| 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
-
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.
|
|
| 2. |
- Sandberg, Nils, et al.
(author)
-
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.
|
|
| 3. |
- Wallenius, Janne, 1968-, et al.
(author)
-
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.
|
|
| 4. |
- Ackland, G. J., et al.
(author)
-
Interatomic forces for transition metals including magnetism
- 2010
-
In: 139th Annual Meeting & Exhibition - Supplemental Proceedings, Vol 2. - 9780873397520 ; , s. 85-92
-
Conference paper (peer-reviewed)abstract
- We present a formalism for extending the second moment tight-binding model[1], incorporating ferro- and anti-ferromagnetic interaction terms which are needed for the FeCr system. For antiferromagnetic and paramagnetic materials, an explicit additional variable representing the spin is required. In a mean-field approximation this spin can be eliminated. and the potential becomes explicitly temperature dependent. For ferromagnetic interactions, this degree of freedom can be eliminated, and the formalism reduces to the embedded atom method (EAM[2]), and we show the equivalence of existing EAM potentials to "magnetic" Potentials.
|
|
| 5. |
- Berglöf, Carl, 1980-
(author)
-
On measurement and monitoring of reactivity in subcritical reactor systems
- 2010
-
Doctoral thesis (other academic/artistic)abstract
- Accelerator-driven systems have been proposed for incineration of transuranic elements from spent nuclear fuel. For safe operation of such facilities, a robust method for reactivity monitoring is required. Experience has shown that the performance of reactivity measurement methods in terms of accuracy and applicability is highly system dependent. Further investigations are needed to increase the knowledge data bank before applying the methods to an industrial facility and to achieve license to operate such a facility. In this thesis, two systems have been subject to investigation of various reactivity measurement methods. Conditions for successful utilization of the methods are presented, based on the experimental experience. In contrast to previous studies in this field, the reactivity has not only been determined, but also monitored based on the so called beam trip methodology which is applicable also to non-zero power systems. The results of this work constitute a part of the knowledge base for the definition of a validated online reactivity monitoring methodology for facilities currently being under development in Europe (XT-ADS and EFIT).
|
|
| 6. |
- Bortot, Sara, et al.
(author)
-
BELLA : a multi-point dynamics code for safety-informed design of fast reactors
- 2015
-
In: Annals of Nuclear Energy. - : PERGAMON-ELSEVIER SCIENCE LTD. - 0306-4549 .- 1873-2100. ; 85, s. 228-235
-
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. (C) 2015 Elsevier Ltd. All rights reserved.
|
|
| 7. |
- Bortot, Sara, 1983-, et al.
(author)
-
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.
|
|
| 8. |
|
|
| 9. |
- Bubelis, E., et al.
(author)
-
System codes benchmarking on a low sodium void effect SFR heterogeneous core under ULOF conditions
- 2017
-
In: Nuclear Engineering and Design. - : Elsevier. - 0029-5493 .- 1872-759X. ; 320, s. 325-345
-
Journal article (peer-reviewed)abstract
- This paper discusses system codes benchmarking activities on an ASTRID-like heterogeneous fast core under a representative design basis accident condition: the unprotected loss of flow accident (ULOF). The paper provides evidence that all the system codes used in this exercise are capable to simulate the transient behavior of heterogeneous SFR cores up to the initiation of sodium boiling. As a proof of this, a comparison of steady-state results and dynamic simulation results for a ULOF transient (simulated using system codes in combination with neutron point kinetics) are provided and discussed in this paper. The paper contains a brief description of the system codes (TRACE, CATHARE, SIM-SFR, SAS-SFR, ATHLET, SPECTRA, SAS4A) used by the participants (PSI, CEA, EDF, KIT, GRS, UPVLC, NRG, KTH), assumptions made during the simulations, as well as results obtained.
|
|
| 10. |
- Claisse, Antoine, 1988-
(author)
-
Multiscale modeling of nitride fuels
- 2016
-
Doctoral thesis (other academic/artistic)abstract
- Nitride fuels have always been considered a good candidate for GENIV reactors, as well as space reactors, due to their high fissile density, highthermal conductivity and high melting point. In these concepts, not beingcompatible with water is not a significant problem. However, in recent years,nitride fuels started to raise an interest for application in thermal reactors,as accident tolerant or high performance fuels. However, oxide fuels havebenefited from decades of intensive research, and thousands of reactor-years.As such, a large effort has to be made on qualifying the fuel and developingtools to help assess their performances.In this thesis, the modeling side of this task is chosen. The effort istwo-fold: determining fundamental properties using atomistic models andputting together all the properties to predict the performances under irradi-ation using a fuel performance code. The first part is done combining manyframeworks. The density functional theory is the basis to compute the elec-tronic structure of the materials, to which a Hubbard correction is added tohandle the strong correlation effects. Negative side effects of the Hubbardcorrection are tackled using the so-called occupation matrix control method.This combined framework is first tested, and then used to find electronic andmechanic properties of the bulk material as well as the thermomechanicalbehavior of foreign atoms. Then, another method, the self-consistent meanfield (SCMF) one, is used to reach the dynamics properties of these foreignatoms. In the SCMF theory, the data that were obtained performing the abinitio simulations are treated to provide diffusion and kinetic flux couplingproperties.In the second step of the work, the fuel performance code TRANSURA-NUS is used to model complete fuel pins. An athermal fission gas releasemodel based on the open porosity is developed and tested on oxide fuels.A model for nitride fuels is introduced, and some correlations are bench-marked. Major issues remaining are pointed out and recommendations asto how to solve them are made.
|
|
| 11. |
- Costa, Diogo Ribeiro, et al.
(author)
-
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.
|
|
| 12. |
- Costa, Diogo Ribeiro, et al.
(author)
-
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.
|
|
| 13. |
- Costa, Diogo Ribeiro, 1983-
(author)
-
Development of Encapsulated UN-UO₂ Accident Tolerant Fuel
- 2023
-
Doctoral thesis (other academic/artistic)abstract
- Accident tolerant fuels (ATFs) are designed to endure a severe accident in the reactor core longer than the standard UO2-Zr alloy systems used in light water reactors (LWRs). Composite fuels such as UN-UO2 are being considered as an ATF concept to address the lower oxidation resistance of the UN fuel from a safety perspective for use in LWRs, whilst improving the in-reactor behaviour of the UO2 fuel. The main objective of this thesis is to fabricate, characterise, and evaluate an innovative ATF concept for LWRs: encapsulated UN spheres as additives for the standard UO2 fuel. Several development steps were applied to understand the influence of the sintering parameters on the UN-UO2 fuel microstructure, evaluate potential coating candidates to encapsulate the UN spheres by different coating methodologies, assess the oxidation resistance of the composites, and estimate the thermal behaviours of uncoated and encapsulated UN-UO2 fuels. All composites were sintered by the spark plasma sintering method and characterised by many complementary microstructural techniques. Molybdenum and tungsten are shown, using a combination of modelling and experiments, to be good material candidates for the protective coating. It is shown that the powder coating methods form a thick, dense, and non-uniform coating layer onto spheres, while the chemical and vapour deposition methods provide thinner and more uniform layers. Finite element modelling indicates that the fuel centreline temperature may be reduced by more than 400 K when 70 wt% of encapsulated spheres are used as compared to the reference UO2. Moreover, the severity of the degradation of the nitride phase is reduced when embedded in a UO2 matrix and may also be reduced even more by the presence of a coating layer. These results contribute to further developments in methodologies for fabricating, characterising, and evaluating accident tolerant fuels within LWRs.
|
|
| 14. |
- 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.
|
|
| 15. |
|
|
| 16. |
- 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.
|
|
| 17. |
- 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.
|
|
| 18. |
- De Bruyn, D., et al.
(author)
-
Main achievements of the FP7-LEADER collaborative project of the european commission regarding the design of a lead-cooled fast reactor
- 2013
-
In: International Congress on Advances in Nuclear Power Plants, ICAPP 2013. - 9781632660381 ; , s. 281-290
-
Conference paper (peer-reviewed)abstract
- Concerns over energy resource availability, climate change, air quality, and energy security suggest an important role for nuclear power in future energy supplies. While the current Generation II and III nuclear power plant designs provide an economically and publicly acceptable electricity supply in many markets, further advances in nuclear energy system design can broaden the opportunities for the use of nuclear energy. To explore these opportunities, worldwide governments, industries, and research centres started a wide-ranging discussion on the development of new systems known as "Generation IV." The European Commission has organized the Sustainable Nuclear Energy Technology Platform that through its Strategic Research Agenda promoted the development of fast reactors with closed fuel cycle. Among the promising reactor technologies, the Lead Fast Reactor (LFR) has been identified as a technology with great potential to meet needs for both remote sites and central power stations. The LFR system features a fast-neutron spectrum allowing the possibility for a closed fuel cycle for efficient conversion of fertile uranium and management of actinides. A full actinide recycle fuel cycle is therefore envisioned for the design of the reference LFR meant for deployment, while the capabilities of the system to act as a net-burner of actinides from spent fuel are object of further investigation The LEADER project deals with the development of such a technology through two main goals: the conceptual design of an industrial-size LFR (the so-called European LFRor ELFR) and the conceptual design of a scaled down facility, the demonstration reactor called ALFRED (Advanced Lead Fast Reactor European Demonstrator). The European Commission, withinits seventh framework programme, has approved the proposal submitted by 16 partners comprising research centres, industrial partners and universities. The project has started in April 2010 for a duration of three years.The focus of the first part of the LEADER project was the resolution of the key issues of the previous sixth framework programme ELSY project in order to reach a new consistent industrial-size reactor ELFR configuration.With reference to this reactor configuration the design of the ALFRED demonstrator (sized at 300 MWth, about 120 MWe) has been performed. The development of such demonstrator reactor presents obviously strong and interesting synergies with the development of MYRRHA, a material and fuel testing facility proposed by the SCK·CEN research centre in Belgium. In this paper we present a synthesis of the main results of the LEADER project.
|
|
| 19. |
- 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.
|
|
| 20. |
- 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.
|
|
| 21. |
|
|
| 22. |
- Delage, F., et al.
(author)
-
ADS fuel developments in Europe : Results from the EUROTRANS integrated project
- 2011
-
In: Energy Procedia. - : Elsevier BV. ; , s. 303-313
-
Conference paper (peer-reviewed)abstract
- Fuels to be used in Accelerator Driven Systems dedicated to Minor Actinides transmutation can be described as highly innovative in comparison with those used in critical cores. Indeed, ADS fuels are not fertile, so as to improve the transmutation performance and they contain high volumetric concentrations (∼50%) of minor actinides and plutonium compounds. This unusual fuel composition results in high gamma and neutron emissions during its fabrication, as well as degraded performances under irradiation. Ceramic-Ceramic and Ceramic Metallic composite fuels consisting of particles of (Pu, MA)O2 phases dispersed in a magnesia or molybdenum matrix were investigated within the European Research programme for Transmutation, as driver fuels for a prospective 400MWth transmuter: the European Facility for Industrial Transmutation. Fuel performances and safety of preliminary core designs were evaluated to support the project. Out -of-pile as well as in-pile experiments were carried out to gain essential knowledge on properties and behaviour under irradiation of these types of fuel. This paper gives an overview of experimental results within the project.
|
|
| 23. |
- Dudarev, S. L., et al.
(author)
-
The EU programme for modelling radiation effects in fusion reactor materials : An overview of recent advances and future goals
- 2009
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 386, s. 1-7
-
Journal article (peer-reviewed)abstract
- The EU fusion materials modelling programme was initiated in 2002 with the objective of developing a comprehensive set of computer modelling techniques and approaches, aimed at rationalising the extensive available experimental information on properties of irradiated fusion materials, developing capabilities for predicting the behaviour of materials under conditions not yet accessible to experimental tests, assessing results of tests involving high dose rates, and extrapolating these results to the fusion-relevant conditions. The programme presently gives emphasis to modelling a single class of materials, which are ferritic-martensitic EUROFER-type steels, and focuses on the investigation of key physical phenomena and interpretation of experimental observations. The objective of the programme is the development of computational capabilities for predicting changes in mechanical properties, hardening and embrittlement, as well as changes in the microstructure and phase stability of EUROFER and FeCr model alloys occurring under fusion reactor relevant irradiation conditions.
|
|
| 24. |
- 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.
|
|
| 25. |
- Eriksson, Marcus, 1972-
(author)
-
Accelerator-driven systems : safety and kinetics
- 2005
-
Doctoral thesis (other academic/artistic)abstract
- The accelerator-driven system (ADS) is recognized as a promising system for the purpose of nuclear waste transmutation and minimization of spent fuel radiotoxicity. The primary cause for this derives from its accelerator-driven, sub-critical operating state, which introduces beneficial safety-related features allowing for application of cores employing fuel systems containing pure transuranics or minor actinides, thereby offering increased incineration rate of waste products and minimal deployment of advanced (and expensive) partitioning and transmutation technologies. The main theme of the thesis is safety and kinetics performance of accelerator-driven nuclear reactors. The studies are confined to the examination of ADS design proposals employing fast neutron spectrum, uranium-free lattice fuels, and liquid-metal cooling, with emphasis on lead-bismuth coolant. The thesis consists of computational studies under normal operation and hypothetical accidents, and of evaluation and identification of safety design features. By itself, subcritical operation provides a distinct safety advantage over critical reactor operation, distinguished by high operational stability and additional margins for positive reactivity insertion. For a uranium-free minor actinide based fuel important safety parameters deteriorate. Specific analyses suggest that operation of such cores in a critical state would be very difficult. The studies of unprotected transients indicate that lead-bismuth cooled accelerator-driven reactors can be effective in addressing the low effective delayed neutron fraction and the high coolant void reactivity that comes with the minor actinide fuel, but some supportive prompt negative feedback mechanism might be considered necessary to compensate for a weak Doppler effect in case of a prompt critical transient. Although lead-bismuth features a high boiling point, the work underlines the importance of maintaining a low coolant void reactivity value. The transient design studies identified a molybdenum-based Ceramic-Metal (CerMet) fuel with favourable inherent safety features. A higher lattice pitch is suggested to avoid mechanical failure during unprotected loss-of-flow. Detailed coupled neutron kinetics and thermal hydraulic analyses demonstrated that the point kinetics approximation is capable of providing highly accurate transient calculations of subcritical systems. The results suggest better precision at lower keff levels, which is an effect of the reduced sensitivity to system reactivity perturbations in a subcritical state resulting in small spatial distortions. In the course of a beam reliability study, the accelerator was identified as responsible for frequent beam interruptions. It is clear that extensive improvement in the mean-time between beam failures is required.
|
|
| 26. |
- Eriksson, Marcus, et al.
(author)
-
Inherent Safety of Fuels for Accelerator-driven Systems
- 2005
-
In: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 151:3, s. 314-333
-
Journal article (peer-reviewed)abstract
- Transient safety characteristics of accelerator-driven systems using advanced minor actinide fuels have been investigated. Results for a molybdenum-based Ceramic-Metal (CerMet) fuel, a magnesia-based Ceramic-Ceramic fuel, and a zirconium-nitride-based fuel are reported. The focus is on the inherent safety aspects of core design. Accident analyses are carried out for the response to unprotected loss-of-flow and accelerator beam-overpower transients and coolant voiding scenarios. An attempt is made to establish basic design limits for the fuel and cladding. Maximum temperatures during transients are determined and compared with design limits. Reactivity effects associated with coolant void, fuel and structural expansion, and cladding relocation are investigated. Design studies encompass variations in lattice pitch and pin diameter. Critical mass studies are performed. The studies indicate favorable inherent safety features of the CerMet fuel. Major consideration is given to the potential threat of coolant voiding in accelerator-driven design proposals. Results for a transient test case study of a postulated steam generator tube rupture event leading to extensive coolant voiding are presented. The study underlines the importance of having a low coolant void reactivity value in a lead-bismuth system despite the high boiling temperature of the coolant. It was found that the power rise following a voiding transient increases dramatically near the critical state. The studies suggest that a reactivity margin of a few dollars in the voided state is sufficient to permit significant reactivity insertions.
|
|
| 27. |
- Eriksson, Marcus, et al.
(author)
-
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.
|
|
| 28. |
- Fokau, Andrei, et al.
(author)
-
A source efficient ADS for minor actinides burning
- 2010
-
In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 37:4, s. 540-545
-
Journal article (peer-reviewed)abstract
- Taking advantage of the good neutron economy of nitride fuel, a compact accelerator-driven system (ADS) for burning of minor actinide fuels has been designed, based on the fuel assembly geometry developed for the European Facility for Industrial Transmutation (EFIT) within the EUROTRANS project. The small core size of the new design permits reduction of the size of the spallation target region, which enhances proton source efficiency by about 80% compared to the reference oxide version of EFIT. Additionally, adoption of the austenitic steel 15/15Ti as clad material allows to safely reduce the fuel pin pitch, which leads to an increase of fuel volume fraction and therefore makes the neutron energy spectrum faster, consequently increasing minor actinides fission probabilities. Our calculations show that one can dramatically increase neutron source efficiency up to 0.95 without a significant loss of neutron source intensity, i.e. having high proton source efficiency. Consequently, the accelerator current required for operation of the ADS with a fission power of 201 MWth and a burn-up of 27 GW d/t per year (365 EFPD) is reduced by 67%.
|
|
| 29. |
|
|
| 30. |
- Gudowski, Waclaw, et al.
(author)
-
Review of the European project - Impact of Accelerator-Based Technologies on Nuclear Fission Safety (IABAT)
- 2001
-
In: Progress in nuclear energy (New series). - 0149-1970 .- 1878-4224. ; 38:1-2, s. 135-151
-
Journal article (peer-reviewed)abstract
- The IABAT project - Impact of Accelerator Based Technologies on Nuclear Fission Safety - started in 1996 in the frame of 4(th) Framework Programme of the European Union, R&D specific programme Nuclear fission safety 1994-1998, area A.2 Exploring innovative approaches/Fuel cycle concepts, as one of the first common European activities in ADS. The project was completed October 31, 1999. The overall objective of the IABAT project has been a preliminary assessment of the potential of Accelerator-Driven Systems (ADS) for transmutation of nuclear waste and for nuclear energy production with minimum waste generation. Moreover, more specific topics related to nuclear data and code development for ADS have been studied in more detail. Four ADSs have been studied for different fuel/coolant combinations: liquid metal coolant and solid fuel, liquid metal coolant and dispersed fuel, and fast and thermal molten salt systems. Target studies comprised multiple target solutions and radiation damage problems in a target environment. In a tool development part of the project a methodology of subcriticality monitoring has been developed based on Feynman-alpha and Rossi-alpha methods. Moreover, a new Monte-Carlo burnup code taking full advantage of continuous neutron cross-section data has been developed and benchmarked. Impact on the risk from high-level waste repositories fi om radiotoxicity reduction using ADS has been assessed giving no crystal-clear benefits of ADS for repository radiotoxicity reduction but concluding some important prerequisites for effective transmutation. In proliferation studies important differences between critical reactors and ADS have been underlined and non-proliferation measures have been proposed. In assessment of accelerator technology costing models have been created that allow the circular and linear accelerator options to be compared and the effect of parameter variations examined. The calculations reported show that cyclotron systems would be more economical, due mainly to the advantage of the cost of RF power supplies. However, the accelerator community regards with skepticism the possibility of transporting and extracting more than a 10mA beam current from a 1GeV cyclotron and therefore technical factors may limit the application of cyclotrons. Finally, this review summarizes development of nuclear data in the energy region between 20 Mev and 150 MeV. Neutron and proton transport data files for Fe, Ni, Pb, Th, U-238 and Pu-239 have been created. The high-energy part of the data files consists completely of results from model calculations, which are benchmarked against the available experimental data. Although there is obviously future work left regarding fine-tuning of several parts of the data files, the representation of nuclear reaction information up to 150 MeV is already better than can be attained with intranuclear cascade codes.
|
|
| 31. |
|
|
| 32. |
|
|
| 33. |
- Hania, P. R., et al.
(author)
-
Irradiation and post-irradiation examination of uranium-free nitride fuel
- 2015
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 466, s. 597-605
-
Journal article (peer-reviewed)abstract
- Two identical Phénix-type 15-15Ti steel pinlets each containing a 70 mm Pu0.3Zr0.7N fuel stack in a 1-bar helium atmosphere have been irradiated in the HFR Petten at medium high linear power (46-47 kW/m at BOL) and an average cladding temperature of 505 °C. The pins were irradiated to a plutonium burn-up of 9.7% (88 MWd/kgHM) in 170 full power days. Both pins remained fully intact. Post-irradiation examination performed at NRG and PSI showed that the overall swelling rate of the fuel was 0.92 vol-%/%FIHMA. Fission gas release was 5-6%, while helium release was larger than 50%. No fuel restructuring was observed, and only mild cracking. EPMA measurements show a burn-up increase toward the pellet edge of up to 4 times. All investigated fission products except to some extent the noble metals were found to be evenly distributed over the matrix, indicating good solubility. Local formation of a secondary phase with high Pu content and hardly any Zr was observed. A general conclusion of this investigation is that ZrN is a suitable inert matrix for burning plutonium at high destruction rates.
|
|
| 34. |
- 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.
|
|
| 35. |
- 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.
|
|
| 36. |
- 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.
|
|
| 37. |
- Johnson, Kyle D., et al.
(author)
-
Fabrication and microstructural analysis of UN-U3Si2 composites for accident tolerant fuel applications
- 2016
-
In: Journal of Nuclear Materials. - : Elsevier. - 0022-3115 .- 1873-4820. ; 477, s. 18-23
-
Journal article (peer-reviewed)abstract
- In this study, U3Si2 was synthesized via the use of arc-melting and mixed with UN powders, which together were sintered using the SPS method. The study revealed a number of interesting conclusions regarding the stability of the system - namely the formation of a probable but as yet unidentified ternary phase coupled with the reduction of the stoichiometry in the nitride phase - as well as some insights into the mechanics of the sintering process itself. By milling the silicide powders and reducing its particle size ratio compared to UN, it was possible to form a high density UN-U3Si2 composite, with desirable microstructural characteristics for accident tolerant fuel applications.
|
|
| 38. |
- Johnson, Kyle D., 1986-
(author)
-
High Performance Fuels for Water-Cooled Reactor Systems
- 2016
-
Doctoral thesis (other academic/artistic)abstract
- Investigation of nitride fuels and their properties has, for decades, been propelled on the basis of their desirable high metal densities and high thermal conductivities, both of which oer intrinsic advantages to performance, economy, and safety in fast and light water reactor systems. In this time several key obstacles have been identied as impeding the implementation of these fuels for commercial applications; namely chemical interactions with air and steam, the noted diculty in sintering of the material, and the high costs associated with the enrichment of 15N. The combination of these limitations, historically, led to the well founded conclusion that the most appropriate use of nitride fuels was in the fast reactor fuel cycle, where the cost burdens associated with them is substantially less. Indeed, it is within this context that the vast majority of work on nitrides has been and continues to be done.Nevertheless, following the 2011 Fukushima-Daiichi nuclear accident, a concerted governmental-industrial eort was embarked upon to explore the alternatives of so-called \accident tolerant" and \high performance" fuels. These fuels would, at the same time, improve the response of the fuel-clad system to severe accidents and improve the economy of operation for light water reactor systems. Among the various candidates proposed are uranium nitride, uranium silicide, and a third \uranium nitride-silicide" composite featuring a mixture of the former.In this thesis a method has been established for the synthesis, fabrication, and characterization of high purity uranium nitride, and uranium nitride-silicide composites, prepared by the spark plasma sintering (SPS) technique. A specic result has been to isolate the impact of the processing parameters on the microstructure of representative fuel pellets, essentially permitting any conceivable microstructure of interest to be fabricated. This has enabled the development of a highly reproducible technique for the production of pellets with microstructures tailored towards any desired porosity between 88-99.9%TD, any grain size between 6-24 μm, and, in the case of the uranium nitride-silicide composite, a silicide-coated UN matrix. This has permitted the evaluation of these microstructural characteristics on the performance of these materials, specically with respect to their role as accident tolerant fuels. This has generated results which have tightly coupled nitride performance with pellet microstructure, with important implications for the use of nitrides in water-cooled reactors.
|
|
| 39. |
- Johnson, Kyle D., et al.
(author)
-
Spark plasma sintering and porosity studies of uranium nitride
- 2016
-
In: Journal of Nuclear Materials. - : Elsevier. - 0022-3115 .- 1873-4820. ; 473, s. 13-17
-
Journal article (peer-reviewed)abstract
- In this study, a number of samples of UN sintered by the SPS method have been fabricated, and highly pure samples ranging in density from 68% to 99.8%TD-corresponding to an absolute density of 14.25 g/cm3 out of a theoretical density of 14.28 g/cm3-have been fabricated. By careful adjustment of the sintering parameters of temperature and applied pressure, the production of pellets of specific porosity may now be achieved between these ranges. The pore closure behaviour of the material has also been documented and compared to previous studies of similar materials, which demonstrates that full pore closure using these methods occurs near 97.5% of relative density.
|
|
| 40. |
- 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.
|
|
| 41. |
|
|
| 42. |
- Jolkkonen, Mikael, et al.
(author)
-
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.
|
|
| 43. |
- Jolkkonen, Mikael, et al.
(author)
-
Uranium nitride fuels in superheated steam
- 2017
-
In: Journal of Nuclear Science and Technology. - : Informa UK Limited. - 0022-3131 .- 1881-1248.
-
Journal article (peer-reviewed)abstract
- Uranium mononitride (UN) pellets of different densities were subjected to a superheatedsteam/argon mixture at atmospheric pressure to evaluate their resistance to hydrolysis. Completedegradation of pure UN pellets was obtained within 1 hour in 0.50 bar steam at 500 °C. Theidentified reaction products were uranium dioxide, ammonia and hydrogen gas, with no detectableamounts of nitrogen oxides formed. However, the reaction could not be carried to completion, andthe presence of uranium sesquinitride and higher uranium oxides or uranium oxynitrides in the solidresidue is indicated. Evolution of elemental nitrogen was seen in connection with very high reactionrates. The porosity of the pellets was identified as the most important factor determining reactionrates at 400 – 425 °C, and it is suggested that in dense pellets, cracking due to internal volumeincrease initiates a transition from slow surface corrosion to pellet disintegration. The implicationsfor the use of nitride fuels in light water reactors are discussed, with some observations concerninghydrolysis as a method for 15N recovery from isotopically enriched spent nitride fuel.
|
|
| 44. |
|
|
| 45. |
- Lambrinou, K., et al.
(author)
-
Corrosion-resistant ternary carbides for use in heavy liquid metal coolants
- 2016
-
In: Ceramic Engineering and Science Proceedings. - 9781119040439 ; , s. 19-34
-
Conference paper (peer-reviewed)abstract
- A primary concern in the development of accelerator-driven systems (ADS) with liquid leadbismuth eutectic (LBE) spallation target and Gen-IV lead-cooled fast reactors (LFRs) is the compatibility of the candidate structural steels with the heavy liquid metal (HLM) coolant In the accelerator-driven system MYRRHA, the envisaged primary coolant is liquid LBE, a potentially corrosive environment for various nuclear grade steels. The inherent LBE corrosiveness is the driving force behind diverse research incentives aiming at the development of corrosion-resistant materials for specific applications. I3ue to their superb corrosion resistance in contact with liquid LBE, MAX phases are currently being assessed as candidate materials for the construction of pump impellers suitable for MYRRHA and Gen-IV LFRs. In the case of the MYRRHA nuclear system, the pump impeller will be called to operate reliably at ∼270°C in contact with moderately-oxygenated (concentration of dissolved oxygen: [O] ≥ 7×10-7 mass%), fast-flowing LBE (LBE flow velocity: v ≈ 10-20 m/s locally on the impeller surface). Selected MAX phases are currently being screened with respect to their capability of meeting the targeted material property requirements, especially the enhanced erosion resistance requested by this particular application. This work gives a state-of-the-art overview of the processing and characterisation of selected MAX phases that are screened as candidate structural materials for the MYRRHA pump impeller. All considered MAX phases were produced via a powder metallurgical route and their performance was assessed by various mechanical tests in air/vacuum and corrosion/erosion tests in liquid LBE.
|
|
| 46. |
- 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.
|
|
| 47. |
- 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.
|
|
| 48. |
- Malerba, L., et al.
(author)
-
Modelling of radiation damage in Fe-Cr alloys
- 2007
-
In: Journal of ASTM International. - 1546-962X. ; 4:6
-
Journal article (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.
|
|
| 49. |
- 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.
|
|
| 50. |
- Malerba, Lorenzo, et al.
(author)
-
Multiscale modelling of radiation damage and phase transformations : The challenge of FeCr alloys
- 2008
-
In: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 382:2-3, s. 112-125
-
Journal article (peer-reviewed)abstract
- We review the experimental evidence of the non-monotonic behaviour of FeCr alloys versus Cr content, particularly under irradiation (ordering versus segregation tendencies, microstructure and phase evolution, hardening and embrittlement), together with the theoretical efforts done at the electronic and atomic level to interpret them. We summarize the achievements of the two interatomic potentials developed for this system and perform a careful scrutiny of their limitations. We emphasise the difficulties related to the study, at the atomic-level, of concentrated alloys and propose routes to overcome them. Finally, we advance some opinions regarding the crucial points that deserve further investigation in order to fully understand this important binary alloy, at the basis of the steels for current and future nuclear applications.
|
|
