Technical Basis for New Alloy Definition from PQD Testing ...Technical Basis for New Alloy Dfiiti f...
Transcript of Technical Basis for New Alloy Definition from PQD Testing ...Technical Basis for New Alloy Dfiiti f...
Technical Basis for New Alloy D fi iti f PQD T tiDefinition from PQD Testing
Perspective
Ken YuehPrincipal Technical Leader
O k Rid N ti l L b tOak Ridge National LaboratoryApril 29-30, 2015
US NRC 50.46c Draft Regulatory GuidesPublic Meeting
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Public Meeting
Contents
Zirconium Based Alloy Metallurgy
Changes to Cladding During IrradiationChanges to Cladding During Irradiation
Damage Recovery
Ductility Loss Mechanisms During a LOCA
Proposalp
Technical Basis For Proposal
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Zirconium Based Alloy Metallurgy
Commercial zirconium based alloys are > 97 wt% Zr
Alloys are typically used in the α phaseAlloys are typically used in the α-phase– Low alloying element solubility
Th t i l t f i t th β h t 850°CThe material transforms into the β-phase at ~850°C– High alloying element solubility, >=1.9% for all commonly used
alloying elements
Controlled thermal-mechanical processing typically starts with a β-phase quench (~1000°C) in the manufacturing process– Homogenizes the alloying/impurity elements in solution
Eliminates any prior thermal mechanical processing history
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– Eliminates any prior thermal-mechanical processing history
Zirconium-X Binary Phase Diagrams
Zr-Sn Zr-Cr Zr-Fe Zr-Ni Zr-Nb Zr-O
Solubility of alloying elements in the beta phase is high relative to alloying element
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Solubility of alloying elements in the beta phase is high relative to alloying element concentrations – No new phase forms within expected concentration range
Changes to Material During Irradiation
Irradiation damage– Point and dislocation defects generation
Fission fragments embedding near the inner diameter surface
Alloying element re-distribution
Alloying element transmutationAlloying element transmutation
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Changes to Material During a LOCA 1/3
Zircaloy-2 irradiation damage (point defects and dislocation) recovery
Non-Irradiated Irradiated
Irradiation damage recovery is completed below 700°C
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Changes to Material During a LOCA 2/3
Embedded fission fragments– Mostly embedded in the ID oxide, not relevant to the base cladding
materialmaterial– Existing test data do not show they form any “unacceptable phase”
with other alloying/impurity element detrimental to post LOCA d tilitductility
Alloy element re-distribution– No impact since homogenization takes place in β-phase above
~850°C
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Changes to Material During a LOCA 3/3
Transmutation of alloying elements is the only irradiation induced changes not erased during LOCA temperature exposureexposure
Transmuted elements could be accounted for in one of two waysways– Testing of irradiated cladding near end of life– Testing of the alloy with addition of expected transmutation g y p
elements
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Ductility Degradation During a LOCA
Generally accepted oxygen ingress into the beta phase leads to ductility degradation
Hydrogen accelerates degradation– ANL concluded hydrogen increases oxygen diffusion Evidence showing oxygen concentration is not enhanced with hydrogen
– Hydrogen by itself without oxygen diffusion can reduce cladding ductility
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Effect of Hydrogen on Oxygen Concentration
Wavelength Dispersive X-ray (WDX) analysis of ANL samples
(cou
nt) 10ppm H-5%ECR
636ppm H-5%ECR
(cou
nt) 10ppm H-5%ECR
636ppm H-5%ECR
• No significant increase in oxygen concentration from hydrogen
Con
tent
(C
onte
nt ( hydrogen
• Hydrogen contributes to localization of oxygen during cooling
Oxy
gen
CO
xyge
n C oxygen during cooling
0 0.2 0.4 0.6
Distance from OD (mm)0 0.2 0.4 0.6
Distance from OD (mm)
H drogen does not appear to significantl
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Hydrogen does not appear to significantlyenhance oxygen diffusion
Oxidation Temperature Effect
Significant difference in oxygen concentration from different oxidation temperatures
coun
t) 1000C-15%ECR1200C-15%ECR
coun
t) 1000C-15%ECR1200C-15%ECR
• Oxygen concentration difference due to higher oxidation
Con
tent
(C
onte
nt ( temperature is
significant
Oxy
gen
CO
xyge
n C
0 0.2 0.4 0.6Distance from OD (mm)
0 0.2 0.4 0.6Distance from OD (mm)
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Direct Impact of Hydrogen on Ductility
Hydrogen by itself reduces ductility
• Is hydrogen• Is hydrogen and oxygen effects additive oradditive or interactive?
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Hemann, A. et al, “Ductility Degradation of Irradiated Fuel Cladding”, IAEA publication
Proposals
For alloy chemistries bounded by testing experience (e.g., combined licensed alloy composition envelop of alloys tested)
All li it d i d f i di t d h d h d– Allow limits derived from non-irradiated hydrogen pre-charged testing
New Alloy - chemistries outside of the experience database– Bulk of data to be generated using hydrogen pre-charge
V ifi ti t ti f i di t d l ddi th t d d f– Verification testing of irradiated cladding near the expected end of life hydrogen concentration or use as-fabricated materials with transmuted species
– Allow operation within the hydrogen pre-charged test data range
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Rationale for Proposal 1/3
Ductility degradation at high temperatures is due to oxygen diffusion into the beta phase
O diff i i t ll d b th i i t i b lk t– Oxygen diffusion is controlled by the zirconium matrix bulk property– Oxygen diffusion could be modified if secondary phases form at
LOCA temperature and have higher oxygen diffusion rates– Phase diagram of commonly used alloying element showed high
solubility in the beta phase and therefore would not form secondary phasesp
Testing of existing alloys manufactured by different processes showed no significant difference in ductility p g ydegradation– Verifying metallurgical data that the commonly used alloying
elements do not form a “short cut” for oxygen diffusion
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elements do not form a short-cut for oxygen diffusion
Rationale for Proposal 2/3
ANL Zircaloy-4, ZIRLO and M5 PQD ductile-to-brittle transition ECR
• Sample from multiple vendors fit on the same trend 20
25
%)
Zirc-4; 800°C Quench, Non-IrradiatedZirc-2; 800°C Quench, Non-IrradiatedZIRLO - 800°C Quench, Non-IrradiatedM5 - 800°C Quench, Non-IrradiatedM5 - 800°C Quench, IrradiatedZIRLO - 800°C Quench, IrradiatedZi 4 800°C Q h I di t d line
• Irradiated data also consistent
15
men
t CP-
ECR
(%
Zirc-4 - 800°C Quench, IrradiatedPol. Fit
a so co s s ewith hydrogen pre-charged
5
10
Embr
ittle
m
Old Zirc-4
00 100 200 300 400 500 600 700 800
Hydrogen Content (wppm)
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Pre-hydriding is a good surrogate for irradiation
Rationale for Proposal 3/3
ANL test data demonstrated transmuted species from currently used alloying elements do not lead oxygen diffusion short cutdiffusion short-cut
Therefore definition of new alloys from PQD testing perspective should be outside of the
combined testing experience database
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Together…Shaping the Future of Electricityg p g y
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