Corrosion of Zirconium alloy componentsin light water reactors

By

Udit Kumar (163110065)

Overview

Introduction Light water reactors Environment and why zirconium alloys as fuel rod cladding Zirconium Alloys Corrosion of Zirconium Alloys in Water without Irradiation Heat Flow Conditions & Impact of Metallurgical Parameters on Oxidation

Resistance Hydrogen Pickup and Hydriding Effect of Irradiation on Microstructure and Corrosion Corrosion of Fuel Rods in Reactors

Introduction

Presentation highlights the various uses and properties of zirconium alloy cladding and structural components used in nuclear power light water reactors. Specific attributes including dimensional stability, corrosion resistance, irradiation effects and mechanical properties.

Zirconium alloys are used as the prime structural material in light water reactors (LWRs). As such, they have to meet several requirements: low neutron absorption cross section; corrosion resistance in 280–350 ° C water; resistance to radiation in both mechanical behaviors and dimensional stability; reasonable strength, ductility and fabricability; affordable cost; and availability in large quantities

Light water reactors

The light-water reactor (LWR) is a type of thermal-neutron reactor that uses normal water, as opposed to heavy water, as both its coolant and neutron moderator.

Pressurized water reactor (PWR) fuel consists of cylindrical rods put into bundles. A uranium oxide ceramic is formed into pellets and inserted into Zircaloy tubes that are bundled together

Fig :1 Schematic Boiling water reactor

Environment

Highly aggressive Environment Pressurized water environment High temperature Irradiation

Why Zirconium alloy ? Less neutron absorbing material Very stable oxide Layer

Characteristics of water reactor fuel assemblies

Reactor Type

Dimension(dia.mm*thickness mm*length m)

Alloy Temperature 0C

PressureMPa

PWR 9.5*0.6*3.7 Zry4 290-335 15.5

BMR 12*0.8*4 Zry2 275-285 7

CANDU 13.8*0.41*1.6 Zry4 265-310 10

Stephen D. Cramer and Bernard S. Covino, J. (n.d.). ASM Handbook Volume 13C: Corrosion: Environments and Industries. ASM International, Ch. 46

Zirconium Alloys

Two Major Families Zr-Sn-O (+Fe ,Cr ,Ni) Ze-Nb-O ( with 1 nad 2.5 wt % Nb)

Highly reactive with oxygen and this oxide is very stable (much like Al2O3 ). Alloying Addition

Tin (Sn) – Improves mechanical strength, creep resistance Niobium (Nb) – very effective in corrosion resistance Oxygen – around 1250ppm as a Hardening agent

Grade UNS No.

Sn Fe Cr Ni Nb

Zircaloy 2

R60802

1.2–1.7 0.07–0.2

0.05–0.15

0.03–.08

-

Zircaloy 4

R60804

1.2–1.7 0.18–.24

0.07–.13

- -

Zr-1Nb - - - - - 1Zr-2.5Nb 60901 - - - - 2.5

Stephen D. Cramer and Bernard S. Covino, J. (n.d.). ASM Handbook Volume 13C: Corrosion: Environments and Industries. ASM International, Ch. 46

Zirconium Alloys

BLOCK, S., DA JORNADA, J. A. H., & PIERMARINI, G. J. (1985). Pressure-Temperature Phase Diagram of Zirconia. Journal of the American Ceramic Society, 68(9), 497–499. http://doi.org/10.1111/j.1151-2916.1985.tb15817.x

At atmospheric pressure

Fig :3 Pressure Temperature phase diagram of Zirconia

Corrosion of Zirconium Alloys in Water without Irradiation• The behavior of zirconium in contact with

high-temperature water leads to an oxidation reaction• Zr + H2O ZrO2 + 4H

• The growth of the oxide layer has been shown to occur by diffusion of O2- ions through the oxide and formation of new oxide layer at the metal/oxide interface.

• This protective behavior of the zirconia layer is, however, limited by a transition occurring at a thickness of a few micrometers.

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Fig: 4 General oxidation kinetics of Zircaloy in high-temperature water (1 mmZrO2 ~15 mg/dm2)

B.C. Cheng, R. Kruger, and R. Adamson, Corrosion Behavior of Irradiated Zircaloy, Zirconium in the Nuclear Industry, 10th Intl. Symp., STP 1245, ASTM, 1994, p 400–418

Corrosion of Zirconium Alloys

The crystallographic form of the zirconia at standard temperature and pressure is the monoclinic structure. At high pressure, either the cubic or the tetragonal phase is observed, depending on temperature

As the oxide layer grows, these compressive stresses enhance the stability of the tetragonal phase.

In an ionic binding material, such as zirconia, this process induces internal cracking. Other

The zirconia to transform to the monoclinic structure. In an ionic binding material, such as zirconia, this process induces internal cracking. Other

The oxide consists of an internal layer, rich in the tetragonal phase and considered to be dense and protective, above which the outer layer is mostly monoclinic and porous.

Hydrogen Pickup and Hydriding

The major part of the atomic hydrogen released in the pores of the granular oxide, by the reduction of the water molecules, recombine and form H2 molecules.

Hydrogen form solid soln with zirconium at high temperature and precipitate as ZrH1.6, and other hydrides.

The solubility of hydrogen in zirconium alloys is strongly temperature dependent, with a heat of solution of 64 kJ/mol. At 320 C (610 F) the solubility of hydrogen is approximately 90 to 110 ppm, while at room temperature it is below 1 ppm

Fig: 5 Temperature profile through the oxide layer in heat flow condition. Temperature decreasing from zirconium alloy to coolant

Stephen D. Cramer and Bernard S. Covino, J. (n.d.). ASM Handbook Volume 13C: Corrosion: Environments and Industries. ASM International, Ch. 46

Oxide Morphology

Fig: 6 Nodular corrosion in Zircaloy 2 fuel cladding. (a)Unaided view.(b) Scanning electron-micrograph enlargement

Stephen D. Cramer and Bernard S. Covino, J. (n.d.). ASM Handbook Volume 13C: Corrosion: Environments and Industries. ASM International, Ch. 46

Effect of Irradiation on Microstructure and Corrosion

The irradiation damage, induced by the fast neutrons knocking the metal atoms and inducing displacement cascades and point defects, leads to changes in microstructure

For typical LWR conditions, this is the physical origin of an irradiation induced crystalline-to-amorphous phase transformation observed within these phases.

This irradiation-induced micro- structure change has been observed to severely affect the corrosion resistance after irradiation

D. Gilbon and C. Simonot, Effect of Irradiation on the Microstructure of Zircaloy-4, Zirconium in the Nuclear Industry, 10th Intl Symp., STP 1245, ASTM, 1994, p 521–548

Fig: 7 Corrosion behavior of several Zircaloy specimens in autoclave with and without irradiation.

Corrosion of Fuel Rods in Reactors

Oxide thickness generally increases along the fuel rod because the coolant tempera- ture increases as it is heated by the fuel

In the lower part of the core, the corrosion rate is low due to a temperature close to the inlet temperature, while the upper parts of the rods exhibit higher oxide thickness

J.C. Clayton and R.L. Fischer, Corrosion and Hydrating of Zry Fuel Rod Cladding in 633K Water and Reactor Environments, ANS Light Water Reactor Fuel Performance (Orlando, FL), April 21–24, 1985, American Nuclear Society, p 3–1 to 3–16

Fig: 8 Oxide thickness along two PWR fuel rods, R1 and R2, of the same assembly

Conclusions

MOST OF THE POWER REACTORS currently in operation are using natural (light) water as the coolant and moderator and are referred to as light water reactors (LWRs)

Fuel cladding works in a aggressive environment and naturally it suffers heavy corrosion problem.

Corrosion problem proceed in two stages, 1st one is formation of zirconia on the surface and the phase transformation of zirconia with temperature and stress condition. Which leads to our 2nd stage of damage when hydrogen atom diffuses into bulk of zircaloy and form hydrides which leads to hydrogen embrittlement.

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