| 1. |
- Juslin, N., et al.
(author)
-
Simulation of threshold displacement energies in FeCr
- 2007
-
In: Nuclear Instruments and Methods in Physics Research Section B. - : Elsevier BV. - 0168-583X .- 1872-9584. ; 255:1, s. 75-77
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Journal article (peer-reviewed)abstract
- We have studied the role of chromium on threshold displacement energies in FeCr for the fusion reactor steel relevant concentration 10% Cr. We have used molecular dynamics simulations in order to determine whether the observed Cr-content dependence of macroscopic properties can be due to the defect production. We compare FeCr-alloys with pure iron and chromium, employing two different potential sets for the Fe-Cr system. We find that there are no significant differences between pure iron and FeCr with 10% Cr for the 100, 110 and 111 directions and the average threshold energy.
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| 2. |
- Sandberg, Nils, et al.
(author)
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Carbon impurity dissolution and migration in bcc Fe-Cr : First-principles calculations
- 2008
-
In: Physical Review B. Condensed Matter and Materials Physics. - 1098-0121 .- 1550-235X. ; 78:9
-
Journal article (peer-reviewed)abstract
- First-principles density-functional theory calculations for C solution enthalpies, H-sol, and diffusion activation enthalpies, H-diff, in body-centered-cubic Fe and Cr are presented. The results for C in Fe compare well with experiments, provided that the effect of magnetic disordering is accounted for. Likewise, in Cr, the calculated Hsol and Hdiff agree well with available experiments. In both materials, the deviation between calculated enthalpies and critically assessed experimental enthalpies are less than 0.05 eV. Further, first-principles calculations for the interaction energies between a solute (e.g., a Cr atom in bcc Fe) and an interstitial C atom are presented. The results are in conflict with those inferred from internal friction (IF) experiments in disordered Fe-Cr-C alloys. A simple model of C relaxation in disordered Fe-Cr is used to compare theoretical and experimental IF curves directly. The results suggest that a more extensive study of the energetic, thermodynamic, and kinetic aspects of C migration in Fe-Cr is needed.
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| 3. |
- Wallenius, Janne, 1968-, et al.
(author)
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Muonic atom deexcitation via formation of metastable molecular states in light of experimental verification
- 2001
-
In: Hyperfine Interactions. - 0304-3843 .- 1572-9540. ; 138:04-jan, s. 285-288
-
Journal article (peer-reviewed)abstract
- In a recent experiment performed at PSI, a peak in the time-of-flight distribution of pmu(1s) atoms could be identified with decay of ppmu* molecular ions situated below the 2s threshold, providing 900 eV of kinetic energy to the pmu atom. This finding may be interpreted in terms of the side path model which suggests that metastable muonic molecules may form with high probability in resonant collisions between muonic hydrogen in the 2s state and hydrogen molecules, e.g., pmu(2s)H-2-->[(ppmu*)(vJ)(pq) - pee(])v(K) --> [(ppmu*)(v'J')(p'q') - pe](+) + e(-). The Coulombic decay of the Auger stabilised ppmu* molecular ion then leads to the formation of highly energetic pmu(1s) atoms. In the present paper calculations of resonant formation rates in pure hydrogen are presented and compared to the quenching rate of pmu(2s) atoms measured at low hydrogen density.
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| 4. |
- Bortot, Sara, 1983-, et al.
(author)
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BELLA : a multi-point dynamics code for simulation of fast reactors
-
In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100.
-
Journal article (peer-reviewed)abstract
- In this paper, the multi-point dynamics code BELLA and its benchmarking with respect to SAS4A/SASSYS- 1 is described for a small fast reactor cooled with natural convection of lead (ELECTRA). It is shown that BELLA is capable of reproducing the magnitude of mass-flow, reactivity, power and temperature excursions during design extension conditions with an accuracy better than 10%. Hence, the BELLA code can be used for safety-informed design and stability analyses of fast reactor systems, permitting to isolate essential phenomena and trends of significance for their safety assessment.
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| 5. |
- Costa, Diogo Ribeiro, et al.
(author)
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Coated UN microspheres embedded in UO2 matrix as an innovative advanced technology fuel: Early progress
- 2021
-
In: TopFuel 2021 Light Water Reactor Fuel Performance Conference, Santander, Spain, October 24-28, 2021..
-
Conference paper (peer-reviewed)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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| 6. |
- Costa, Diogo Ribeiro, et al.
(author)
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Coated ZrN sphere-UO2 composites as surrogates for UN-UO2 accident tolerant fuels
- 2022
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 567, s. 153845-
-
Journal article (peer-reviewed)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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| 7. |
- Costa, Diogo Ribeiro
(author)
-
Encapsulated additive nuclear fuels as an innovative accident tolerant fuel concept : fabrication, characterisation and oxidation resistance
- 2023
-
In: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820.
-
Journal article (peer-reviewed)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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| 8. |
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| 9. |
- Costa, Diogo Ribeiro, et al.
(author)
-
Oxidation of UN/U 2 N 3 -UO 2 composites: an evaluation of UO 2 as an oxidation barrier for the nitride phases
- 2021
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 544
-
Journal article (peer-reviewed)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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| 10. |
- Costa, Diogo Ribeiro, et al.
(author)
-
UN microspheres embedded in UO2 matrix: An innovative accident tolerant fuel
- 2020
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 540
-
Journal article (peer-reviewed)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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| 11. |
- Dehlin, Fredrik, 1994-, et al.
(author)
-
Activation analysis of the lead coolant in SUNRISE-LFR
- 2023
-
In: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 414
-
Journal article (peer-reviewed)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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| 12. |
- Dehlin, Fredrik, 1994-, et al.
(author)
-
An analytic approach to the design of passively safe lead-cooled reactors
- 2022
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 169, s. 108971-108971
-
Journal article (peer-reviewed)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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| 13. |
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| 14. |
- Ekberg, Christian, 1967, et al.
(author)
-
Fuel fabrication and reprocessing issues: the ASGARD project
- 2020
-
In: EPJ NUCLEAR SCIENCES & TECHNOLOGIES. - : EDP Sciences. - 2491-9292. ; 6
-
Research review (peer-reviewed)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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| 15. |
- Eriksson, Marcus, et al.
(author)
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Safety Analysis of Na and Pb-Bi Coolants in Response to Beam Instabilities
- 2003
-
In: UTILISATION AND RELIABILITY OF HIGH POWER PROTON ACCELERATORS, WORKSHOP PROCEEDINGS. - 9264102116 ; , s. 227-236
-
Conference paper (peer-reviewed)abstract
- A comparative safety study has been performed on sodium vs. lead/bismuth as coolant for accelerator-driven systems. Transient studies are performed for a beam overpower event. We examine a fuel type of recent interest in the research on minor actinide burners, i.e. uranium-free oxide fuel. A strong positive void coefficient is calculated for both sodium and lead/bismuth. This is attributed to the high fraction of americium in the fuel. It is shown that the lead/bismuth-cooled reactor features twice the grace time with respect to fuel or cladding damage compared to the sodium-cooled reactor of comparable core size and power rating. This accounts to the difference in void reactivity contribution and to the low boiling point of sodium. For improved safety features the general objective is to reduce the coolant void reactivity effect. An important safety issue is the high void worth that could possibly drive the system to prompt criticality.
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| 16. |
- Henriksson, Krister O. E., et al.
(author)
-
Carbides in stainless steels : Results from ab initio investigations
- 2008
-
In: Applied Physics Letters. - : AIP Publishing. - 0003-6951 .- 1077-3118. ; 93:19
-
Journal article (peer-reviewed)abstract
- The useful properties of steels are due to a complicated microstructure containing iron and chromium carbides. Only some basic physical properties of these carbides are known with high precision, although the carbides may have a vital impact on the performance and longevity of the steel. To improve on this situation, we have performed extensive density-functional theory calculations of several carbides. The quantitative results are in perfect agreement with the relative empirical stability of the carbides. Also, in contradiction with experimental data, we find that Cr23C6 responsible for the hardness of stainless steels is not the most stable chromium-dominated carbide.
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| 17. |
- Hernandez, Cuauhtemoc Reale, et al.
(author)
-
Development of a CFD-based model to simulate loss of flow transients in a small lead-cooled reactor
- 2022
-
In: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 392, s. 111773-
-
Journal article (peer-reviewed)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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| 18. |
- Huang, Zi-Nan, et al.
(author)
-
Analysis of the stress field in the reactor vessel of the China Initiative Accelerator Driven System during postulated ULOF and UTOP transients
- 2023
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 194
-
Journal article (peer-reviewed)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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| 19. |
- Johnson, Kyle, et al.
(author)
-
Oxidation of accident tolerant fuel candidates
- 2017
-
In: Journal of Nuclear Science and Technology. - : Taylor & Francis. - 0022-3131 .- 1881-1248. ; 54:3, s. 280-286
-
Journal article (peer-reviewed)abstract
- In this study, the oxidation of various accident tolerant fuel candidates produced under different conditions have been evaluated and compared relative to the reference standard–UO2. The candidates considered in this study were UN, U3Si2, U3Si5, and a composite material composed of UN–U3Si2. With the spark plasma sintering (SPS) method, it was possible to fabricate samples of UN with varying porosity, as well as a high-density composite of UN–U3Si2 (10%). Using thermogravimetry in air, the oxidation behaviors of each material and the various microstructures of UN were assessed. These results reveal that it is possible to fabricate UN to very high densities using the SPS method, such that its resistance to oxidation can be improved compared to U3Si5 and UO2, and compete favorably with the principal ATF candidates, U3Si2, which shows a particularly violent reaction under the conditions of this study, and the UN–U3Si2 (10%) composite.
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| 20. |
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| 21. |
- Jolkkonen, Mikael, et al.
(author)
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Thermo-chemical modelling of uranium-free nitride fuels
- 2004
-
In: Journal of Nuclear Science and Technology. - : Informa UK Limited. - 0022-3131 .- 1881-1248. ; 41:4, s. 457-465
-
Journal article (peer-reviewed)abstract
- A production process for americium-bearing, uranium-free nitride fuels was modelled using the newly developed ALCHYMY thermochemical database. The results suggested that the practical difficulties with yield and purity are of a kinetic rather than a thermodynamical nature. We predict that the immediate product of the typical decarburisation step is not methane, but hydrogen cyanide. HCN may then undergo further reactions upon cooling, explaining the difficulty in observing any carbophoric molecules in the gaseous off stream. The thermal stability of nitride fuels in different environments was also estimated. We show that sintering of nitride compounds containing americium should be performed under nitrogen atmosphere in order to the avoid the excessive losses of americium reported from sintering under inert gas. Addition of nitrogen in small amounts to fuel pin filling gas also appears to significantly improve the in-pile stability of transuranium nitride fuels.
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| 22. |
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| 23. |
- Lindroth, E., et al.
(author)
-
Decay rates of excited muonic molecular ions
- 2003
-
In: Physical Review A. Atomic, Molecular, and Optical Physics. - 1050-2947 .- 1094-1622. ; 68:3
-
Journal article (peer-reviewed)abstract
- Muonic molecular ions in excited states have been predicted to form in collisions between excited muonic atoms and hydrogen molecules. We have calculated radiative and Coulombic decay rates for ppmu(*) and ddmu(*) molecular states located below the 2s threshold, using the complex rotation method. The x-ray spectrum from the radiative decay is shown to exhibit several maxima, corresponding to the vibrational motion of the decaying molecule. The branching ratio of the radiative decay mode was calculated to be less than 15% for ppmu(*), while a radiative yield of more than 80% is predicted for the decay of ddmu(*). Our results have a significant impact on the analysis of the muon catalyzed fusion cycle as well as on the interpretation of exotic hydrogen spectroscopy.
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| 24. |
- Malerba, L., et al.
(author)
-
Modelling of Radiation Damage in Fe-Cr Alloys
- 2008
-
In: EFFECTS OF RADIATION ON MATERIALS. - 9780803134218 ; , s. 159-176
-
Conference paper (peer-reviewed)abstract
- High-Cr ferritic/martensitic steels are being considered as structural materials for a large number of future nuclear applications, from fusion to accelerator-driven systems and GenIV reactors. Fe-Cr alloys can be used as model materials to investigate some of the mechanisms governing their microstructure evolution under irradiation and its correlation to changes in their macroscopic properties. Focusing on these alloys, we show an example of how the integration of computer simulation and theoretical models can provide keys for the interpretation of a host of relevant experimental observations. In particular we show that proper accounting for two basic features of these alloys, namely, the existence of a fairly strong attractive interaction between self-interstitials and Cr atoms and of a mixing enthalpy that changes sign from negative to positive around 8 to 10 % Cr, is a necessary and, to a certain extent, sufficient condition to rationalize and understand their behavior under irradiation. These features have been revealed by ab initio calculations, are supported by experimental evidence, and have been adequately transferred into advanced empirical interatomic potentials, which have been and are being used for the simulation of damage production, defect behavior, and phase transformation in these alloys. The results of the simulations have been and are being used to parameterize models capable of extending the description of radiation effects to scales beyond the reach of molecular dynamics. The present paper intends to highlight the most important achievements and results of this research activity.
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| 25. |
- Malerba, L., et al.
(author)
-
Molecular dynamics simulation of displacement cascades in Fe-Cr alloys
- 2004
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 329-33, s. 1156-1160
-
Journal article (peer-reviewed)abstract
- An embedded atom method (EAM) empirical potential recently fitted and validated for Fe-Cr systems is used to simulate displacement cascades up to 15 keV in Fe and Fe-10%Cr. The evolution of these cascades up to thermalisation of the primary damage state is followed and quantitatively analysed. Particular attention is devoted to assessing the effect of Cr atoms on the defect distribution versus pure Fe. Using the Wigner-Seitz cell criterion to identify point defects, first results show that the main effect of the presence of Cr in the system is the preferential formation of mixed Fe-Cr dumbbells and mixed interstitial clusters, with expected lower mobility than in pure Fe.
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| 26. |
- Mishchenko, Yulia, et al.
(author)
-
Design and fabrication of UN composites : From first principles to pellet production
- 2021
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 553, s. 153047-
-
Journal article (peer-reviewed)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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| 27. |
- Mishchenko, Yulia
(author)
-
Engineered microstructure composites as means of improving the oxidation resistance of uranium nitride
- 2023
-
Doctoral thesis (other academic/artistic)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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| 28. |
- Mishchenko, Yulia, et al.
(author)
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Potential accident tolerant fuel candidate : Investigation of physical properties of the ternary phase U2CrN3
- 2022
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 568
-
Journal article (peer-reviewed)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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| 29. |
- Mishchenko, Yulia, et al.
(author)
-
Thermophysical properties and oxidation behaviour of the U0.8Zr0.2N solid solution
- 2023
-
In: Nuclear Materials and Energy. - : Elsevier Ltd. - 2352-1791. ; 35
-
Journal article (peer-reviewed)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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| 30. |
- Mishchenko, Yulia, et al.
(author)
-
Uranium nitride advanced fuel : an evaluation of the oxidation resistance of coated and doped grains
- 2021
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 556
-
Journal article (peer-reviewed)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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| 31. |
- Morelová, Ikoleta, et al.
(author)
-
IAEA'S Coordinated Research Projects on Thermal Hydraulics of Fast Reactors
- 2023
-
In: 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).
-
Conference paper (peer-reviewed)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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| 32. |
- Nordlund, K., et al.
(author)
-
Molecular dynamics simulations of threshold displacement energies in Fe
- 2006
-
In: Nuclear Instruments and Methods in Physics Research Section B. - : Elsevier BV. - 0168-583X .- 1872-9584. ; 246:2, s. 322-332
-
Journal article (peer-reviewed)abstract
- We compare systematically the threshold displacement energy surface of 11 interatomic potentials in Fe. We discuss in detail different possible definitions of threshold displacement energies, and how they relate to different kinds of experimental threshold displacement energies. We compare the threshold results to experiments, and find that none of the 11 tested potentials agrees fully with experiments. However, all the potentials predict some qualitative features in the same way, most importantly that the threshold energy surface close to the 100 crystal direction is flat and that the largest threshold energies occur around very roughly the 123 crystal direction.
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| 33. |
- Olsson, Pär, et al.
(author)
-
Ab initio formation energies of Fe-Cr alloys
- 2003
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 321:1, s. 84-90
-
Journal article (peer-reviewed)abstract
- We have calculated ab initio lattice parameters, formation energies, bulk moduli and magnetic moments of Fe-Cr alloys. The results agree well with available experimental data. In addition to body centered cubic (bcc) alloys, which are representative of ferritic steels used in fast neutron reactors, face centered cubic (fcc) and hexagonal close packed (hcp) phases were considered in order to complete a theoretical database of thermodynamic properties. Calculations were done for the ferromagnetic phase, as well as for a phase with local moment disorder, simulating the magnetic structure at high temperatures. For the latter case, the formation energy of the alloy is strictly positive smooth function of chromium concentration, in agreement with experiments performed at high temperature. In the ferromagnetic case, a negative mixing enthalpy is found for chromium concentrations below 6 %. Our observation is consistent with the experimentally observed inversion of the ordering trend, as well as with formation of the chromium rich alpha phase at Cr-concentrations above 9%, occurring at T < 900 K.
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| 34. |
- Olsson, Pär, et al.
(author)
-
Ab initio study of Cr interactions with point defects in bcc Fe
- 2007
-
In: Physical Review B. Condensed Matter and Materials Physics. - 1098-0121 .- 1550-235X. ; 75:1, s. 014110-
-
Journal article (peer-reviewed)abstract
- The properties of Cr in alpha Fe have been investigated by ab initio calculations based on density functional theory. The intrinsic point defect formation energies were found to be larger in model bcc Cr as compared to those in ferromagnetic bcc Fe. The interactions of Cr with point defects (vacancy and self-interstitials) have been characterized. Single Cr atoms interact weakly with vacancies but significantly with self-interstitial atoms (SIA). Mixed interstitials of any interstitial symmetry are bound. Configurations where two Cr atoms are in nearest-neighbor position are generally unfavorable in bcc Fe except when they are a part of a < 111 > interstitial complex. Mixed < 110 > interstitials do not have as strong directional stability as pure Fe interstitials have. The effects on the results using the atom description scheme of either the ultrasoft pseudopotential (USPP) or the projector augmented wave (PAW) formalisms are connected to the differences in local magnetic moments that the two methods predict. As expected for the Fe-Cr system, the results obtained using the PAW method are more reliable than the ones obtained with USPP.
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| 35. |
- Olsson, Pär, et al.
(author)
-
Electronic origin of the anomalous stability of Fe-rich bcc Fe-Cr alloys
- 2006
-
In: Physical Review B. Condensed Matter and Materials Physics. - : American Physical Society. - 1098-0121 .- 1550-235X. ; 73:10
-
Journal article (peer-reviewed)abstract
- The binary Fe-Cr alloy is a system with a miscibility gap. The decomposition occurs either via the nucleation and growth mechanism or as spinodal decomposition, depending on the Cr content. However, at low chromium concentrations the alloys are anomalously stable. This is shown to be true only for the ferromagnetic body centered cubic (bcc) phase. The stability stems from the negative mixing enthalpy at low concentrations of chromium. We show that the effect has an electronic origin, that is, it is directly related to variations of the electronic structure in the alloy with concentration. We also demonstrate that the variation in the state density of the majority channel at the Fermi level in the concentration interval below 20 at. % Cr indicates increasing tendency of the system towards the spinodal decomposition in the system. Moreover, in the equimolar concentration region, significant deviations of the spin up band from its canonical shape are observed, which destabilize the bcc phase.
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| 36. |
- Olsson, Pär, et al.
(author)
-
Two-band modeling of alpha-prime phase formation in Fe-Cr
- 2005
-
In: Physical Review B. Condensed Matter and Materials Physics. - 1098-0121 .- 1550-235X. ; 72:21
-
Journal article (peer-reviewed)abstract
- We have developed a two-band model of Fe-Cr, fitted to properties of the ferromagnetic alloy. Fitting many-body functionals to the calculated mixing enthalpy of the alloy and the mixed interstitial binding energy in iron, our potential reproduces changes in sign of the formation energy as a function of Cr concentration. When applied in kinetic Monte Carlo simulations, the potential correctly predicts decomposition of initially random Fe-Cr alloys into the alpha-prime phase as function of Cr concentration.
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| 37. |
- Pillon, Sylvie, et al.
(author)
-
Oxide and nitride TRU fuels : Lessons drawn from the CONFIRM and FUTURE projects of the 5th European Framework Program
- 2006
-
In: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 153:3, s. 245-252
-
Journal article (peer-reviewed)abstract
- The FUTURE and CONFIRM projects of the 5th European Framework Program address the issues of the design and fabrication of oxide and nitride fuels, respectively, for the transmutation in an accelerator-driven system (ADS). They started in December 2001 and September 2000, respectively. Advantages and drawbacks of transuranic oxides and nitrides in terms of performance and fabricability have been analyzed. Recommendations on the fuel design will be given and used for the next step of the 6th European Framework Program related to the design and the feasibility assessment of an industrial ADS prototype dedicated to transmutation.
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| 38. |
- Pontikis, V., et al.
(author)
-
An analytic n-body potential for bcc iron
- 2007
-
In: Nuclear Instruments and Methods in Physics Research Section B. - : Elsevier BV. - 0168-583X .- 1872-9584. ; 255:1, s. 37-40
-
Journal article (peer-reviewed)abstract
- We have developed an analytic n-body phenomenological potential for bcc iron made of two electron-density functionals representing repulsion via the Thomas-Fermi free-electron gas kinetic energy term and attraction via a square root functional similar to the second moment approximation of the tight-binding scheme. Electron-density is given by radial, hydrogen-like orbitals with effective charges taken as adjustable parameters fitted on experimental and ab-initio data. Although the set of adjustable parameters is small, prediction of static and dynamical properties of iron is in excellent agreement with the experiments. Advantages and shortcomings of this model are discussed with reference to published works.
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| 39. |
- Reale Hernandez, C., et al.
(author)
-
Dynamic sensitivity and uncertainty analysis of a small lead cooled reactor
- 2020
-
In: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 144
-
Journal article (peer-reviewed)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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| 40. |
- Reale Hernandez, Cuauhtemoc, et al.
(author)
-
Simulation of a loss of flow transient of a small Lead-Cooled reactor using a CFD-Based model
- 2023
-
In: Nuclear Engineering and Design. - : Elsevier BV. - 0029-5493 .- 1872-759X. ; 412
-
Journal article (peer-reviewed)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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| 41. |
- Seltborg, Per, et al.
(author)
-
Definition and application of proton source efficiency in accelerator driven systems
- 2003
-
In: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 145:3, s. 390-399
-
Journal article (peer-reviewed)abstract
- In order to study the beam power amplification of an accelerator-driven system (ADS), a new parameter, the proton source efficiency psi* is introduced. psi* represents the average importance of the external proton source, relative to the average importance of the eigenmode production, and is closely related to the neutron source efficiency rho*, which is frequently used in the ADS field. rho* is commonly used in the physics of subcritical systems driven by any external source (spallation source, (d,d), (d, t), Cf-252 spontaneous fissions, etc.). On the contrary, psi* has been defined in this paper exclusively for ADS studies where the system is driven by a spallation source. The main advantage with using psi* instead of rho* for ADS is that the way of defining the external source is unique and that it is proportional to the core power divided by the proton beam power, independent of the neutron source distribution. Numerical simulations have been performed with the Monte Carlo code MCNPX in order to study psi* as a function of different design parameters. It was found that, in order to maximize psi* and therefore minimize the proton current needs, a target radius as small as possible should be chosen. For target radii smaller than similar to30 cm, lead-bismuth is a better choice of coolant material than sodium, regarding the proton source efficiency, while for larger target radii the two materials are equally good. The optimal axial proton beam impact was found to be located similar to 20 cm above the core center. Varying the proton energy, psi*/E-p was found to have a maximum for proton energies between 1200 and 1400 MeV Increasing the americium content in the fuel decreases psi* considerably, in particular when the target radius is large.
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| 42. |
- Seltborg, Per, et al.
(author)
-
Proton source efficiency for heterogeneous distribution of actinides in the core of an accelerator-driven system
- 2006
-
In: Nuclear science and engineering. - 0029-5639 .- 1943-748X. ; 154:2, s. 202-214
-
Journal article (peer-reviewed)abstract
- The distribution of actinides in the core of an accelerator-driven system loaded with plutonium, americium, and curium has been studied in order to optimize the proton source efficiency psi*. The optimization of psi* was performed by keeping some important characteristics of the system, e.g., the radial power profile and the reactivity of the core, constant. One of the basic assumptions of the study, that the magnitude of psi* is sensitive primarily to the composition of actinides in the inner part of the core, whereas only marginally to that in the outer part, has been confirmed. It has been shown that the odd-N nuclides (those nuclides with an even number of neutrons) in general and Am-241 and Cm-244 in particular have favorable properties with respect to improving psi* if they are placed in the innermost part of the core. The underlying reason for this phenomenon is that the energy spectrum of the source neutrons in the inner part of the core is harder than that of the average fission neutrons. Moreover, it has been shown that loading the inner part of the core with only curium increases psi* by similar to 7%. Plutonium, on the other hand, in particular high-quality plutonium consisting mainly of Pu-239 and Pu-241, was found to be a comparatively source inefficient element and is preferably located in the outer part of the core. The differences in psi* are due to combined effects from relative changes in the average fission and capture cross sections and in the average fission neutron yield.
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| 43. |
- Suvdantsetseg, Erdenechimeg, 1983-, et al.
(author)
-
An assessment of prompt neutron reproduction time in a reflector dominated fast critical system : ELECTRA
- 2014
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 71, s. 159-165
-
Journal article (peer-reviewed)abstract
- In this paper, an accurate method to evaluate the prompt neutron reproduction time for a reflector dominated fast critical reactor, ELECTRA, is discussed. To adequately handle the problem, explicit time dependent Monte Carlo calculations with MCNP, applying repeated time cut-off technique, is used and compared against the σ ∼ 1/v time dependent absorber method, applying artificial cross section data in the Monte Carlo code SERPENT. The results show that when a reflector plays a major role in criticality for fast neutron reactor, the two methods predict different physical parameters (Λ = 69 ± 2 ns and Λ = 83 ± 1 ns for time cut-off and the 1/v method respectively). The reason is explained by applying Avery-Cohn’s two-region prompt neutron model.
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| 44. |
- Suvdantsetseg, Erdenechimeg, 1983-, et al.
(author)
-
Optimization of the reactivity control drum system of ELECTRA
- 2012
-
In: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 252:1, s. 209-214
-
Journal article (peer-reviewed)abstract
- In this paper, an optimized control drum system for the European Lead Cooled Training Reactor (ELECTRA) is proposed. By changing the number of rotating drums from 6 to 12, we succeed in reducing the maximum rotational worth of a single drum from 4 $ to 1.64 $. As a consequence, the safety margin during reactivity insertions is significantly improved.
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| 45. |
|
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| 46. |
- Suvdantsetseg, Erdenechimeg, 1983-, et al.
(author)
-
Unprotected Loss of Heat Sink Analysis for the Design of ELECTRA
-
In: Annals of Nuclear Energy. - 0306-4549 .- 1873-2100.
-
Journal article (other academic/artistic)abstract
- In this paper, the transient behavior of a small fast reactor cooled by natural convection of lead (ELECTRA) is studied under unprotected loss of heat sink conditions, as a parametric function of the radiative heat loss via its outer vessel surface. The results show that when this heat loss is higher than the decay heat level, the negative reactivity feedback mechanisms of the system lead to re-criticality, bringing it into a new steady state. As a consequence, coolant freezing may be avoided during such accidents.
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| 47. |
- Tucek, Kamil, et al.
(author)
-
Studies of an accelerator-driven transuranium burner with hafnium-based inert matrix fuel
- 2007
-
In: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 157:3, s. 277-298
-
Journal article (peer-reviewed)abstract
- Neutronic and burnup characteristics of an accelerator-driven transuranium burner in a startup mode were studied. Different inert and absorbing matrices as well as lattice configurations were assessed in order to identify suitable fuel and core design configurations. Monte Carlo transport and burnup codes were used in the analyses. The lattice pin pitch was varied to optimize the source efficiency and coolant void worth while respecting key thermal and material-related design constraints posed by fuel and cladding. A HfN matrix appeared to provide a good combination of neutronic, burnup, and safety characteristics: maintaining a hard neutron spectrum, yielding acceptable coolant void reactivity and source efficiency, and alleviating the burnup reactivity swing. A conceptual design of a (TRU,Hf)N fueled, lead-bismuth eutectic-cooled accelerator-driven system was developed. Twice higher neutron fission-to absorption probabilities in Am isotopes were achieved compared to reactor designs relying on ZrN or YN inert matrix fuel. The production of higher actinides in the fuel cycle is hence limited, with a Cm fraction in the equilibrium fuel being similar to 40% lower than for cores with ZrN matrix-based fuel. The burnup reactivity swing and associated power peaking in the core are managed by an appropriate choice of cycle length (100 days) and by core enrichment zoning. A safety analysis shows that the system is protected from instant damage during unprotected beam overpower transient.
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| 48. |
- Wallenius, Janne, 1968-, et al.
(author)
-
A new paradigm for breeding of nuclear fuel
- 2019
-
In: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 133, s. 816-819
-
Journal article (peer-reviewed)abstract
- Breeding of nuclear fuel from fertile nuclides may allow to extend known nuclear fuel resources from a century to thousands of years. In this article, we show that the requirement for a breeder reactor fuel to feature an effective neutron production per absorption larger than 2 (eta > 2) breaks down for fertile and fissionable nuclides meeting two criteria related to the relative magnitude of capture and fission cross sections. Moreover, we find that a breeding ratio larger than unity can be achieved for fuels consisting of a single nuclide, in spite of the this nuclide featuring eta < 2. In particular, neptunium is identified as a nuclear fuel that can sustain a reactivity increase over time up to a burn-up exceeding 240 GWd/ton in a fast neutron spectrum.
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| 49. |
- Wallenius, Janne, 1968-, et al.
(author)
-
A small lead-cooled reactor with improved Am-burning and non-proliferation characteristics
- 2018
-
In: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 122, s. 193-200
-
Journal article (peer-reviewed)abstract
- In this paper, a novel approach for transmutation of americium in fast reactors is presented. Using enriched uranium as fissile support, rather than plutonium, it is shown that a minor actinide burning rate of 25 kg/TWh(th) is possible to achieve in a passively safe, critical lead-cooled reactor. Moreover, the plutonium produced by transmutation of Am-241 features up to 38% (PU)-P-238, making it difficult to use for weapons production.
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| 50. |
- Wallenius, Janne, 1968-
(author)
-
An improved correlation for gas release from nitride fuels
- 2022
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 558
-
Journal article (peer-reviewed)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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