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At least 1,027 records · Page 57
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Coated particle fuel: Historical perspectives and current progress

We report that coated particle fuel concepts date back some 60 years, and have evolved significantly from the relatively primitive pyrocarbon-coated kernels envisioned by the first pioneers. Improvements in particle design, coating layer properties, and kernel composition have produced the modern tristructural isotropic (TRISO) particle, capable of low statistical coating failure fractions and good fission product retention under extremely severe conditions, including temperatures of 1600 °C for hundreds of hours. The fuel constitutes one of the key enabling technologies for high-temperature gas-cooled reactors, allowing coolant outlet temperatures approaching 1000 °C and contributing to enhanced reactor safety due to the hardiness of the particles. TRISO fuel development has taken place in a number of countries worldwide, and several fuel qualification programs are currently in progress. Here, we discuss the unique history of particle fuel development and some key technology advances, concluding with some of the latest progress in UO 2 and UCO TRISO fuel qualification.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

UMo Solutions Processing in La Hague Cold Crucible Induction Melter: the Feedback From the First Years of Operation - 16376

250 cubic meters of high-level liquid waste from reprocessed U-Mo-Sn-Al spent fuel, used in Gas Cooled Reactors (GCR), were produced during the mid-1960's at La Hague facility. These 'UMo' solutions are less radioactive than the current fission product concentrates coming from ongoing reprocessing activities, but are very rich in molybdenum and phosphorus whose contents make the molten glass quite corrosive and require a high-temperature glass formulation to obtain sufficiently high waste loading factors (12% in molybdenum oxide). Hence the use of the Cold Crucible Induction Melter (CCIM) technology to process such solutions has been deemed a good opportunity for AREVA to meet its performance expectations. In addition to being very corrosive, the UMo waste is quite challenging to process as the molybdenum has a strong tendency to stick in the calciner and causes clogging issues in off-gas treatment equipment. Therefore, the process and technological qualifications were deployed in order to address these specific issues. UMo solutions processing in the La Hague CCIM has started in January 2013 and is currently ongoing. During this period (from 2013 to 2015), many data have been collected to confirm the process parameters that were defined during the qualification of this innovating process. Even if some difficulties occurred, operations teams experience along with engineering and R and D support allowed managing them. This paper presents the start-up methodology and the feedback from the first years of UMo solutions processing with the CCIM technology at La Hague site. Lessons learned are presented with the difficulties encountered and the solutions implemented, emphasizing the benefits of a close integration between R and D, engineering and operations teams. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1994

Numerical Analysis of an Impinging Jet Reactor for the CVD and Gas-Phase Nucleation of Titania

We model a cold-wall atmospheric pressure impinging jet reactor to study the CVD and gas-phase nucleation of TiO2 from a titanium tetra-iso-propoxide (TTIP)/oxygen dilute source gas mixture in nitrogen. The mathematical model uses the computational code FIDAP and complements our recent asymptotic theory for high activation energy gas-phase reactions in thin chemically reacting sublayers. The numerical predictions highlight deviations from ideality in various regions inside the experimental reactor. Model predictions of deposition rates and the onset of gas-phase nucleation compare favorably with experiments. Although variable property effects on deposition rates are not significant (approximately 11 percent at 1000 K), the reduction rates due to Soret transport is substantial (approximately 75 percent at 1000 K).

Gokoglu, Suleyman A.↗
DOE OSTI2017

Use of spectral shift to greatly improve uranium utilisation in FHR designs

Fluoride-salt-cooled high-temperature reactors (FHRs) are thermal-spectrum nuclear reactors defined by their use of liquid-fluoride-salt coolants, together with tri-isotropic (TRISO)-coated particle fuels and graphite moderator materials (Forsberg, 2004). FHRs are also characterized by their improved thermodynamic efficiency relative to gas-cooled reactors. The liquid salt coolant is designed to operate with high core outlet temperature and nearly atmosphere pressure. These features allow the FHR to possess higher core power density compared to gas-cooled TRISO-based systems. However, industry fabrication limitations result in a low volumetric fraction of fuel kernel in TRISO particles. In order to obtain reasonable operation periods, the enrichment needs to be in the high LEU (low enriched uranium) range as was proposed in the original design (Greene, 2010). Recent studies (Ilas, 2014) presented newly repurposed SmAHTR-CTC reactor concept that relies on prior experience obtained from the 2010 initial SmAHTR design (Greene, 2010). This design consists of 19 hexagonal fuel assemblies arranged in a 3-ring hexagonal geometry with 18 fuel plates each. The active length of fuel plates in each fuel assembly is 4 meters. The proposed SmAHTR-CTC achieves substantial improvements from the neutronic performance standpoint over the original concept. The results indicate that a fuel cycle length of over two years with discharged fuel burnup of 40 gigawatt-day per metric ton of initial heavy metal (HM) is possible with an 8% fuel enrichment in a once-through fuel cycle. However, the lower uranium enrichment loading requires a more frequent refueling strategy and hence considerably reduces the cycle length. It is necessary to achieve reasonable cycle length values but at the same time to extend the residence time of the fuel, especially since the costs associated with TRISO-fuel are substantially higher (Shropshire et al., 2008) than traditional UO{sub 2} fuel ($10,000/kg for TRISO vs $200/kg for UO{sub 2}). This paper presents an approach to extend the operational time and improve fuel utilization of FHR by means of spectral shift control. This is achieved by incorporating movable graphite structures into the design, instead of the static graphite blocks presently envisaged. At beginning of life (BOL), the graphite structures are withdrawn. The reactor is under-moderated, with excess neutrons being primarily captured in {sup 238}U, breeding {sup 239}Pu. Towards the end of life (EOL), the graphite structures are inserted, thermalizing the neutron spectrum and increasing reactivity. The extra {sup 239}Pu bred during the cycle is then burned, allowing the cycle and discharge burnup to be extended. Such movable structures regime should be feasible from an engineering point of view, given that it is broadly equivalent to inserting and withdrawing control rods. The preliminary results indicate that such a design may considerably improve the fuel utilization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

The DISTINCTIVE University Consortium: AGR, Magnox and Exotic Spent Fuels - 18345

The Engineering and Physical Sciences Research Council (EPSRC) sponsored consortium Decommissioning, Immobilisation and Storage Solutions for Nuclear Waste Inventories (DISTINCTIVE) is contributing to the development of skilled cross-disciplinary scientists and engineers with the knowledge and experience required to develop engineering solutions tailored for application within radiologically contaminated sites. Of the four themes addressed by the DISTINCTIVE consortium, the AGR (Advanced Gas-Cooled Reactor), Magnox and Exotic Spent Fuels theme considers the UK's diverse spent nuclear fuel inventory and the preferred options for disposal; this includes the retrieval of fuel from current storage facilities and the repackaging options available. Specifically, the research is concerned with AGR, Magnox and so-called 'Exotic' spent fuels; the goal being to increase knowledge and mechanistic understanding of the processes affecting long-term management and mitigation of any potential risks. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

The IAEA coordinated research programme on HTGR uncertainty analysis: Phase I status and Ex. I-1 prismatic reference results

The quantification of uncertainties in design and safety analysis of reactors is today not only broadly accepted, but in many cases became the preferred way to replace traditional conservative analysis for safety and licensing analysis. The use of a more fundamental methodology is also consistent with the reliable high fidelity physics models and robust, efficient, and accurate codes available today. To facilitate uncertainty analysis applications a comprehensive approach and methodology must be developed and applied, in contrast to the historical approach where sensitivity analysis were performed and uncertainties then determined by a simplified statistical combination of a few important input parameters. New methodologies are currently under development in the OECD/NEA Light Water Reactor (LWR) Uncertainty Analysis in Best-Estimate Modelling (UAM) benchmark activity. High Temperature Gas-cooled Reactor (HTGR) designs require specific treatment of the double heterogeneous fuel design and large graphite quantities at high temperatures. The IAEA has therefore launched a Coordinated Research Project (CRP) on HTGR Uncertainty Analysis in Modelling (UAM) in 2013 to study uncertainty propagation specifically in the HTGR analysis chain. Two benchmark problems are defined, with the prismatic design represented by the General Atomics (GA) MHTGR-350 and a 250 MW modular pebble bed design similar to the Chinese HTR-PM. Work has started on the first phase and the current CRP status is reported in the paper. A comparison of the Serpent and SCALE/KENO-VI reference Monte Carlo results for Ex. I-1 of the MHTGR-350 design is also included. It was observed that the SCALE/KENO-VI Continuous Energy (CE) k ∞ values were 395 pcm (Ex. I-1a) to 803 pcm (Ex. I-1b) higher than the respective Serpent lattice calculations, and that within the set of the SCALE results, the KENO-VI 238 Multi-Group (MG) k ∞ values were up to 800 pcm lower than the KENO-VI CE values. The use of the latest ENDF-B-VII.1 cross section library in Serpent lead to ~180 pcm lower k ∞ values compared to the older ENDF-B-VII.0 dataset, caused by the modified graphite neutron capture cross section. Furthermore, the fourth beta release of SCALE 6.2 likewise produced lower CE k∞ values when compared to SCALE 6.1, and the improved performance of the new 252-group library available in SCALE 6.2 is especially noteworthy. A SCALE/TSUNAMI uncertainty analysis of the Hot Full Power variant for Ex. I-1a furthermore concluded that the 238 U(n,γ) (capture) and 235 U(View the MathML source) cross-section covariance matrices contributed the most to the total k ∞ uncertainty of 0.58%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Calcium looping process for high purity hydrogen production integrated with capture of carbon dioxide, sulfur and halides

A process for producing hydrogen comprising the steps of: (i) gasifying a fuel into a raw synthesis gas comprising CO, hydrogen, steam, sulfur and halide contaminants in the form of H.sub.2S, COS, and HX, wherein X is a halide; (ii) passing the raw synthesis gas through a water gas shift reactor (WGSR) into which CaO and steam are injected, the CaO reacting with the shifted gas to remove CO.sub.2, sulfur and halides in a solid-phase calcium-containing product comprising CaCO.sub.3, CaS and CaX.sub.2; (iii) separating the solid-phase calcium-containing product from an enriched gaseous hydrogen product; and (iv) regenerating the CaO by calcining the solid-phase calcium-containing product at a condition selected from the group consisting of: in the presence of steam, in the presence of CO.sub.2, in the presence of synthesis gas, in the presence of H.sub.2 and O.sub.2, under partial vacuum, and combinations thereof.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2015

Calcium looping process for high purity hydrogen production integrated with capture of carbon dioxide, sulfur and halides

A process for producing hydrogen comprising the steps of: (i) gasifying a fuel into a raw synthesis gas comprising CO, hydrogen, steam, sulfur and halide contaminants in the form of H.sub.2S, COS, and HX, wherein X is a halide; (ii) passing the raw synthesis gas through a water gas shift reactor (WGSR) into which CaO and steam are injected, the CaO reacting with the shifted gas to remove CO.sub.2, sulfur and halides in a solid-phase calcium-containing product comprising CaCO.sub.3, CaS and CaX.sub.2; (iii) separating the solid-phase calcium-containing product from an enriched gaseous hydrogen product; and (iv) regenerating the CaO by calcining the solid-phase calcium-containing product at a condition selected from the group consisting of: in the presence of steam, in the presence of CO.sub.2, in the presence of synthesis gas, in the presence of H.sub.2 and O.sub.2, under partial vacuum, and combinations thereof.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Determination of oxidation rates and volatile oxidation products for HTGR graphite matrix material exposed to steam atmospheres

High-temperature gas-cooled reactors (HTGRs) in operation use tristructural isotropic (TRISO) particles embedded in graphite and carbonized resin matrix to form the fuel element. This graphite matrix material serves as a supportive structural element, heat transfer medium, and neutron moderator. In HTGR designs, fuel compacts are exposed to helium coolant, which facilitates high outlet temperatures (750°C 2 O, 800°C 2 , and H 2 , are quantified in varied oxidant atmospheres using a coupled thermogravimetric analyzer and mass spectrometer. Furthermore, oxidation rates reported here for varied steam (H 2 O [g]) atmospheres are predominantly linear and comparable with literature values in the range of tested temperatures (800–1200°C). Changes in dominant matrix oxidation products from primarily CO to a mixture of CO, CO 2 , and H 2 were observed at higher temperatures (≥1000°C) and steam atmospheres (≥5 kPa pH 2 O). Kinetic data indicates that there was no shift in oxidation regime with chemical oxidation occurring at all temperatures and H2O (g) atmospheres tested. These data provide insight into the oxidation behavior of graphite matrix material and will inform future testing conditions, notably mixed atmospheric conditions, of HTGR fuel elements.

36 MATERIALS SCIENCE↗
DOE OSTI2018

IAEA CRP on HTGR UAM: Propagation of Phase I cross section uncertainties to Phase II neutronics steady state using SCALE/SAMPLER and PHISICS/RELAP5-3D

The IAEA Coordinated Research Project (CRP) on High Temperature Gas Cooled Reactor (HTGR) Uncertainties in Modeling (UAM) has been actively supported by the HTGR Methods group at Idaho National Laboratory since 2013. The INL team is responsible for the development of the prismatic reactor design exercises, based on the 350 MW General Atomics Modular HTGR (MHTGR). The focus of this paper is the reporting of the INL results obtained for the propagation of the lattice cross section uncertainties to the stand-alone neutronics simulation of the full MHTGR-350 core. For the nominal (i.e. best-estimate) model, a series of fresh and depleted single block and super cell SCALE/KENO and SCALE/NEWT models were developed at INL to assess spectral effects in the MHTGR-350, and a combination of these lattice cells were used to construct 8-group cross section libraries for use in the INL-developed PHISICS/RELAP5-3D code suite. The block-level homogenized libraries were assigned to the 220 fuel zones utilized in the MHTGR-350 PHISICS core model, according to the fresh and mixed fresh/depleted core loading pattern. A set of 1,000 perturbed NEWT lattice and PHISICS/RELAP5-3D core calculations were performed using the stochastic sampling module SAMPLER in SCALE 6.2. The paper discusses the mean and standard deviation values obtained for the eigenvalue, power offset and peak axial power density uncertainty results for this sequence of 1,000 lattice-to-core MHTGR-350 simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Graphite Waste Tank Cleanup and Decontamination under the Marcoule UP1 D and D Program - 13166

The UP1 plant in Marcoule reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. During more than 40 years, the decladding operations produced thousands of tons of processed waste, mainly magnesium and graphite fragments. In the absence of a French repository for the graphite waste, the graphite sludge content of the storage pits had to be retrieved and transferred into a newer and safer pit. After an extensive R and D program, the equipment and process necessary for retrieval operations were designed, built and tested. The innovative process is mainly based on the use of two pumps (one to capture and the other one to transfer the sludge) working one after the other and a robotic arm mounted on a telescopic mast. A dedicated process was also set up for the removal of the biggest fragments. The retrieval of the most irradiating fragments was a challenge. Today, the first pit is totally empty and its stainless steel walls have been decontaminated using gels. In the second pit, the sludge retrieval and transfer operations have been almost completed. Most of the non-pumpable graphite fragments has been removed and transferred to a new storage pit. After more than 6 years of operations in sludge retrieval, a lot of experience was acquired from which important 'lessons learned' could be shared. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Cleanup and Dismantling of Highly Contaminated Ventilation Systems Using Robotic Tools - 13162

The UP1 plant reprocessed nearly 20,000 tons of used natural uranium gas cooled reactor fuel coming from the first generation of civil nuclear reactors in France. Following operating incidents in the eighties, the ventilation system of the continuous dissolution line facility was shut down and replaced. Two types of remote controlled tool carriers were developed to perform the decontamination and dismantling operations of the highly contaminated ventilation duct network. The first one, a dedicated small robot, was designed from scratch to retrieve a thick powder deposit within a duct. The robot, managed and confined by two dedicated glove boxes, was equipped for intervention inside the ventilation duct and used for carrying various cleanup and inspection tools. The second type, consisting of robotic tools developed on the base of an industrial platform, was used for the clean-up and dismantling of the ventilation duct system. Depending on the type of work to be performed, on the shape constraints of the rooms and any equipment to be dismantled, different kinds of robotic tools were developed and installed on a Brokk 40 carrier. After more than ten years of ventilation duct D and D operations at the UP1 plant, a lot of experience was acquired about remote operations. The three main important lessons learned in terms of remote controlled operation are: characterizing the initial conditions as much as reasonably possible, performing non-radioactive full scale testing and making it as simple and modular as possible. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Chemical compatibility issues associated with use of SiC/SiC in advanced reactor concepts

Silicon carbide/silicon carbide (SiC/SiC) composites are of interest for components that will experience high radiation fields in the High Temperature Gas Cooled Reactor (HTGR), the Very High Temperature Reactor (VHTR), the Sodium Fast Reactor (SFR), or the Fluoride-cooled High-temperature Reactor (FHR). In all of the reactor systems considered, reactions of SiC/SiC composites with the constituents of the coolant determine suitability of materials of construction. The material of interest is nuclear grade SiC/SiC composites, which consist of a SiC matrix [high-purity, chemical vapor deposition (CVD) SiC or liquid phase-sintered SiC that is crystalline beta-phase SiC containing small amounts of alumina-yttria impurity], a pyrolytic carbon interphase, and somewhat impure yet crystalline beta-phase SiC fibers. The interphase and fiber components may or may not be exposed, at least initially, to the reactor coolant. The chemical compatibility of SiC/SiC composites in the three reactor environments is highly dependent on thermodynamic stability with the pure coolant, and on reactions with impurities present in the environment including any ingress of oxygen and moisture. In general, there is a dearth of information on the performance of SiC in these environments. While there is little to no excess Si present in the new SiC/SiC composites, the reaction of Si with O 2 cannot be ignored, especially for the FHR, in which environment the product, SiO 2 , can be readily removed by the fluoride salt. In all systems, reaction of the carbon interphase layer with oxygen is possible especially under abnormal conditions such as loss of coolant (resulting in increased temperature), and air and/ or steam ingress. A global outline of an approach to resolving SiC/SiC chemical compatibility concerns with the environments of the three reactors is presented along with ideas to quickly determine the baseline compatibility performance of SiC/SiC.

FHR↗
DOE OSTI2020

Two decades of DOE investment lays the foundation for TRISO-fueled reactors

Tristructural isotropic (TRISO) coated particle fuel is a robust, microencapsulated fuel form developed originally for use in high-temperature gas-cooled reactors (HTGRs). The particles consist of a spherical fissile kernel surrounded by several layers of pyrocarbon and a silicon carbide (SiC) layer (Figure 1). The particles are formed into cylindrical or spherical fuel forms using a resinated graphite matrix material for insertion into an HTGR. The kernel and coating layers together act to retain fission products within the particle during normal reactor operation and during postulated accidents; TRISO particles can maintain structural integrity at extremely high temperatures, reaching as high as approximately 1,600°C in limiting HTGR accidents. This limits the fission product activity circulating in the helium coolant and the activity released to the environment during accidents. Acceptable performance of TRISO particles is therefore essential for reactor safety.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Experimental investigation on multiscale hydrodynamics in a novel gas–Liquid–Solid three phase jet-Loop reactor

Multiphase flow hydrodynamics in a novel gas–liquid–solid jet-loop reactor (JLR) were experimentally investigated at the macroscales and mesoscales. The chord length distribution was measured by an optical fiber probe and transformed for bubble size distribution through the maximum entropy method. The impacts of key operating conditions (superficial gas and liquid velocity, solid loading) on hydrodynamics at different axial and radial locations were comprehensively investigated. JLR was found to have good solid suspension ability owing to the internal circulation of bubbles and liquid flow. The gas holdup, axial liquid velocity, and bubble velocity increase with gas velocity, while liquid velocity has little influence on them. Compared with the gas–liquid JLRs, solids decrease the gas holdup and liquid circulation, reduces the bubble velocity and delays the flow development due to the enhanced interaction between bubbles and particles (Stokes number >1). This work also provides a benchmark data for computational fluid dynamics (CFD) model validation.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

Start-up of Silo 130 Waste Retrieval at La Hague: first Step Towards Reducing Legacy Inventory - 20020

On June 25, 2019, the first grapple of waste and debris was lifted from Silo 130, one of the most challenging legacy waste storage silos at La Hague. This was the achievement of more than ten years of studies and efforts focused on solving some very unique technical and safety challenges, and on designing an innovative, efficient solution. The waste is from de-cladding materials from past Gas-Cooled Reactor fuel reprocessing operations. It is an unsorted mixture of magnesium cladding pieces, graphite debris, and various other metallic waste types. These materials can be chemically reactive and therefore need specific dispositions for handling and storing. The building is more than 40 years old and does not meet modern standards. French safety authorities have required the silo to be emptied. After about three years of operation, the bulk of this waste will be removed and placed in safe storage, pending the selection of a final conditioning solution. Removal of bottom debris will then be possible. When finished, this operation will bring a conclusion to a significant legacy-related hazard on the site. The silo will then be available for final cleaning and de-activation. This paper will describe the technical, safety, and economical challenges, and the various decisions and optimizations that led to the definition of the process and technology to be implemented. Several innovative features were developed and implemented, in all sections of the project: project organization, progressive approach for the waste disposition strategy, light building design, customized and robust retrieval technology, processing technologies involving the most modern robotic and sorting techniques, storage method... Several of these first-of-a-kind innovative features have been developed and qualified specifically for this application, using the detailed qualification procedure specified for all La Hague projects. The facility was built in less than five years, overcoming various difficulties along the line. Commissioning is now finished with the combined efforts of the project, engineering, start-up, and operating teams assisted by dedicated suppliers. The facility is now ramping-up. The paper will also report on the commissioning and start-up of the facility and describe the first operational results. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Enhancements to a dissolver burn up credit safety case to reduce further GD use in the thermal oxide reprocessing plant (THORP), Sellafield - 14229

THORP reprocesses fuel from Britain's Advanced Gas-cooled Reactors (AGRs) and fuel from Light Water Reactors (LWRs). The THORP dissolvers use nitric acid poisoned with Gadolinium (Gd) to ensure criticality safety. The dissolver criticality safety case uses a burnup credit approach to determine a minimum required Gd concentration. The burn up credit approach followed originates from a time when fuel receipts for THORP were ongoing and hence models a deliberately conservative and artificial fuel type to cover all potential fuel receipts. External fuel receipts have now ceased and detailed information on post irradiation U and Pu isotopics is available for all fuel scheduled for reprocessing. Work has therefore been undertaken to determine whether the burn up credit safety case could be revised to be more representative of the fuel actually remaining for reprocessing and hence to reduce further the required Gd concentration for the dissolvers. This would ultimately reduce the quantity of high active liquid waste requiring treatment and storage. The great majority of fuel remaining for reprocessing in THORP is AGR fuel, for which the typical burnup is known to be much higher than was previously assumed in the safety case. Pu and U isotopics for irradiated AGR fuel were examined with the fuel inventory code FISPIN. Modelling with the code MONK then showed that the Gd requirement for irradiated AGR fuel of a given burnup was much lower than in the previous safety case due to the change in Pu:U ratio and Pu{sup 240} in Pu for AGR fuel compared with the previous conservative assumptions. The burnup of all remaining LWR fuel was found to be high and hence lower dissolver Gd concentrations could also be tolerated for reprocessing this fuel. The work has concluded that the THORP dissolver Gd concentration can be halved for all remaining LWR and AGR fuel campaigns with a consequent cost saving of ∼ UK pound 4 m in Gd and high active waste processing time. Implementation of this improvement is currently ongoing on the plant. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗