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Issues and considerations for fuel cladding materials of LFR reactor P. Agostini, A. Gessi, D. Rozzia, M.Tarantino – ENEA Contributions by participants.

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Presentation on theme: "Issues and considerations for fuel cladding materials of LFR reactor P. Agostini, A. Gessi, D. Rozzia, M.Tarantino – ENEA Contributions by participants."— Presentation transcript:

1 Issues and considerations for fuel cladding materials of LFR reactor P. Agostini, A. Gessi, D. Rozzia, M.Tarantino – ENEA Contributions by participants of MATTER Project LEADER Meeting Petten, February 2013 1

2 2 Primary Vessel Tnom : 380-430°C Damage modes: corrosion LM embrittlement, ratchetting,fatigue, creep Pump Tnom : 380-480°C Damage modes: Erosion-corrosion, ratchetting, fatigue Steam Generator Tnom : 380-480°C (Pb) – 450°C (steam) Damage modes: corrosion, LM embrittlement, ratchetting, fatigue, creep-fatigue, buckling, Inner Vessel Tnom : 380°C- 480°C Damage modes: corrosion, ratchetting, buckling, creep-fatigue Overview of damage modes in a LFR Fuel Assembly claddings Tnom : 380°C- 550°C Damage modes: irradiation damage (swelling, creep, embrittlement), thermal creep, thermal fatigue, LM corrosion, LM embrittlement Fuel Assembly Structures Tnom : 380°C- 530°C Damage modes: irradiation damage (swelling, creep, embrittlement), LMcorrosion, thermal creep, LM embrittlement

3 FUEL CLADDING CONDITIONS Fuel criteria defined in ELSY EU Project Max allowed peak linear power 32kW/m; Max clad and fuel temperatures of 560 °C and 2100 °C, respectively; Max neutron flux 2.4*10 15 n/cm 2 s Peak clad damage of 100 dpa, in correspondence of a fuel burn-up of 100 MWd/kgHM (200 dpa are assumed as a long term option); Fuel pin OD 10.5 mm, overall length 2520 mm Hoop stress to be examined for creep 160 MPa (200 Mpa as long term option)

4 Neutron spectrum of LFR core 4

5 Irradiation swelling of the cladding tubes Phenix experience on cladding materials exposed at high neutron flux Excessive swelling of the cladding tubes : prevents and distorts the adequate coolant flow generates contact stress at interaction with fuel assembly structures (e.g. grids). In a first approximation a swelling limit of 6% is allowed 9 Cr F/M steel is the best one, nevertheless also 15/15 Ti has acceptable swelling at 150 dpa

6 Swelling: Comparison of proven materials Austenitic steels are proven materials by FR technology The swelling performance dominates the qualification CW 15-15Ti Si enriched highlights good swelling performance  demonstrated at 160 dpa with possibility to reach 200dpa Swelling of Ferritic-Martensitic steels  the evolution of swelling with dose is slow  The swelling rates are much smaller than those for austenitic steels

7 Advanced austenitic steels for low swelling 7

8 Thermal Creep resistance of austenitic vs. ferritic/martensitic steels 8 Comparison of creep resistance at 600°C between austenitic and ferritic/martensitic steels The creep resistance is an imporant parameter for cladding material selection. For ELSY a hoop stress of 160 MPa is envisaged. In such conditions, if the cladding temperature unexpectedly rises up to 600 °C, the rupture time becomes very short. The thermal creep resistance of T91 at 600°appears too poor. Nevertheless reliable creep data of 15/15 Ti have to be recovered and re-measured.

9 9 Austenitic steel  The creep vary close to linear with respect to the applied load  The creep is proportional to the irradiation dose  The creep proportionality to the dose is valid only in the domain of the swelling incubation period  The creep performance is not largely dependent from alloying elements 40 dpa Comparison with Ferritic-Martensitic steel For high temperature or high stresses,  the creep do not vary linearly with respect to the applied load  The thermal creep greatly contributes to dimensional changes  Where the creep is proportional to the irradiation dose, the creep/swelling correlation is similar to that for austenitic  At 520°C the creep behavior is acceptable, at 590 °C is no more acceptable. Irradiation Creep: Comparison of proven materials

10 Creep rupture of F/M steels in HLM 10 Creep to rupture tests of T91, 10 -6 wt% oxygen performed at Prometey St. Petersburg – V. Markov A. Jianu, G. Mueller, A.Weisenburger Significant reduction of creep strength of T91 in contact with liquid LBE. This experiment shows the necessity to protect the cladding steel by a compliant layer different from the oxides layer LBE 160 MPa 3107h Ø ~ 2.5mm In LBE cracks in and through oxides scale The lower the stress the larger the cracks

11  Tertiary stage switched on by the threshold strain  th  Threshold strain appears time dependent, decreasing during thermal exposure due to the precipitation and coarsening of Laves phases  Damage strongly depends from accumulated strain Modelling of Tertiary Creep of F/M steels

12 Correlation between the behavior of the threshold strain and the evolution of Laves dimension. evolution of  th is proportional with the inverse of evolution of mean Laves radius during ageing voids formation close to Laves The threshold strain for tertiary creep of F/M is associated with Laves phase and voids formation P91 micrographic analysis Normalized  th function Normalized mean Laves radius Microstructure observations Voids formation close to Laves nucleation Fe2(Mo,W) C.Testani “MATTER workshop 2012”

13 Fatigue resistance of F/M steel 13 Several tests of thermal fatigue were performed on Eurofer 97 by ENEA in the frame of the Fusion Programs.The studies are reported in: G. Filacchioni, The Thermo-Mechanical Fatigue Testing Facility of Casaccia’s Laboratories, MAT TEC, March 2002 The softening effect of strain controlled fatigue is evident after few cycles. Eurofer chemical composition is 9Cr and 1 W instead of 9Cr and 1Mo as T91 Eurofer (low activation Ferritic /martensitic) 316 L steel for fatigue comparison

14 14 Embrittlement of Ferritic-Martensitic steel  DBTT value for T91 and EM10 after irradiation remains below room temperature  Martensitic steels behave better than ferritic steels Irradiation Embrittlement: comparison of proven materials  In CW steels hardening at irradiation temperatures <450°C and ductility increase at higher irradiation temperatures is observed.  Loss of ductility is observed at higher irradiation conditions  It has been proved that the enhancements that lead to higher swelling resistance also have beneficial effects on mechanical properties

15 15 Results by PSI for T91 based on Total elongation Results by PROMETEY Institute for 10Ch9NSMFB based on % necking to rupture Ferritic martensitic steels present Liquid Metal Embrittlement in the temperature range 300 – 420 °C when exposed to HLM. Similar results where obtained by PSI for T91 and by Prometey Institute for notched 10Ch9NSMFB steel (9.4 Cr, 1.3 Si, 0.84 Ni) HLM Embrittlement of grade 91 steel Necking in air Necking in Pb

16 Liquid Metal Embrittlement comparison 16 LME observed in T91 under specific conditions and after UTS Tests performed in LBE at 350° 5×10 -5 s -1 No LME observed in 316L Tests performed in LBE at 350° 5×10 -5 s -1 Observations by SCK-CEN

17 WELDING ISSUES OF Grade91 17 CEA experiments to account for the reduced fatigue resistance of welded P91 At low cycles the type IV cracks were observed At high cycles the cracks in the base metal were oserved HAZ BMWM BM HAZ WM The determination of the welding coefficient for P91 deserves additional efforts. The filler metal, the welding method and the post weld heat treatment are under study.

18 HLM corrosion The HLM presents high solubility of the chemical elements of structural steels: Fe, Cr and mainly Ni In both austenitic and ferritic martensitic steels, a partial protection vs. dissolution is achieved by formation of protective oxides Nevertheless at 550 C and 10 -6 wt% O2 (high oxygen) the dissolution is not completely prevented As shown, the protective oxides are ruptured under stress Moreover the picture shows that for T91 in lead at 500°C, the oxide layer looses its adherence to the matrix and is fractured and removed by the Pb flow. 18 T 91 AISI 316

19 HLM CORROSION 19 316 @ 500°C, O2 10 -6 wt% 10000h stagnant PbBi 316 @500°C, O2 10 -6 wt% 10000h Flowing Pb (ENEA) Temperature limits for corrosion (dissolution) of steels in Pb/PbBi 316 type steels: T limit < 450°might be 500°in Pb – to be assured T91 type F/M steels T limit < 550 °C The oxide scale of austenitic steel is thinner and more stable than that of T91. The additional material protection appears to be necessary to face the corrosion by flowing lead. The suitable coating must be: Resistant to neutron irradiation Resistant to mechanical stress Thin to reduce risk of rupture (about 40 microns)

20 Comparison of materials for ALFRED cladding Swelling performance of grade 91 is better than that of austenitic steels: advanced austenitic have to be developed Thermal creep resistance of grade 91 is poor and Irradiation creep is not linear with load Grade 91 is subject to fatigue softening Cyclic strength of Grade 91 is 50% lower than that of 15-15 Ti Irradiation embrittlement for both 15-15 Ti and Gr.91 is acceptable Gr.91 is subject to HLM embrittlement at T< 420 C. Gr.91 welds are subject to type IV rupture and require special heat treatment Both 15-15Ti and Gr.91 are subject to HLM corrosion (elemental dissolution). The only oxides scale is not an effective corrosion barrier: ruptured under stress, spalled at higher temperatures 20 Austenitic Ferritic/Martensitic

21 Considerations on materials for ALFRED cladding It is confirmed that the fuel cladding of the first core of ALFRED will not be made of Gr.91 steel, since its mechanical properties (creep, fatigue, HLM embrittlement, welds) appear too poor and subject to ageing. An intensive R&D is being addressed in France for austenitic steels resistant to irradiation swelling. ENEA also is very much interested to this research line It is confirmed that the weak point of LFR technology is represented by the dissolution of main steel elements. The naturally formed oxides scale, although mitigating the dissolution effect, cannot represent an effective protection for long time in stressed condition and high temperature. In short term, the reference material for ALFRED fuel cladding is 15-15 Ti, Si stabilized, protected by a well qualified corrosion barrier. The potential candidates for corrosion barriers include : Fe-Al, TiN (BLUE), Al oxide, GESA, Ta and possibly others. In the long term, corrosion resistant austenitic steels have to be selected and qualified for fuel cladding: Si or Al containing steels 21

22 Exposed for 2000h in Pb Coatings under test: T91 “BLUE” coated No apparent damages on the layer No lead penetrations are observed Exposed for 4000h in Pb

23 Coatings under test: T91 “SS39L” coated 5000 hours of exposure for SS39L, the last CHEOPEIII run. The coating appears heavily damaged, with random thickness Oxygen inner precipitation.

24 Coatings under test: T91 “FeAl” coated 5000 hours of exposure of FeAl, the last CHEOPEIII run. The coating appears untouched where its original quality is good, locally damaged with Oxygen precipitation where detachments are present. No changes in chemical composition Inner Oxygen precipitation in conjuction with defects, near the limit of the coated area Pefect result

25 Coating under test: AISI 316 Ta coated 25 Ta coating Successfully tested as bulk material in PbBi. Successfully tested with plastic deformation in room conditions. Not yet tested in creep-rupture tests. The use in the core has to be clarified due to high neutron capture and transmutation to W 1µm

26 Further steps Extensive testing campaign of steel corrosion barriers in controlled corrosion conditions Extensive testing campaign of steel corrosion barriers in HLM under stress and strain conditions PIE after irradiation tests performed in BOR 60 at 16 dpa Development of additional corrosion barriers for austenitic and F/M steels Qualification of corrosion resistant steels for cladding Collaborations to get irradiation data on advanced austenitic steels 26 1µm 253MA Average thickness < 1 µm  253 MA (21wt% Cr, 11wt% Ni, 2wt% Si)

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