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

© 2015 Electric Power Research Institute, Inc. All rights reserved.

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

2© 2015 Electric Power Research Institute, Inc. All rights reserved.

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


– 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


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


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



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



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


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


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


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)


Direct Impact of Hydrogen on Ductility

Hydrogen by itself reduces ductility

• Is hydrogen• Is hydrogen and oxygen effects additive oradditive or interactive?


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


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


elements do not form a short-cut for oxygen diffusion


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)


Pre-hydriding is a good surrogate for irradiation


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


Together…Shaping the Future of Electricityg p g y


technical basis for new alloy definition from pqd testing ...technical basis for new alloy dfiiti f...

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