| 1. |
- Costa, Diogo Ribeiro, et al.
(författare)
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Coated UN microspheres embedded in UO2 matrix as an innovative advanced technology fuel: Early progress
- 2021
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Ingår i: TopFuel 2021 Light Water Reactor Fuel Performance Conference, Santander, Spain, October 24-28, 2021..
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Konferensbidrag (refereegranskat)abstract
- Uranium nitride (UN)-uranium dioxide (UO2) composites have been proposed as an innovative advanced technology fuel (ATF) option for light water reactors (LWRs). However, the interdiffusion of oxygen and nitrogen during fabrication result in the formation of α-U2N3. A way to avoid this interaction is to coat the UN with a material that is impermeable to oxygen and nitrogen, has a high melting point, high thermal conductivity, and reasonable low neutron cross-section. Among many candidates,refractory metals may be the first option. In this study, we present an early progressresult of fabricating an innovative ATF concept: coated UN microspheres embedded in UO2 matrix. To do so, the following steps are performed: 1) diffusion couple experiments of UN-X-UO2 (X=W, Mo, Ta, Nb, V) to evaluate the interactions between the coating candidates (X) and the fuels; 2) selection of the most promising candidates; 3) use a surrogate material (ZrN microspheres) to develop processes to coat the microspheres with nanopowders: dry and wet methods; 4) coating the UN microspheres with a selected method; 5) finally, sinter a coated UN-UO2 composite using spark plasma sintering (SPS), and compare the results with an uncoated UNUO2 composite sintered at the same SPS conditions (1500 °C, 80 MPa, 3 min,vacuum). The diffusion couple results indicate W and Mo as the most promising candidates, with the wet method showing the smoothest surface. So, dense (~95 %TD) W/UN-UO2 and Mo/UN-UO2 were sintered and the preliminary results show that the tungsten coating was not efficient due to poor adhesion. Conversely, the Mo coating (~15 µm) was efficient against the α-U2N3 formation. Therefore, this early progress indicates the possibility of fabricating an innovative ATF concept using a low cost and potentially applicable coating method.
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| 2. |
- Costa, Diogo Ribeiro, et al.
(författare)
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Coated ZrN sphere-UO2 composites as surrogates for UN-UO2 accident tolerant fuels
- 2022
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 567, s. 153845-
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Tidskriftsartikel (refereegranskat)abstract
- Uranium nitride (UN) spheres embedded in uranium dioxide (UO2) matrix is considered an innovative accident tolerant fuel (ATF). However, the interaction between UN and UO2 restricts the applicability of such composite in light water reactors. A possibility to limit this interaction is to separate the two materials with a diffusion barrier that has a high melting point, high thermal conductivity, and reasonably low neutron cross-section. Recent density functional theory calculations and experimental results on interface interactions in UN-X-UO2 systems (X = V, Nb, Ta, Cr, Mo, W) concluded that Mo and W are promising coating candidates. In this work, we develop and study different methods of coating ZrN spheres, used as a surrogate material for UN spheres: first, using Mo or W nanopowders (wet and binder); and second, using chemical vapour deposition (CVD) of W. ZrN-UO2 composites containing 15 wt% of coated ZrN spheres were consolidated by spark plasma sintering (1773 K, 80 MPa) and characterised by SEM/FIB-EDS and EBSD. The results show dense Mo and W layers without interaction with UO2. Wet and binder Mo methods provided coating layers of about 20 µm and 65 µm, respectively, while the binder and CVD of W methods layers of about 12 µm and 3 µm, respectively.
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| 3. |
- Costa, Diogo Ribeiro
(författare)
-
Encapsulated additive nuclear fuels as an innovative accident tolerant fuel concept : fabrication, characterisation and oxidation resistance
- 2023
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Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820.
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Tidskriftsartikel (refereegranskat)abstract
- UN-UO2 composites are considered an accident tolerant fuel (ATF) option for light water reactors (LWRs). However, the interactions between UN and UO2 and the low oxidation resistance of UN limit the application of such ATF composite concept in LWRs. A potential alternative to overcome these issues is encapsulating the UN fuel before sintering. Based on our recent studies, molybdenum and tungsten are selected to encapsulate UN spheres. In this article, different coating techniques, such as powder coating, chemical vapour deposition (CVD), and physical vapour deposition (PVD), were developed and applied to encapsulate surrogates and UN spheres. Encapsulated UN-UO2 pellets fabricated by the spark plasma sintering (SPS) method (1773 K, 80 MPa) were characterised by complementary techniques and evaluated against their oxidation resistance in air up to 973 K. The results show inert, dense, and non-uniform Mo and W layers of about 28 μm and 32 μm, respectively, obtained by the powder coating method. PVD provided uniform and dense layers of Mo and W of approximately 1.0 μm and 4.0 μm, respectively, but with cracks at the interface with the surrogate spheres. PVD-Mo onto UN spheres shows a dense and well-adhered layer of about 0.5 μm but with W contamination from the previous coating. The PVD-W and CVD-W results and the oxidation experiments will be in the final version of this manuscript.
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| 4. |
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| 5. |
- Costa, Diogo Ribeiro, et al.
(författare)
-
Oxidation of UN/U 2 N 3 -UO 2 composites: an evaluation of UO 2 as an oxidation barrier for the nitride phases
- 2021
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 544
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Tidskriftsartikel (refereegranskat)abstract
- Composite fuels such as UN-UO2 are being considered to address the lower oxidation resistance of the UN fuel from a safety perspective for use in light water reactors, whilst improving the in-reactor behaviour of the more ubiquitous UO2 fuel. An innovative UN-UO2 accident tolerant fuel has recently been fabricated and studied: UN microspheres embedded in UO2 matrix. In the present study, detailed oxidative thermogravimetric investigations (TGA/DSC) of high-density UN/U2N3-UO2 composite fuels (91-97 %TD), as well as post oxidised microstructures obtained by SEM, are reported and analysed. Triplicate TGA measurements of each specimen were carried out at 5 K/min up to 973 K in a synthetic air atmosphere to assess their oxidation kinetics. The mass variation due to the oxidation reactions (%), the oxidation onset temperatures (OOTs), and the maximum reaction temperatures (MRTs) are also presented and discussed. The results show that all composites have similar post oxidised microstructures with mostly intergranular cracking and spalling. The oxidation resistance of the pellet with initially 10 wt% of UN microspheres is surprisingly better than the UO2 reference. Moreover, there is no significant difference in the OOT (~557 K) and MRT (~615 K) when 30 wt% or 50 wt% of embedded UN microspheres are used. Therefore, the findings in this article demonstrate that the UO2 matrix acts as a barrier to improve the oxidation resistance of the nitride phases at the beginning of life conditions.
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| 6. |
- Costa, Diogo Ribeiro, et al.
(författare)
-
UN microspheres embedded in UO2 matrix: An innovative accident tolerant fuel
- 2020
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 540
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Tidskriftsartikel (refereegranskat)abstract
- Uranium nitride (UN)-uranium dioxide (UO2) composite fuels are being considered as an accident tolerant fuel (ATF) option for light water reactors. However, the complexity related to the chemical interactions between UN and UO(2 )during sintering is still an open problem. Moreover, there is a lack of knowledge regarding the influence of the sintering parameters on the amount and morphology of the alpha-U2N3 phase formed. In this study, a detailed investigation of the interaction between UN and UO2 is provided and a formation mechanism for the resulting alpha-U2N3 phase is proposed. Coupled with these analyses, an innovative ATF concept was investigated: UN microspheres and UO2,13 powder were mixed and subsequently sintered by spark plasma sintering. Different temperatures, pressures, times and cooling rates were evaluated. The pellets were characterised by complementary techniques, including XRD, DSC, and SEM-EDS/WDS/EBSD. The UN and UO2 interaction is driven by O diffusion into the UN phase and N diffusion in the opposite direction, forming a long-range solid solution in the UO2 matrix, that can be described as UO2-xNx. The cooling process decreases the N solubility in UO2-xNx, causing then N redistribution and precipitation as alpha-U2N3 phase along and inside the UO2 grains. This precipitation mechanism occurs at temperatures between 1273 K and 973 K on cooling, following specific crystallographic grain orientation patterns. (C) 2020 The Authors. Published by Elsevier B.V.
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| 7. |
- Dehlin, Fredrik, 1994-, et al.
(författare)
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Activation analysis of the lead coolant in SUNRISE-LFR
- 2023
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Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 414
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Tidskriftsartikel (refereegranskat)abstract
- A lumped, zero-dimensional, mass transport model is combined with a depletion matrix solver to study the influence of coolant circulation on radionuclide build-up in a small lead-cooled fast reactor. It is shown that the addition of coolant circulation results in a lower activity for a minority of studied nuclides, and it is thus recommended to consider stagnant coolant when licensing a reactor. Activation analysis of three different lead qualities potentially used in SUNRISE-LFR is performed, and the result shows that a low silver content is desirable to simplify maintenance and decommissioning.
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| 8. |
- Dehlin, Fredrik, 1994-, et al.
(författare)
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An analytic approach to the design of passively safe lead-cooled reactors
- 2022
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 169, s. 108971-108971
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Tidskriftsartikel (refereegranskat)abstract
- A methodology to assist the design of liquid metal reactors, passively cooled by natural circulation duringoff-normal conditions, is derived from first principle physics. Based on this methodology, a preliminarydesign of a small LFR is accomplished and presented with accompanying neutronic and reactor dynamiccharacterizations. The benefit of using this methodology for reactor design compared to other availablemethods is discussed.
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| 9. |
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| 10. |
- Ekberg, Christian, 1967, et al.
(författare)
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Fuel fabrication and reprocessing issues: the ASGARD project
- 2020
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Ingår i: EPJ NUCLEAR SCIENCES & TECHNOLOGIES. - : EDP Sciences. - 2491-9292. ; 6
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Forskningsöversikt (refereegranskat)abstract
- The ASGARD project (2012-2016) was designed to tackle the challenge the multi-dimensional questions dealing with the recyclability of novel nuclear fuels. These dimensions are: the scientific achievements, investigating how to increase the industrial applicability of the fabrication of these novel fuels, the bridging of the often separate physics and chemical communities in connection with nuclear fuel cycles and finally to create an ambitious education and training platform. This will be offered to younger scientists and will include a broadening of their experience by international exchange with relevant facilities. At the end of the project 27 papers in peer reviewed journals were published and it is expected that the real number will be the double. The training and integration success was evidenced by the fruitful implementation of the Travel Fund as well as the unique schools, e.g. practical and theoretical handling of plutonium.
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| 11. |
- Hernandez, Cuauhtemoc Reale, et al.
(författare)
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Development of a CFD-based model to simulate loss of flow transients in a small lead-cooled reactor
- 2022
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 392, s. 111773-
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Tidskriftsartikel (refereegranskat)abstract
- With the deployment of advanced and small modular reactors (SMRs), it is important to develop the tools to assess their safety. This work presents the different components of a CFD based model for simulating transients in a pool-type small lead cooled reactor. The model encompasses the entire primary circuit with a simplification of the fuel channels, pumps and steam generators. Those parts are modelled through heat and momentum sources (or sinks), similar to the porous medium used in other studies. The CFD solver is coupled with a finite volume solver for fuel pin temperature and a point kinetics solver for neutronics. Free surface is modelled in CFD with multiphase volume of fluid method. The set of methods that is used in this work constitute a novelty for modelling lead cooled reactors. The goal is to have a model that is relatively simple to implement in order to study the effect of some parameters on reactor transients like an unprotected loss of flow. The focus of this study is to describe in detail every individual component of the model, namely the fuel channels, fuel pin temperature, neutronics, coupling strategy, pump and steam generators. In addition, CFD simulations are compared against experimental data from the TALL-3D facility. The purpose of this comparison is to verify that the models and parameters of the CFD software (STAR-CCM+) are capable of reproducing a flow of heavy metal. A future publication will provide the simulation results of an integrated model with all the components.
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| 12. |
- Huang, Zi-Nan, et al.
(författare)
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Analysis of the stress field in the reactor vessel of the China Initiative Accelerator Driven System during postulated ULOF and UTOP transients
- 2023
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Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 194
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Tidskriftsartikel (refereegranskat)abstract
- The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The subcritical reactor in CiADS is a liquid Lead Bismuth Eutectic (LBE) cooled fast reactor. When the reactor core is in operation, the LBE coolant will directly contact and corrode the inner surface of reactor vessel. Due to the high temperature, the corrosion will be more severe. If the stress on the reactor vessel exceeds the limit, the plastic deformation will occur, leading to the generation and expansion of defects and cracks, and the safety of the reactor will be affected. Therefore, evaluating the stress field of the reactor vessel under different operating conditions is a very important research project. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS, and the ASME Code was used as stress assessment standards. We can preliminarily prove that the stress assessments of the vessel during the postulated Unprotected Loss of Flow (ULOF) accidents satisfy the requirements of ASME Code. The limit reactivity insertion to protect the vessel from plastic deformation is 0.58$ in the postulated Unprotected Transient over Power (UTOP) accidents based on our current results. Therefore, we can preliminarily conclude that the current material selection and structural design of the reactor vessel in CiADS could survive most of the postulated transient accidents considering the stress effect.
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| 13. |
- Meng, Lu, et al.
(författare)
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Study on the safety performance of an offshore stationary lead-cooled fast reactor design loaded with nitride fuel
- 2024
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Ingår i: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100. ; 208
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Tidskriftsartikel (refereegranskat)abstract
- As a new generation of reactor type, lead-cooled fast reactors, have better safety behaviors, higher reliability, and better economic performance, aiming at island power supply through nuclear energy, seawater desalination, optimization of nuclear submarines, etc. Its evolution of nuclear waste and the advantages of nuclear non-proliferation provide a good prospect for development. This paper studied the safety performances of an offshore stationary lead-cooled reactor (OSLR) proposed in the National Key Research and Development Program of China. The transient analysis code SAS4A/SASSYS-1 was used to perform simulations of unprotected over-power accidents (UTOP) and unprotected loss of heat sink (ULOHS) accidents. The results indicated that offshore stationary lead-cooled reactors can withstand a maximum positive reactivity insertion of 0.5$ within 1 s during UTOP accidents without exceeding the working limits of the core. In ULOHS accidents, the inherent safety characteristics of OSLR allowed it to withstand 75 % heat removal capability of IHX. The simulation results were used to analyze the response of this stationary offshore reactor to transient accident conditions and the limits of its ability to withstand accidents in order to provide reference data for subsequent design and ideas for possible development of natural cycle lead-cooled reactors in the future.
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| 14. |
- Mishchenko, Yulia, et al.
(författare)
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Design and fabrication of UN composites : From first principles to pellet production
- 2021
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 553, s. 153047-
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Tidskriftsartikel (refereegranskat)abstract
- In this study the composite UN-AlN, UN-Cr, UN-CrN and UN-AlN-CrN pellets were fabricated, and the advanced microstructure with different modes of interaction between the phases was obtained. The dopants for this study were selected based on the results of the ab-initio modeling calculations, that identified the AlN phase as insoluble and CrN and Cr as soluble in the UN matrix. This method allowed to investigate the possibility of improving the corrosion resistance of UN by protecting the grain boundaries with insoluble AlN and by hindering the diffusion of oxygen through the bulk by adding soluble CrN and Cr. The UN powder was produced by hydriding-nitriding method and mixed with the AlN, CrN and Cr powders. High density (>90 %TD) composite pellets were sintered by Spark Plasma Sintering (SPS). The microstructure of the pellets was analysed using SEM coupled with EDS. The phase purity was determined by XRD. For the first time the presence of the ternary U2CrN3 phase was observed in the composite pellets containing Cr and CrN dopants. The results obtained in this study allowed to assess the methodology for fabrication of the UN composites with controlled microstructure.
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| 15. |
- Mishchenko, Yulia
(författare)
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Engineered microstructure composites as means of improving the oxidation resistance of uranium nitride
- 2023
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Doktorsavhandling (övrigt vetenskapligt/konstnärligt)abstract
- Owing to its high uranium density and good thermophysical properties,uranium nitride (UN) fuel has been considered as a potential Accident TolerantFuel (ATF) candidate for use in Light Water Reactors (LWRs). However,the main disadvantage of UN is its low oxidation resistance in water/steamcontaining atmospheres at the operating temperatures of LWRs.The main objective of this thesis is to investigate a concept of engineeredmicrostructure composites as means of improving the response of UN to watersidecorrosion. The methodology for incorporating the corrosion resistantadditives in the form of metals, nitrides and oxides into the UN matrix hasbeen developed and tested. The additives were proposed to produce coated(no interaction with UN) or doped (incorporation of the additive into theUN bulk) grains, which will be able to shield the UN from the oxidising environmentand slow down the oxygen diffusion through the bulk. The UNcomposite pellets containing the selected additives were sintered using theSpark Plasma Sintering (SPS) technique. The resulting microstructures ofthe composite pellets were well characterised prior to subjecting some of theengineered microstructure representative samples to oxidation testing in airand steam containing environments.The obtained results indicate that the response to air and steam oxidationof the composite samples differs from that of pure UN. Moreover, a delay inthe oxidation onset was observed for the composite samples UN-20CrNpremixand UN-20ZrNpremix in steam and for UN-20CrNpremix pellet in air. Theimproved response to oxidation was accompanied by the formation of theternary oxides, an observation that could be applied to the screening processof the additive candidates for waterproofing of UN.
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| 16. |
- Mishchenko, Yulia, et al.
(författare)
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Potential accident tolerant fuel candidate : Investigation of physical properties of the ternary phase U2CrN3
- 2022
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 568
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Tidskriftsartikel (refereegranskat)abstract
- In the present study, physical properties of the ternary phase U2CrN3 are evaluated experimentally and by modeling methods. High density pellets containing the ternary phase were prepared by spark plasma sintering (SPS). The microstructural and crystallographic analyses of the composite pellets were performed using scanning electron microscopy (SEM), standardised energy dispersive spectroscopy (EDS) and electron backscatter diffraction (EBSD). Evaluation of the mechanical properties was performed by nanoindentation test. The impact of temperature on lattice properties was evaluated using high temperature X-ray diffraction (XRD) coupled with modeling. Progressive change in the lattice parameters was obtained from room temperature (RT) to 673 K, and the result was used to calculate average linear thermal expansion coefficients, as well as an input for the density functional theory (DFT) modeling to reassess the degradation of the mechanical properties. The ab-initio calculation provides an initial assessment of electronic configuration of this ternary phase in a direct comparison with one of UN phase. For this goal, modeling was also employed to evaluate point defect formation energies and electronic charge distribution in the ternary phase. Results indicate that the U2CrN3 phase has similar mechanical properties to UN (Young's, bulk, shear moduli, hardness). No preferential crystallographic orientation was observed in the composite pellet. However, charge electron density distribution highlights the significant directionality of chemical bonds, which is in agreement with the anisotropy and non-linear behaviour of the obtained thermal expansion (α¯(aa) = 9.12 × 10−6/K, α¯(ab) = 5.81 × 10−6/K and α¯(ac) = 6.08 × 10−6/K). As a consequence, uranium was found to be more strongly bound in the ternary structure which may delay diffusion and vacancy formation, promising an acceptable performance as nuclear fuel.
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| 17. |
- Mishchenko, Yulia, et al.
(författare)
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Thermophysical properties and oxidation behaviour of the U0.8Zr0.2N solid solution
- 2023
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Ingår i: Nuclear Materials and Energy. - : Elsevier Ltd. - 2352-1791. ; 35
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Tidskriftsartikel (refereegranskat)abstract
- Thermophysical properties and oxidation behaviour of the composite pellet UN–20 vol%ZrN were investigated experimentally and compared with the behaviour of the pure UN pellet. A compound of a single phase, a solid solution of the average composition U0.8Zr0.2N, was obtained by Spark Plasma Sintering (SPS) of the powders UN and ZrN. Crystallographic and microstructural characterisation of the composite was performed using Scanning Electron Microscopy (SEM), standardised Energy Dispersive Spectroscopy (EDS) and Electron Backscatter Diffraction (EBSD). Nano hardness and Young's modulus were also measured by the nanoindentation method. High-Temperature X-ray diffraction (XRD) was applied to obtain the lattice expansion as a function of temperature (room temperature to 673 K). Thermogravimetric Analysis (TGA) was applied to evaluate oxidation behaviour in air. Results demonstrate that the fabrication method results in a matrix of solid solution with homogeneous composition averaged to U0.8Zr0.2N. The mechanical properties of such solution are uniform, with variation only due to the crystallographic orientation of the grains of the solution phase, similar to pure UN. The obtained value for the average linear thermal expansion coefficient is α¯ = 7.94 × 10-6/K, which compares well to UN (α¯ = 7.95 × 10-6/K) for the same temperature range. The degradation behaviour of the composite pellet UN-20 vol%ZrN in air shows a lower oxidation onset temperature, compared to pure UN, with the final product of oxidation being mainly U3O8. Smaller crystallites in the product of corrosion of the composite pellet indicate that the mechanism of degradation of the solid solution phase U0.8Zr0.2N is accompanied by the formation of two distinct oxides and their interaction.
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| 18. |
- Mishchenko, Yulia, et al.
(författare)
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Uranium nitride advanced fuel : an evaluation of the oxidation resistance of coated and doped grains
- 2021
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 556
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Tidskriftsartikel (refereegranskat)abstract
- The oxidation behaviour of the composite UN-AlN, UN-Cr 2 N/CrN and UN-AlN-Cr 2 N/CrN pellets in air and anoxic steam under thermal transient conditions was investigated and compared with the pure UN pellet. The composite pellets were manufactured to contain the engineered microstructure of coated (the addition of matrix-insoluble AlN) and doped (the addition of matrix-soluble Cr 2 N/CrN) grains. The composite powders were produced by powder metallurgy and sintered into pellets using the SPS method. Sintered composite pellets were subjected to a thermal transient up to 1273 K in an STA-EGA (TGA-DSC-Gas-MS) system, followed by crystallographic characterization by XRD and morphological and elemental analysis by FEG-SEM. Improved oxidation behaviour in air compared to pure UN was demonstrated by the UN-Cr 2 N/CrN composite pellet. The formation of the ternary oxide UCrO 4 from the ternary (U 2 Cr)N 3 phase (doped grain) was observed, consistent with the delayed oxidation onset and slower reaction rates. In an anoxic steam environment UN-Cr 2 N/CrN exhibited a higher onset oxidation temperature relative to UN, although the reaction progressed faster than for UN sample. Composite UN-AlN pellet oxidised at a lower temperature in both air and steam, compared to pure UN, due to internal stresses in the fuel matrix. A mechanism for degradation of the composite materials is proposed and the influence of the individual phases on the oxidation behaviour of the composites is discussed.
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| 19. |
- Morelová, Ikoleta, et al.
(författare)
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IAEA'S Coordinated Research Projects on Thermal Hydraulics of Fast Reactors
- 2023
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Ingår i: Proceedings of the 30th International Conference on Nuclear Engineering "Nuclear, Thermal, and Renewables: United to Provide Carbon Neutral Power", ICONE 2023. - : American Society of Mechanical Engineers (ASME).
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Konferensbidrag (refereegranskat)abstract
- A Coordinated Research Project on “Benchmark Analysis of FFTF Loss of Flow Without Scram Test” was launched by the International Atomic Energy Agency (IAEA) in 2018. A series of passive safety tests were conducted from 1980-1992 at the Fast Flux Test Facility (FFTF), 400 MW(th) liquid sodium cooled nuclear test reactor owned by U.S. Department of Energy (DOE) to demonstrate the potential of FFTF to survive severe accident initiators with no core damage. Amongst these tests was a series of Loss of Flow Without Scram (LOFWOS) tests from power levels up to 50%, also commonly referred to as Unprotected Loss of Flow (ULOF) tests, which were studied in the IAEA CRP. The data were provided by the Argonne National Laboratory (ANL) and Pacific Northwest National Laboratory (PNNL). Another Research Coordinated Project on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop” was launched by the IAEA in 2022. Three tests were conducted in 2017 to study the thermal-hydraulic behavior of a test fuel assembly cooled by lead-bismuth eutectic alloy during transition from forced to natural convection at the NACIE-UP facility at Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Italy. This project is the first IAEA CRP that is dedicated to the thermal hydraulics of lead and lead bismuth eutectic (LBE) technology. The paper provides a general overview of the two CRPs within the framework of the IAEA activities on thermal hydraulics of fast reactors.
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| 20. |
- Reale Hernandez, C., et al.
(författare)
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Dynamic sensitivity and uncertainty analysis of a small lead cooled reactor
- 2020
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 144
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Tidskriftsartikel (refereegranskat)abstract
- A sensitivity and uncertainty analysis was performed on a small lead cooled reactor for two types of transients: an unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP). Transients were simulated with the code BELLA, which is a point-kinetics and lumped-parameter model. A Monte Carlo based method was used with 5000 simulations. Input parameters are reactor dimensions, neutronics properties, material properties and thermal hydraulic properties. Outputs are maximum temperatures (clad, coolant and fuel), mass flow disturbance, natural convection mass flow, maximum power and energy deposition. For ULOF, it was found that the most sensitive parameters were the gap between fuel and clad, the flow area in the core, the friction factors in core and steam generator and the pump coastdown time. A deeper analysis recommends increasing pump coastdown time to avoid mass flow disturbances during coastdown. For UTOP, the most sensitive parameters are the gap between fuel and clad, the reactivity feedback coefficients, and to a lesser extent, fuel conductivity and fuel heat capacity. In any case, the uncertainties never bring the reactor beyond safety limits.
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| 21. |
- Reale Hernandez, Cuauhtemoc, et al.
(författare)
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Simulation of a loss of flow transient of a small Lead-Cooled reactor using a CFD-Based model
- 2023
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 412
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Tidskriftsartikel (refereegranskat)abstract
- The recent development of small modular reactors needs to be followed by safety analysis using the newest available tools. This work focuses on one type of reactor, SEALER, which is a small lead cooled reactor intended for remote communities in Canada. Simulations of a loss of flow transients are performed using a CFD-based model that was specifically developed for this project. The CFD geometry includes the entire primary circuit with some simplifications. The fuel channel, steam generator and pumps use a simple geometry with momentum source and heat source/sink. Free surface level is modelled with the multiphase volume of fluid (VOF) method. The CFD part of the model is coupled to a custom code for heat transfer in the fuel rods and point kinetics for neutronics. Transient results show that core temperatures do not increase significantly and stay well below coolant boiling and fuel melting points. The CFD-based model presented here is compared against a lumped-parameter model using the same transient. It is shown that the evolution of the mass flow and temperature is significantly different and more detailed with the CFD-based model. Finally, the influence of the moment of inertia of the pump flywheel on the transient is explored.
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| 22. |
- Wallenius, Janne, 1968-
(författare)
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An improved correlation for gas release from nitride fuels
- 2022
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 558
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Tidskriftsartikel (refereegranskat)abstract
- An improved correlation for gas release from nitride fuels is elaborated. Introducing empirical activation energies for migration of fission gases in presence of solid fission products and oxide impurities, it be-comes possible to better reproduce existing experimental data sets for gas release in sodium and helium bonded rods. The suggested approach may assist in resolving the previously poorly understood dispersion in measured gas release for identical irradiation conditions.
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| 23. |
- Wallenius, Janne, 1968-
(författare)
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Anomalous reactivity swing in the 238U-233U system
- 2020
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Ingår i: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 139
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Tidskriftsartikel (refereegranskat)abstract
- An anomalous negative reactivity swing is discovered in the iso-breeding 238U-233U system, resulting from an increase in conversion ratio with burn-up. Based on this discovery, a plutonium-free ternary fuel composition is identified allowing to reach a burn-up of 100 GWd/ton with a reactivity swing of ≃0.3$. Potential applications of such fuels are discussed.
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| 24. |
- Wallenius, Janne, 1968-
(författare)
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Micro-reactors
- 2021
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Ingår i: Encyclopedia of Nuclear Energy. - : Elsevier BV. ; , s. 749-756
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Bokkapitel (övrigt vetenskapligt/konstnärligt)
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| 25. |
- Wallenius, Janne, 1968-
(författare)
-
Nitride Fuels
- 2020
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Ingår i: Comprehensive Nuclear Materials: Second Edition. - : Elsevier BV. ; , s. 88-101
-
Bokkapitel (övrigt vetenskapligt/konstnärligt)
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| 26. |
- Wang, Di-Si, et al.
(författare)
-
Analysis of the Accelerator-Driven System Fuel Assembly during the Steam Generator Tube Rupture Accident
- 2021
-
Ingår i: Materials. - : MDPI AG. - 1996-1944. ; 14:8
-
Tidskriftsartikel (refereegranskat)abstract
- China is developing an ADS (Accelerator-Driven System) research device named the China initiative accelerator-driven system (CiADS). When performing a safety analysis of this new proposed design, the core behavior during the steam generator tube rupture (SGTR) accident has to be investigated. The purpose of our research in this paper is to investigate the impact from different heating conditions and inlet steam contents on steam bubble and coolant temperature distributions in ADS fuel assemblies during a postulated SGTR accident by performing necessary computational fluid dynamics (CFD) simulations. In this research, the open source CFD calculation software OpenFOAM, together with the two-phase VOF (Volume of Fluid) model were used to simulate the steam bubble behavior in heavy liquid metal flow. The model was validated with experimental results published in the open literature. Based on our simulation results, it can be noticed that steam bubbles will accumulate at the periphery region of fuel assemblies, and the maximum temperature in fuel assembly will not overwhelm its working limit during the postulated SGTR accident when the steam content at assembly inlet is less than 15%.
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| 27. |
- Wang, Di -Si, et al.
(författare)
-
Finite element analysis of the main reactor vessel in the China Initiative Accelerator Driven System
- 2023
-
Ingår i: Engineering Failure Analysis. - : Elsevier BV. - 1350-6307 .- 1873-1961. ; 146
-
Tidskriftsartikel (refereegranskat)abstract
- The China Initiative Accelerator Driven System (CiADS) was proposed by China Academy of Science since 2015. The reactor in CiADS is a subcritical fast neutron reactor cooled by a liquid lead-bismuth eutectic. The reactor operates at high temperature and bears high thermal stress. In addition to the heavy weight of the whole reactor, the vessel will bear large effective stress. If the effective stress exceeds the limit of the material, defects and cracks may occur on the main reactor vessel, which will affect the safety performances of the reactor. Therefore, it is very important to analyze the effective stress field of the reactor vessel. In this paper, the finite element analysis software ADINA was applied to analyze the reactor vessel in CiADS. We can preliminarily prove that the maximum effective stress that the vessel will bear during the postulated Unprotected Loss of Flow (ULOF) and Unprotected Transient over Power (UTOP) accidents is less than the yield strength of 316L stainless steel. Therefore, we can preliminarily conclude that the current ma-terial selection and structural design of the CiADS vessel could survive the postulated transient accidents considering the effective stress effect.
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| 28. |
- Wang, Guan, et al.
(författare)
-
Cladding fatigue analysis under frequent beam trips for China initiative Accelerator Driven System
- 2024
-
Ingår i: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 426
-
Tidskriftsartikel (refereegranskat)abstract
- China initiative Accelerator Driven System (CiADS) is a 10 MW lead–bismuth eutectic (LBE) cooled subcritical reactor, which is designed to demonstrate the engineering feasibility of the ADS concept. An elaborated methodology was developed to deal with the nonlinear mechanical behaviors of fuels under transients. Based on this methodology, the fuel mechanical module was accomplished and coupled with the neutron dynamics module and thermo-hydraulics module in the extended BELLA code. The cladding mechanical simulation under frequent beam trips for CiADS was carried out using BELLA, and fatigue analysis based on the experimental data was discussed. Combined with the CiADS beam-trip transients, it is concluded that the tolerance of the CiADS cladding to beam trips is between 1000 and 11,000 cycles.
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| 29. |
- Wang, Guan, et al.
(författare)
-
Transient analyses for China initiative Accelerator Driven System using the extended BELLA code
- 2023
-
Ingår i: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 190
-
Tidskriftsartikel (refereegranskat)abstract
- China initiative Accelerator Driven System (CiADS) is a 10 MW lead-bismuth eutectic (LBE) cooled subcritical reactor loaded with UO2 fuels in the first phase, which is designed to demonstrate the engineering feasibility of the ADS concept. The transient analyses were performed to show the safety potential and dynamic characteristics of the CiADS subcritical reactor using the extended BELLA code. Typical scenarios, such as beam-trip transients, unprotected beam overpower (UBOP), unprotected and protected loss of flow (ULOF and PLOF), unprotected and protected loss of heat sink (ULOHS and PLOHS) and self-defined station blackout (SBO), were analyzed and simulated. The results indicate that, as a low-power reactor, the CiADS subcritical reactor has a large margin to avoid severe damage in the core. Short-term cladding failure caused by thermal fatigue under frequent beam trips may not happen since the variations of temperatures are relatively small. However, accelerated LBE corrosion combined with accumulated creep tends to be a risk under ULOF and ULOHS.
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| 30. |
- Xi, Bin, et al.
(författare)
-
Influence of TIG and Laser Welding Processes of Fe-10Cr-4Al-RE Alloy Cracks Overlayed on 316L Steel Plate
- 2022
-
Ingår i: Materials. - : MDPI AG. - 1996-1944. ; 15:10, s. 3541-
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, the possibility of applying different welding strategies to overlay an FeCrAl layer against corrosion from heavy liquid metal on a plain plate made of 316L austenitic stainless steel was investigated. This technology could be used in manufacturing the main vessel of CiADS, which may be considered as a more economic and feasible solution than production with the corrosion-resistant FeCrAl alloy directly. The main operational parameters of the laser welding process, including laser power, weld wire feeding speed, diameter of the welding wire, etc., were adjusted correspondingly to the optimized mechanical properties of the welded plate. After performing the standard nuclear-grade bending tests, it can be preliminarily confirmed that the low-power pulse laser with specific operational parameters and an enhanced cooling strategy will be suitable to surface an Fe-10Cr-4Al-RE layer with a thickness of approximately 1 mm on a 40 mm-thick 316L stainless steel plate, thanks to the upgraded mechanical properties incurred by refined grains with a maximum size of around 300 mu m in the welded layer.
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| 31. |
- Yang, Sheng, et al.
(författare)
-
Development of a welding process to overlay FeCrAl alloy on a thin wall austenitic stainless steel tube
- 2021
-
Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 27
-
Tidskriftsartikel (refereegranskat)abstract
- In this paper, we investigated the possibility of applying the low power pulse laser welding technology to surface a protective layer against heavy liquid metal corrosion in fuel cladding tube made of austenitic 316L stainless steel. Based on results from flaring, flattening, bending tests and metallographic microscope investigations, we can preliminarily confirm the possibility of using low power pulse laser with specific power input to weld Fe-10Cr-4Al-RE alloy on the outer surface of a 316L stainless steel tube with the inner diameter of 11.8 mm and wall thickness of 0.65 mm.
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