from 1 to 2 μg/ml in the Series 2 Tests versus about 0.1 to 0.4 μg/ml in Series 3 with the lowest concentrations occurring in the 85°C tests. Formation of a calciumuranium-silicate phase identified as uranophane in the 85°C Series 3 Tests is thought to have limited uranium concentration in these tests. Americium-241, Pu-239 and Pu-240 activities measured in filtered solution decreased to less than 1 pCi/ml in the 85°C tests. Preferential release of fission products Cs, I, Sr and Tc, and activation product C-14, was indicated relative to the actinides. Tc-99 and Cs-137 activities measured in solution after Cycle 1 increased linearly with time, with the rate of increase greater at 85°C than at 25°C. Continuous preferential release of soluble fission products is thought to result primarily from the dissolution of fine particles of fission product phases concentrated on grain boundaries.

Vitrification of high molybdenum feeds in the presence of reprocessing waste liquor

2013 ◽  
Vol 1518 ◽  
pp. 21-39
Author(s):  
Rick Short ◽  
Barbara Dunnett ◽  
Nick Gribble ◽  
Hannah Steel ◽  
Carl James Steele
Keyword(s):  
Borosilicate Glass ◽  
Solubility Limit ◽  
Test Rig ◽  
Waste Liquor ◽  
Highly Active ◽  
Waste Storage ◽  
Waste Vitrification ◽  
Processing Line ◽  
Active Waste ◽  
Glass Formulation

ABSTRACTAt Sellafield, the Post Operational Clean Out (POCO) of solids from the base of the highly active waste storage tanks, in preparation for decommissioning, will result in a high molybdenum stream which will be vitrified using the current Waste Vitrification Plant (WVP). In order to minimise the number of containers required for POCO, the high molybdenum feed could be co-vitrified by addition to reprocessing waste, using the borosilicate glass formulation currently utilised on WVP. Co-vitrification of high molybdenum feeds has been carried out using non-active simulants, both in the laboratory and on the Vitrification Test Rig (VTR) which is a full scale working replica of a WVP processing line.In addition, a new borosilicate glass formulation containing calcium has been developed by NNL which allows a higher incorporation of molybdenum through the formation of a durable CaMoO4 phase, after the solubility limit of molybdenum in the glass has been reached. Vitrification of the high molybdenum feed in the presence of varying quantities of reprocessing waste liquor using the new glass formulation has been carried out in the laboratory. Up to ∼10 wt% MoO3 could be incorporated without any detrimental phase separation in the product glass, but increasing the fraction of reprocessing waste was found to decrease the MoO3 incorporation. Soxhlet and static powder leach tests have been performed to assess the durability of the glass products. This paper discusses the results of the vitrification of high molybdenum feeds in the presence of reprocessing liquor in both the borosilicate glass formulation currently utilised on WVP and the modified formulation which contain calcium.

scholarly journals Corrosion tests in water of fuel elements irradiated in the world’s first NPP reactor

2019 ◽  
Vol 5 (4) ◽  
pp. 337-343
Author(s):  
Sergey N. Ivanov ◽  
Sergey I. Porollo ◽  
Yury D. Baranaev ◽  
Vladimir F. Timofeev ◽  
Yury V. Kharizomenov
Keyword(s):  
Dose Rate ◽  
Water Samples ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Fuel Composition ◽  
Spent Fuel ◽  
Artificial Defect ◽  
Corrosion Tests ◽  
Long Term Storage ◽  
Fuel Elements

Spent nuclear fuel (SNF) storage in reactor spent fuel pools (SFP) is one of the crucial stages of SNF management technology: it requires special measures to ensure nuclear and radiation safety. During long-term storage in water-filled SFPs, leak-tight canisters in which SFAs are usually placed can become unsealed, which will result in the development of corrosion processes in the fuel element (FE) claddings. We studied fragments of spent fuel elements of the AM reactor of the World’s First NPP during their long exposure in the aqueous medium. The aim of the study was to obtain experimental data on the corrosion changes in the FE claddings and fuel composition during storage as well as on the release of radioactive fission products from them. For the study, a laboratory facility for exposing fuel elements in the water was developed and experimental fragments of fuel elements were made. The study was carried out in the hot chamber of the SSC RF-IPPE. The change in the activity of the water was estimated by the γ-dose rate from the selected water sample. The diameter measurements and metallographic studies were carried out in various sections of FE fragments. Corrosion tests were carried out on fragments of spent fuel elements of the AM reactor of the World’s First NPP that were stored for a long time (more than 50 years – FEs with U-Mo fuel and ~ 20 years – FEs with UO2 fuel) using standard technology – first in SFP canisters filled with water and then in dry canisters in the air. Placing the fuel elements in the water did not lead to through damage to the FE claddings and a significant change in the size (diameter) of the outer cladding. Metallographic studies of the FE fragments after the corrosion tests showed the presence of intergranular and local frontal corrosion on the surface of the claddings, the depth of which exceeded the depth of the cladding corrosion defects before testing. The rate of radionuclide release from the fuel composition was estimated by the γ-dose rate of water samples taken from the glasses with FE fragments. Throughout the test period, the dose rate of water samples from the glasses with defect-free FEs remained constant. The dose rate from water samples taken from the glasses with the FE fragments with an artificial defect grew during storage.

scholarly journals Modern Methods of Radiochemical Reprocessing of Spent Nuclear Fuel

2019 ◽  
pp. 52-57
Author(s):  
T. Maltseva ◽  
А. Shyshuta ◽  
S. Lukashyn
Keyword(s):  
Nuclear Fuel ◽  
Large Scale ◽  
Spent Nuclear Fuel ◽  
Fuel Cycle ◽  
Fission Products ◽  
Purex Process ◽  
Minor Actinides ◽  
Closed Fuel Cycle ◽  
Highly Active ◽  
High Level

The paper is devoted to the history of development and the current state of technological and scientific advances in radiochemical reprocessing of spent nuclear fuel from water-cooled power reactors. Regarding spent nuclear fuel (SNF) of NPP power reactors, long-term energy security involves adopting a version of its radiochemical treatment, conditioning and recirculation. Recycling SNF is required for the implementation of a closed fuel cycle and the re-use of regeneration products as energy reactor fuels. The basis of modern technological schemes for the reprocessing of the spent nuclear fuel is the “Purex” process, developed since the 60s in the USA. The classic approach to the use of U and Pu nuclides contained in spent nuclear fuel is to separate them from fission products, re-enrich regenerated uranium and use plutonium for the production of mixed-oxide (MOX) fuel with depleted uranium. The modern reprocessing plants are able to deal with fuel with further increase of its main characteristics without significant changes in the initial project. In order to close the fuel cycle, it is needed to add the following technological steps: (1) removal of high-level and long-lived components and minor actinides; (2) return of actinides to the technological cycle; (3) safe disposal of unused components. Each of these areas is under investigation now. Several new promising multi-cycle hydrometallurgical processes based on the joint extraction of trivalent lanthanides and minor actinides with their subsequent separation have been developed. A number of promising materials is suggested to be potential matrices for the immobilization of high-level components of radioactive wastes. To improve the compatibility of fuel processing with the environment, non-aqueous technologies are being developed, for instance, pyro-chemical methods for the reprocessing of various types of highly active fuels based on metals, oxides, carbides, or nitrides. An important scientific and technological task under investigation is transmutation of actinides. The results of international large-scale experiments on the partitioning and transmutation of fuel with various minor actinides and long-lived fission products confirm the real possibility and expediency of closing the nuclear fuel cycle.

scholarly journals Aluminosilicate Sediment Dissolution in Liquid Radioactive Waste Storage Facilities

2020 ◽  
Vol 11 (2) ◽  
pp. 75-84
Author(s):  
T. A. Kulagina ◽  
◽  
V. A. Kulagin ◽  
Keyword(s):  
Radioactive Waste ◽  
Spent Nuclear Fuel ◽  
Liquid Radioactive Waste ◽  
Storage Tanks ◽  
Spent Nuclear Fuel Reprocessing ◽  
Long Term Storage ◽  
Waste Storage ◽  
Poorly Soluble ◽  
Hydrodynamic Effects

The article deals with the extraction of insoluble sediments formed in storage tanks during long-term storage of liquid radioactive waste from spent nuclear fuel reprocessing. Results of a thermodynamic analysis are presented enabling to assess the structure of the precipitation formed and to select most effective modes for thermal and hydrodynamic effects (cavitation technology) produced by liquid medium on eroded sediments. The paper presents the results of studies on the extraction of poorly soluble pulp components from storage tanks using cavitation technology.

scholarly journals New Methods for Evaluation of Spent Fuel Condition during Long-Term Storage in Slovakia

2009 ◽  
Vol 2009 ◽  
pp. 1-5
Author(s):  
M. Mikloš ◽  
V. Kršjak
Keyword(s):  
Nuclear Fuel ◽  
Spent Nuclear Fuel ◽  
Detection System ◽  
Fission Products ◽  
Spent Fuel ◽  
Limit Values ◽  
Long Term Storage ◽  
Fuel Assemblies ◽  
Leak Tightness ◽  
Evaluation And Monitoring

Experiences with an advanced spent nuclear fuel management in Slovakia are presented in this paper. The evaluation and monitoring procedures are based on practices at the Slovak wet interim spent fuel storage facility in NPP Jaslovské Bohunice. Since 1999, leak testing of WWER-440 fuel assemblies are provided by special leak tightness detection system “Sipping in pool” delivered by Framatomeanp with external heating for the precise defects determination. In 2006, a new inspection stand “SVYP-440” for monitoring of spent nuclear fuel condition was inserted. This stand has the possibility to open WWER-440 fuel assemblies and examine fuel elements. Optimal ways of spent fuel disposal and monitoring of nuclear fuel condition were designed. With appropriate approach of conservativeness, new factor for specifying spent fuel leak tightness is introduced in the paper. By using computer simulations (based on SCALE 4.4a code) for fission products creation and measurements by system “Sipping in pool,” the limit values of leak tightness were established.

Materials Characteristics and Dissolution Behavior of Spent Nuclear Fuel

MRS Bulletin ◽  
1994 ◽  
Vol 19 (12) ◽  
pp. 24-27 ◽  
Author(s):  
L.H. Johnson ◽  
L.O. Werme
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Nuclear Power ◽  
Spent Nuclear Fuel ◽  
Fission Products ◽  
Waste Form ◽  
Dissolution Behavior ◽  
Spent Fuel ◽  
Power Stations ◽  
Geologic Disposal

The geologic disposal of spent nuclear fuel is currently under consideration in many countries. Most of this fuel is in the form of assemblies of zirconium-alloy-clad rods containing enriched (1–4% 235U) or natural (0.71% 235U) uranium oxide pellets. Approximately 135,000 Mg are presently in temporary storage facilities throughout the world in nations with commercial nuclear power stations.Safe geologic disposal of nuclear waste could be achieved using a combination of a natural barrier (the host rock of the repository) and engineered barriers, which would include a low-solubility waste form, long-lived containers, and clay- and cement-based barriers surrounding the waste containers and sealing the excavations.A requirement in evaluating the safety of disposal of nuclear waste is a knowledge of the kinetics and mechanism of dissolution of the waste form in groundwater and the solubility of the waste form constituents. In the case of spent nuclear fuel, this means developing an understanding of fuel microstructure, its impact on release of contained fission products, and the dissolution behavior of spent fuel and of UO2, the principal constituent of the fuel.

scholarly journals Management of the Energy and Economic Potential of Nuclear Waste Use

Energies ◽  
2021 ◽  
Vol 14 (12) ◽  
pp. 3709
Author(s):  
Bader Alshuraiaan ◽  
Sergey Pushkin ◽  
Anastasia Kurilova ◽  
Magdalena Mazur
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Spent Nuclear Fuel ◽  
Integrated Approach ◽  
Economic Potential ◽  
Nuclear Waste Storage ◽  
Scientific Discussion ◽  
Management Algorithm ◽  
Waste Storage ◽  
The Eu

Recently, issues related to the effects (benefit or harm) of processing nuclear waste and its further use as fuel have been increasingly often raised in the scientific discussion. In this regard, the research aims to investigate issues related to the assessment of the economic potential of nuclear waste use, as well as the cooperation between states in the context of the reduction of risks associated with nuclear waste storage and processing. The research methodology is based on an integrated approach, including statistical, factor analysis, and the proposed system of performance indicators for managing spent nuclear fuel use. The research was carried out on the basis of materials from Russia and the EU countries. In the course of the study, a model of cooperation between states has been developed (based on the example of technologies and methods of processing nuclear waste used in the EU and Russia) according to the nuclear waste (spent nuclear fuel) management algorithm. The model considers the risks and threats associated with ecology and safety. The developments and other results described in the study should be used in further research devoted to the use of nuclear waste as heat-producing elements.

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