**Oxidation, Embrittlement, and Growth of TREAT Zircaloy-3 Cladding**

Charles W. Solbrig, Anthony LaPorta, Katelyn M. Wachs and James R. Parry

Additional information is available at the end of the chapter

http://dx.doi.org/10.5772/62708

#### **Abstract**

This chapter analyzes the effects of oxidation, embrittlement, and cladding growth on the Zircaloy-3 alloy used for 25 mil thick TREAT fuel assembly cladding. The fuel cladding is a protective shell which is used to prevent damage to the enclosed fuel. Therefore, its integrity is important to guarantee this protection. The above three factors which can affect the Zircaloy-3 cladding are considered in this chapter and investigated. Limits to operation are determined. The oxidation of Zircaloy-3 in air is of interest to air-cooled reactors and Zircaloy-2 and 4 for accidents in fuel storage pools. The temperature range of interest is from room temperature where the fuel is stored for long periods of time, through the temperature range encountered in normal operation (400 to 600°C) to the highest temperatures which are possible in extreme accident situations. This tempera‐ ture range is considered in this chapter to be from room temperature to 1200°C.

**Keywords:** cladding, zircaloy, oxidation, embrittlement, metal growth

## **1. Introduction**

This chapter describes the corrosion rate of Zirconium-2, 3, and 4 in air. Zircaloy-2 and 4 are used to clad fuel for all commercial power reactors. The Kendall Zircaloy-2 oxidation correla‐ tion is derived based on the well-known physically justified Arrhenius equation using legacy rate data at 500, 600, and 700°C. Additional data obtained by several other authors have shown that this correlation can be extended so that it adequately covers the range of 200 to 1100°C. This correlation also bounds the oxidation rate of Zircaloy-3, which is used to clad the fuel in the TREAT reactor. Zircaloy-4 rates are seen to be higher than the Zircaloy-2 correlation. Zirca‐

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loy-2ratesarehigherthantheZircaloy-3rates.TheZircaloy-2correlationisverifiedforZircaloy-3 by comparison to oxide thickness measurement on Zircaloy-3 coupons oxidized in air for different time periods in the temperature range of 500 to 1100°C: many coupons proceeded to disintegration. Although the latter samples still had a metal layer, oxygen incursion into the metal grain boundaries caused the metal to become brittle and crack like rust. The Kendall correlation is shown to adequately describe the behavior of these samples over the entire range 500 to 1100°C. Maximum oxide limit of 15.52 mils growth for TREAT cladding was deter‐ mined from these data to ensure that the fuel assemblies could still be removed without disintegration.

The TREAT reactor, a graphite-moderated thermal reactor, is designed primarily for operation in the transient or pulsed mode for destructive testing of prototypic fast reactor fuel pins. It is designed conservatively to produce a pulse with a thermal neutron fluence of at least 3.5 × 1015 neutrons/cm2 averaged over the core. The operating TREAT core temperature limit at the peak is 600°C.

The standard TREAT fuel assembly consists of upper and lower graphite reflector sections and a central section of uranium oxide-bearing graphite fuel. The fuel section is 4 ft long and contains six fuel blocks, each 8 in. long and 3.96 in. square with chamfered corners. The reactor fuel blocks consist of small particles (mean size, 10 microns) of fully enriched 235U dispersed in a graphite-carbon matrix. The carbon-to-uranium U235 atom ratio is nominally 10,000:1. The graphite-carbon-urania blocks are sealed within evacuated Zircaloy-3 cans.

Zirconium alloys are used for cladding in all commercial power thermal reactors because of the high corrosion resistance, low cross section for thermal neutrons, and high temperature capability. As shown later in this chapter, Zircaloy-3 (Zr-3) is more corrosion resistant in air than either Zr-2 or Zr-4 and was used to clad the TREAT fuel. In low-temperature reactors such as TRIGAs, aluminum can be used. Since TREAT has a very thermalized spectrum, the low cross section and high temperature capability were the reasons Zr-3 was used.

The compositions of the common zirconium alloys are listed in **Table 1** (Gibbons [1], Blanchard [2] for Zr-2, Zr-4, and Alloy Digest [3] for Zr-3). Zr-3 has much less zinc than either Zr-2 or Zr-4.


**Table 1.** Composition percentages of commercial Zirconium alloys (w/o).

A considerable amount of research has been carried out in investigating the oxidation of zirconium alloys used in different environments. The early research was carried out on Zircaloy-2 and the later work was carried out on Zircaloy-4. The recent oxidation research has been carried out because of concern about loss of cooling water from spent fuel pools where temperatures range from 300 to 600°C. Natesan [4] in 2004 conducted air oxidation tests on unirradiated Zircaloy-4 cladding starting with a 25 to 30 μm (1 mil) oxide layer representative of the current inventory of spent fuel discharged after a medium or high level of fuel burnup. Temperatures were in the range of 300–600°C, which is representative of cladding heat up in the event of a partial or full draining of spent fuel pool coolant. Ji Min Lee [5] in 2012 investi‐ gated the oxidation of Zircaloy-4 under transient conditions from 500 to 800°C. Duriez [6] in 2009 summarized results of several studies comparing them in terms of kinetics and oxide scale structure and composition. Steinbrück [7] in 2009 studied the mechanism of the reaction between Zircaloy-4 and air at temperatures from 800 to 1500°C. Both of these studies consid‐ ered pre-oxidized metal. Beuzet [8] in 2009 used existing correlations to simulate the zirconia scale growth under air atmosphere in the MAAP4.07 Severe Accident code. Duriez [9] in 2008 studied the degradation of Zircaloy-4 and M5 cladding tubes in air at high temperature by thermo-gravimetric analysis, in isothermal conditions, from 600 to 1200°C. Steinbrück [10] in 2007 conducted experiments on the reaction between Zircaloy-4 and air under mixed air (nitrogen) steam atmospheres and pre-oxidation conditions for severe nuclear reactor accident temperatures 800–1500°C. The earlier Zircaloy-2 and 3 work is described in the next section.
