Alloying Effect of Nb and Sn on the Zirconium Alloy

Fuel Claddings Behavior at High Temperature

Oxidation in Steam

A. Malgin, V. Markelov, A. Gusev, A. Nikulina, V. Novikov, I. Shelepov

(SC “VNIINM”, Moscow, Russia)

V. Donnikov, V.Latunin, J. Kosihina

(JSC “VTI”, Moscow, Russia)

«ROSATOM» STATE ATOMIC ENERGY CORPORATION

A.A. BOCHVAR HIGH-TECHNOLOGY RESEARCH INSTITUTE

OF INORGANIC MATERIALS (SC «VNIINM»)

18TH INTERNATIONAL SYMPOSIUM ON ZIRCONIUM IN THE

NUCLEAR INDUSTRY

MAY 15-19, 2016

Introduction

2

The current development trend of nuclear energy is to increase the fuel

efficiency determined by the level of the reached burnup and fuel cycle

versatility. In this regard increased requirements to the functional properties of

the fuel cladding material are needed to ensure the operating safety of reactors

PWR and WWER under normal operating conditions and in emergency

situations. In addition to the alloy Zircaloy-4 (PWR), the zirconium alloys systems

Zr-Nb (E110, М5, J-alloys) и Zr-Nb-Sn-Fe (E635, ZIRLO, AXIOM, Q-alloys as well as

others) are used in these reactors as the fuel cladding material. Operational

alloys are improved and new zirconium alloys are developed at the same time.

The purpose of the present work was to study the influence of Nb, Sn

and (Fe) on corrosion and loss of ductility under conditions of high temperature

steam oxidation of cladding tubes made of binary Zr-Nb (Nb=1.0÷2.5 wt.%)

alloys and alloys of the Zr-Nb-Sn-Fe system with different content of

Nb (0.60÷2.37 wt.%), Sn (0.38÷1.08 wt.%) and Fe (0.18÷0.34 wt.%).

Investigated Materials

3

ALLOY Content, wt.%

Nb Sn Fe O

Zr-1.0Nb 1.0 - 0.03 0.06

Zr-1.7Nb 1.7 - 0.04 0.06

Zr-2.5Nb 2.5 - 0.04 0.06

O 1 0.60 0.38 0.33 0.11

О 2 1.00 0.41 0.33 0.09

О 3 1.64 0.40 0.33 0.09

О 4 2.33 0.40 0.32 0.09

О 5 1.03 0.59 0.34 0.09

О 6 1.02 0.58 0.18 0.10

О 7 2.37 0.41 0.23 0.09

О 8 1.00 1.08 0.33 0.09

O 9 1.00 0.24 0.32 0.10

Investigations were carried

out on samples of cladding tubes

sizes Ø9.10×7.73(7.93) mm

produced in industrial conditions

of the experimental ingots

(weight ~ 50 kg) of various alloys

on zirconium sponge base.

As-received cladding tubes

made of Zr-xNb alloys had etched

outer and inner surfaces and

tubes from O1÷О9 alloys had

polished outer and etched inner

surfaces.

All heat treatments of tubes at

the cold work stage were carried

out for all alloys within the

temperature range of α-Zr with

final annealing at 580°C-3 h.

I

II

III

IV

High Temperature Steam Oxidation Tests

4

cooling in steam

Tem

pe

ratu

re,

°C

Oxidation Time, s U-127 facility scheme

Testing conditions:

● Double side oxidation;

● Steam environment; ● Atmospheric pressure;

● Temperature of oxidation: 1000-1200°С;

● Steam flow rate: 20-90 g/h (~ 1.5-5.5 mg/cm2/s);

● The error of temperature keeping is ~ ± 1°С;

● The instrumental error of a mass measurement is ±0.1 mg;

● The heating rate ~ 50°C/s; ● The cooling rate in steam ~ 20°C/s;

● Short-term overheating of the sample on the heating stage to the

nominal temperature not exceeding 15°C;

● Weighing the sample before, during, and after the experience on an

analytical balance with an accuracy of ±0.1 mg; ● Sample length: 30 mm.

sample

balance

furnace

water

thermocouple

steam generator

quartz tube

Pt suspension

Investigations of Oxidized Samples

5

Oxidation kinetics

Microstructure analysis

Microhardness meas.

(DUROSCAN 70)

Ring Compression Tests

(INSTRON 8861) Hydrogen analysis

(LECO TCH600)

Oxidized Sample

Oxidation kinetics at 1000°С

6

Nb=0.6%

Nb=1.0%

Nb=2.3%

Sn=1.08%

Sn=0.59%

Sn=0.41%

Sn=0.24%

Zr-хNb

Zr-1.0Nb-ySn-zFe

Zr-xNb-0.4Sn-zFe Appearance of the

oxidized samples

“Breakaway” effect

7

- acceleration of oxidation (oxygen uptake)

- formation of cracked oxide films

- absorption of hydrogen released during

interaction with steam

Enhancing of cladding embrittlement

“breakaway”

The transition from the parabolic

oxidation according to the linear

– “breakaway” effect

Zr + 2H2O → ZrO2 + 2H2

“Breakaway” effect at 1000 °С

8

For Zr-Nb-Sn-Fe alloy system the increasing Sn content leads to a decrease in

duration until the development of “breakaway” while the increase in the Nb content

of over 1.6 wt.% increases the duration to its development. Significant influence of

the Fe content was not revealed.

Breakaway oxidation time, s

Hydrogen content in samples after oxidation at 1000°C

9

At the same duration of oxidation the hydrogen

content in the oxidized samples can vary by

several orders of magnitude.

Hydrogen content, ppm

Oxidation time - 3350 s

Microstructure of oxidized samples (1000°C / 5000 s)

10

Zr-1.0%Nb

Zr-2.5%Nb

+ 1.5%Nb + 1.7%Nb + 0.8%Sn

O1 (Zr-0.6Nb-0.38Sn-0.33Fe)

O4 (Zr-2.33Nb-0.40Sn-0.32Fe)

O9 (Zr-1.0Nb-0.24Sn-0.32Fe)

O8 (Zr-1.0Nb-1.08Sn-0.33Fe)

ZrO2

αZr(O)

“ex-β”

ZrO2

αZr(O)

“ex-β”

ZrO2

αZr(O)

“ex-β”

ZrO2

αZr(O)

“ex-β”

ZrO2

αZr(O)

“ex-β”

ZrO2

αZr(O)

“ex-β”

α-incursions

Distributions of alloying elements within wall thickness of oxidized

samples (Tox=1000 °С)

11

ZrO2

αZr(O)

“ex-β” ZrO2

αZr(O)

“ex-β” ZrO2

αZr(O)

“ex-β”

OK

FeK

ZrK

NbK

OK

FeK

ZrK

NbK

αZr(O)

OK

SnL

ZrK

NbK

FeK

Zr-1.0%Nb Zr-2.5%Nb O8 (Zr-1.0Nb-1.08Sn-0.33Fe)

Under the action of oxygen diffusion into the metal there was a redistribution

of Nb and Fe. There is no essential redistribution of Sn in the O8 alloy, the level of Sn

concentration in the -Zr(O) and “ex-” layers was the same.

Oxidation kinetics at 1100°С and 1200 °С

12

Zr-хNb

Zr-xNb-0.4Sn-zFe

As the temperature of oxidation increased, the differences in

the oxidation kinetics of alloys of Zr-Nb and Zr-Nb-Sn-Fe

systems decreased. "Breakaway" oxidation was not observed.

Appearance of the oxidized samples

The hydrogen content

in oxidized samples

(1100°C/1000s) of all

alloys was not higher

than 60 wppm.

Zr-1.0Nb-ySn-zFe

Microstructure of oxidized samples (1100 °С/1200 s & 1200°С /300 s)

13

O4 (Zr-2.33Nb-0.40Sn-0.32Fe)

+ 1.5%Nb + 1.7%Nb + 0.8%Sn

Zr-2.5%Nb (2000s/500s) O8 (Zr-1.0Nb-1.08Sn-0.33Fe)

1100°С 1200°С Zr-1.0%Nb

(2000s/500s)

ZrO2

αZr(O)

“ex-β”

ZrO2

αZr(O)

“ex-β”

1100°С 1200°С 1100°С 1200°С

O1 (Zr-0.6Nb-0.38Sn-0.33Fe) O9 (Zr-1.0Nb-0.24Sn-0.32Fe)

α-incursions

Microstructure studies

14

Average oxide layer thickness

Volume fraction of “ex-β” phase (Δexβ) ∆𝑒𝑥𝛽 =𝑉"𝑒𝑥𝛽"

𝑉𝑡𝑜𝑡

= 𝑆"𝑒𝑥𝛽"

𝑆𝑡𝑜𝑡

Zr-хNb Zr-xNb-0.4Sn-zFe Zr-1.0Nb-ySn-zFe

Microhardness of oxidized samples

15

Zr-xNb-0.4Sn-zFe Zr-1.0Nb-ySn-zFe

Zr-хNb

“ex-β” layer

αZr(O)-outer αZr(O)-inner

Nb

Nb

α-incursions

Ring Compression Tests (20°C)

16

Zr-xNb-0.4Sn-zFe Zr-1.0Nb-ySn-zFe

Zr-хNb Typical deformation diagrams

Conclusions - I

17

For alloys of the Zr-Nb system the increase of the Nb content has no

significant influence on the kinetics of oxidation but effects the formation of a

more dispersed structure of α-Zr(O) and “ex-β” layers and makes the oxygen

solubility enhance in the βZr phase at a high temperature. In this case the residual

ductility of oxidized tube samples decreases with increasing Nb content in the

alloy.

The most resistant to high temperature oxidation and embrittlement in the

temperature range 1000-1200°C is the alloy Zr-1.0Nb.

Conclusion - II

18

In alloys of the Zr-Nb-Sn-Fe system increase in the Nb content also

influences the structure formation of α-Zr(O) and “ex-β” layers and leads to a

decrease of the residual ductility. The increase in Nb content also reduces the

intensity of the breakaway at 1000°C.

The increase of Sn in the alloy accelerates the formation of "breakaway"

oxidation, which ultimately leads to a significant reduction in residual ductility due

to violation of protective properties of an oxide film and a more intensive

absorption of oxygen and hydrogen by the metal.

The decrease of the content of Fe in the alloy has no influence on the

kinetics of high temperature steam oxidation, but leads to an increase in residual

ductility.

To ensure the corrosion resistance and the residual ductility after the

oxidation in steam at 1000-1200°C for alloys of the Zr-Nb-Sn-Fe system the content

of Sn and Fe should be reduced while maintaining the content of Nb at the level of

1 wt.%.

19

THANK YOU FOR

YOUR ATTENTION

High Temperature Steam Oxidation of Zr Alloys

20

Investigations of zirconium fuel claddings under

LOCA simulating conditions have demonstrated that as a

result of interaction with high temperature steam and

subsequent sharp cooling the following processes occur

in zirconium alloys :

1) oxidation to form surface oxide ZrO2 layer and

oxygen deep diffusion into metal;

2) formation of a novel microstructure

corresponding to temperature–phase conditions as a

result of transformation;

3) diffusion induced redistribution of alloying

elements and possibly impurities within cladding

thickness (ZrO2-Zr(O)”ex-”);

4) absorption of hydrogen released during

interaction with steam.

These processes leads to a significant decrease of ductility and

embrittlement of the zirconium cladding.

Microstructure investigations

21

∆𝑒𝑥𝛽 =𝑉"𝑒𝑥𝛽"

𝑉𝑡𝑜𝑡

= 𝑆"𝑒𝑥𝛽"𝑆𝑡𝑜𝑡

Typical microstructure of tube sample from zirconium alloy after high

temperature steam oxidation and subsequent cooling (O5 alloy/ 1000°C/ 5000 s)

Increasing / reduction of «ex-β» grain - 1100°C

22

Zr-1.0%Nb (2000 s) Zr-2.5%Nb (2000 s)

O9 (Zr-1.0Nb-0.24Sn-0.32Fe) O8 (Zr-1.0Nb-1.08Sn-0.33Fe)

+ 1.5%Nb

+ 0.8%Sn

Influence of Hydrogen content on Oxygen solubility in βZr

23

M. Négyesi, J. Burda , O. Bláhová , S. Linhart , V. Vrtílková, The influence of hydrogen on oxygen distribution inside Zry-4 fuel cladding, Journal of Nuclear

Materials 416 (2011) 288–292

M. Négyesi et. al. Contribution to the study of the pseudobinary Zr1Nb–Oxygen phase diagram by local oxygen measurements of Zr1Nb fuel cladding after high

temperature oxidation. Journal of Nuclear Materials 420 (2012) 314–319