Introduction

Zirconium (Zr) alloys have been used internationally for over 6 decades to encapsulate ceramic uranium dioxide (UO2) pellets in fuel rods for light water reactors (LWR).1,2 The Zr alloy cladding isolates the ceramic fuel and the fission products from the coolant water. The cladding also provides a surface for the removal of the heat produced during the U235 nuclear fission reaction. Under normal operation conditions, the external surface of the cladding would be in contact with water in the order of 300°C. The temperature of the external surface of the cladding is expected to be < 400°C. Under normal operation conditions, the oxidation rate of the Zr base cladding in the hot water is acceptable for the residence time of the fuel rods in the reactor (< ~ 8 years). However, if the temperature of the Zr alloy increases above 450°C, the reaction rate with water increases rapidly via a highly exothermic reaction.

$$ {\text{Zr}}\left( {\text{s}} \right) + 2{\text{H}}_{2} {\text{O}}\left( {{\text{l}},{\text{v}}} \right) = = = {\text{ZrO}}_{2} \left( {\text{s}} \right) + 2{\text{H}}_{2} \left( {\text{g}} \right) + {\text{Heat}}\;{\text{of}}\;{\text{Reaction}} $$

For over a decade, the international nuclear materials community has been engaged in finding more robust materials for the fuel rods in LWRs. This includes materials with better corrosion resistance and better mechanical properties beyond design basis accident conditions.3 These materials include selections for the fuel itself and for the cladding of the fuel. The fuel contains uranium 235, which undergoes a fission reaction liberating abundant heat which migrates to the surface of the cladding, where it is removed by the coolant light water, finally generating steam (directly in BWRs or indirectly in PWRs). Because the concept of more robust fuel was born after the Fukushima Daiichi events of March 2011, the improved fuel was initially called accident tolerant fuel (ATF), but later the designation may have evolved to advanced technology fuels (ATF) to put more emphasis on materials performance and plant operation life extension and focusing less on accidents.

Figure 1 shows that the ATF materials proposed for LWR fuel rods can be separated into: (1) cladding materials; and (2) fuel materials. Within each category, the materials can be classified further into: (1) evolutionary or modified materials; and (2) revolutionary or advanced materials.4 The focus of the current manuscript is the idea of iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel and the advantages and disadvantages compared to Zr alloys. In the case of a loss of coolant accident (LOCA), the original desired traits of the ATF cladding were to: (1) have superior oxidation resistance in air and steam at temperatures > 1000°C; (2) generate less hydrogen when reacting with steam; and (3) maintain the geometry of the fuel bundle for longer times, avoiding cladding burst and fuel dispersion and allowing for cooling and retrieval.5 Another benefit of ATF cladding could be a higher resistance to debris fretting wear from the coolant side. The FeCrAl alloys include a range of compositional elements, possibly Cr content in the range 12–21 wt.%, Al content 4.5–6 wt.%, Mo 0–3 wt.% and other minor elements such as Y. In this work, the properties of two alloys will be addressed. These two alloys, generically called APMT and C26M, are also referred to as IronClad materials. The C26M alloy has a nominal composition of Fe + 12Cr + 6Al + 2Mo + Y2O3, which was originally made at Oak Ridge National Laboratory in the traditional manner of melting, forging, rolling, extruding, etc.6 C26M was later produced in the exact geometric requirement for fuel cladding in boiling water reactors (BWRs) using the powder metallurgy (PM) process.3 The other alloy is APMT with a nominal composition of Fe + 21Cr + 5Al + 3Mo + Y2O3. APMT was always manufactured through the powder metallurgy process followed by seamless tube production by hot extrusion and subsequent cold pilgering into final fuel cladding dimensions. Its composition was slightly adapted for fuel cladding application by removing the yttria from its formulation. This newer composition (without Y) was called APMT2 or FA-SMT or nuclear grade APMT. APMT2 was also manufactured to full geometrical requirements for BWR fuel cladding.7

Fig. 1
figure 1

Modified and advanced ATF materials.

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Table I shows side by side the desired qualities of the cladding and how the families of Zr-based or FeCrAl-based alloys may respond.

Table I Comparative Properties of Zr based and FeCrAl based cladding for ATF
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The purpose of this work is to compare the high temperature (> 300°C) properties of FeCrAl materials compared to Zr-based alloys. The areas of discussion will include (1) mechanical properties including creep behavior and (2) fretting wear resistance.

Superior Mechanical Properties and Creep Resistance of FeCrAl

Huang et al. reported excellent mechanical properties of PMC26M and APMT2 tubes at 20°C and 315°C compared to Zirc-2 material.3 The specimens for tensile testing were actual cladding tubes (203 mm long), which were fitted at each end with solid inserted plugs for the load frame to grab them without crushing the tubes’ ends. The wall thickness of the two FeCrAl tubes was 0.3 mm, and the wall thickness for the Zirc-2 tubes was 0.6 mm.

Figure 2 shows the yield stress of the three types of non-irradiated tube specimens revealing the higher strength of APMT2 compared to both PMC26M and Zirc-2.3 Related data such as ultimate tensile stress (UTS) and elongation to failure have been reported before.3,7 APMT2 is the strongest material due to its small grain size and the presence of nano-particles in the matrix. The presence of the nanosized precipitates in APMT2 contributes to its enhanced tensile and creep properties and may provide resistance to irradiation damage. The superior mechanical properties of APMT2 and PMC26M regarding the current cladding material of Zircaloy-2 (Fig. 2) allows for the reduction of the wall thickness of the IronClad material to obtain a zero penalty on the parasitic thermal neutron absorption (Table I).

Fig. 2
figure 2

Comparative yield stress of APMT2, PMC26M and Zirc-2 tubes at 20°C and 315°C.3

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To compare the superior strength of the IronClad tubes to the Zirc-2 tubes, a test was performed in an out-of-pile autoclave where tubes of APMT and C26M with caps welded at both ends were exposed side by side to a Zirc-2 tube also with welded caps at both ends. (The FeCrAl tubes used were earlier versions of IronClad.) All the tubes in Fig. 3 were exposed for 120 h to high-purity liquid water at 350°C and 3000 psi pressure containing 1 ppm dissolved oxygen and flowing at a rate of 175 mL/min. Figure 3 shows that after the autoclave exposure, all the FeCrAl tubes maintained their original geometry, while the Zirc-2 tube collapsed because of a combination of pressure and temperature, probably by a process of creep.

Fig. 3
figure 3

Before and after tubes of IronClad and Zirc-2 tubes exposed for 120 h at 350°C and 3000 psi water.

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In a more recent test in an overall argon atmosphere, IronClad (APMT2 + PMC26M) and Zirc-2 tubes were tested for creep resistance at 450°C by measuring the outside diameter (OD) expansion in internally pressurized tubes.7 The internally applied pressure was constant to obtain a pre-set value of calculated hoop stresses in the OD of the tubing. It was reported that a Zirc-2 tube with nominal 20 ksi hoop stresses had a large diametral strain (creep) in just over 200 min (3.3 h) of testing at 450°C, while both the PMC26M and APMT2 tubes with a nominal hoop stress of 47 ksi did not have measurable diametral strain even after 60,000 min (2500 h) of testing in the argon atmosphere.7 It was also reported that the creep resistance of APMT2 was higher than that of PMC26M when the tests were carried out at 600°C and 800°C. The superior creep resistance of APMT2 was more evident at the higher temperature of 800°C.7 The higher creep resistance of APMT2 compared to PMC26M can be attributed to the microstructure of APMT2, with numerous randomly distributed nanosized precipitates. Figure 4a shows the microstructure of the 0.3-mm wall thickness APMT2 tube that meets the geometry requirements for BWR fuel cladding. The grain size is ASTM 11 (8.5 µm average grain size) with a fully recrystallized microstructure (no retained strain) (Fig. 4b). Figure 4c shows randomly dispersed nanosized precipitates appearing as white or black dots.

Fig. 4
figure 4

Microstructure of APMT2 nuclear grade material with small and uniform grain size. SEM images acquired at (a) 200× magnification, fully recrystallized tube, (b) 500× magnification, grain size ASTM 11 ± 0.3, and (c) 1000× magnification, showing white and black nanosized precipitates.

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Jia et al.8 observed that FeCrAl alloys with oxide dispersion strengthening (ODS) have superior performance compared with traditional FeCrAl alloys. This difference in the creep behavior may not be apparent at the intermediate temperatures (< 800°C) but the slower creep of ODS FeCrAl is evident when the temperature nears 1000°C. Aragón et al.5 observed that at 800°C and at a stress of 60 MPa the creep strain of a generic FeCrAl alloy was approximately four orders of magnitude lower than for Zircaloy-4. (The value of 60 MPa was selected because this is the stress on the Zircaloy cladding before the onset of ballooning).5 Similarly, Bell et al.9 reported that, when compared to Zircaloy-2 tubes, FeCrAl (C26M) tubes had up to 250°C higher burst temperatures at all applied hoop stresses and C26M tubes had limited ballooning and minimal opening areas at burst locations. Massey et al.10 demonstrated the importance of ODS on creep resistance in an alloy fabricated using a sequence of mechanical alloying and extrusion, which had a similar composition as C26M (Fe + 12Cr + 6Al + 0.3Zr + 0.3Y2O3). The tube specimens were tested under both the as-received (AR) and recrystallized (RXA) conditions (950°C for 2 h + 1200°C for 1 h in argon atmosphere).

Resistance to Debris Fretting in Water at 288°C

It is understood that currently the major failure mode of the Zr alloys clad fuel in LWRs is debris fretting from the OD of the tube, which may perforate the wall of the cladding allowing the coolant water to enter in contact with the radioactive elements in the fuel cavity.11 The perforation of the cladding wall by fretting wear may occur by repetitive or oscillatory touching of the cladding wall by a foreign material such as a piece of wire trapped in the separation grids. One of the main considerations for replacing the actual Zr-based material of the cladding by an ATF material is that it must comply with the requirement of better resistance to debris fretting. It has been shown that the FeCrAl alloys are more resistant to fretting wear than a Zr-based alloy.12,13,14 Sakamoto et al.14 showed a lower wear depth of FeCrAl ODS compared to Zircaloy-2 bars under dry and wet sliding wear test conditions at ambient temperature. Winter et al.13 performed fretting tests of APMT specimens from 23°C to 350°C using a normal force of 100 N at 30 Hz for up to 100,000 cycles. They calculated the coefficient of friction (COF) since the higher the COF was, the higher would be the wear effect. They also calculated the volume wear loss using 3D confocal microscopy analysis and identified the type of material in the damaged areas using energy-dispersive spectroscopy (EDS). They found that the COF for APMT decreased as the temperature increased from ambient to 350°C suggesting lower wear at reactor operation conditions.13 This reduction in the COF could be the result of the oxide formed on the surface of APMT. Winter et al. also calculated that the wear volume (in µm3) of APMT decreased as the temperature increased, following the same trend as the COF.13

Measurement of Relative Wear Under BWR Conditions

A system was designed at GE Research to evaluate the relative wear susceptibility of APMT and C26M regarding Zircaloy-2 tubes in water at 288°C. The water was nearly pure, only containing 1 ppm dissolved oxygen, and it was recirculated in the autoclave at a rate of 175 mL/min. The wear test consisted of a type 304SS spring wire of 0.031-inch diameter, which was attached firmly to a central shaft. As the shaft rotated at 30 rpm, the spring wire repeatedly touched up to eight vertically fixed tubes of Zirc-2, APMT2 and PMC26M. The test normally lasted 14 days, and it was repeated three times. Figure 5 shows the equipment set-up, and Fig. 6 shows the position of the test tubes regarding the central rotating shaft. As the wire touched the tubes, it left a groove or ditch in the tubes. Figure 7 shows the relative penetrations of the wear in Zircaloy-2 (a1 and a2) and APMT2 (b1 and b2) tubes. After 14 days of testing, the penetration in the Zircaloy-2 tube was well defined and approximately 600 µm deep (across the wall thickness of the tube) (Fig. 7a1 and a2), and the penetration in the APMT tube wall was less defined and approximately 150 µm deep (Fig. 7b1 and b2). Table II shows the depth of the grooves (in µm) produced by the rotating spring wire in the three types of tested tubes for three different tests sets (Test 1, Test 2 and Test 3). In each test, the conditions of the wire or time of testing could be slightly different so a fair comparison among the three types of tubes should be within the same test (e.g., Test 1). In general, Table II shows that for each test, the deepest groove was always for the Zirc-2 tube, followed by the C26M tube and lastly by the APMT2 tube. That is, APMT showed superior resistance to wear in water at 288°C. The current results confirm the previous finding by other international authors using other test methods.12,13,14

Fig. 5
figure 5

Autoclave set-up for the wear test at GE Research.

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Fig. 6
figure 6

Position of the test tubes regarding the central shaft and the fixed spring wire in the rotating shaft.

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Fig. 7
figure 7

Penetration of a wire wear during testing for 14 days in 288°C water. (a1 and a2) is for a Zirc-2 tube. (b1 and b2) is for an APMT tube. Profiles show that the penetration was deeper for the Zirc-2 tube (a1) than for the APMT tube (b1).

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Table II Comparison of the wear depth (µm) for three tests for spring wire touching sequentially the three types of tubes (Zirc-2, APMT2 and PMC26M)
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Summary and Conclusion

  1. 1.

    Accident-tolerant materials for LWR cladding applications include FeCrAl alloys.

  2. 2.

    The FeCrAl materials include APMT, C26M and ODS FeCrAl, now all made by powder metallurgy processes.

  3. 3.

    Testing performed internationally shows that the mechanical properties (i.e., yield stress) at 300°C and at higher temperatures are higher for the FeCrAl alloys than for Zr-based alloys. If the FeCrAl alloy contains nanosized particles randomly dispersed in its matrix, the mechanical strength is even higher.

  4. 4.

    The creep resistance of the FeCrAl at 450°C and higher temperatures is superior to that of Zr alloys by several orders of magnitude. For FeCrAl alloys containing nanosized particles, the creep resistance is the highest, especially at temperatures near 1000°C.

  5. 5.

    The wear resistance of FeCrAl alloys near 300°C is higher than for Zr alloys. Different test methods arrived at the same conclusion. The FeCrAl alloys seem more resistant to wear near 300°C than at ambient temperature.