17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

STRESS CORROSION CRACKING OF PROTON IRRADIATED ALLOY 33 FOR ACCIDENT TOLERANT FUEL CLADDING

Parag Ahmedabadi, Gary S. Was

Nuclear Engineering & Radiological Sciences, University of Michigan, Bonisteel Boulevard Ann Arbor, MI 48109, USA

ABSTRACT

Alloy 33 is a candidate alloy for accident tolerant fuel cladding due to its good corrosion resistance and low reactivity compared to zirconium alloys. The effect of irradiation on stress corrosion cracking (SCC) behavior of Alloy 33 was evaluated using irradiation with 2 Me protons to 5 dpa at 360°C. The hardness increased from 287 HV in the un-irradiated condition to 567 HV (97% increase) in the irradiated condition. Stress corrosion cracking behavior was evaluated in a simulated boiling water reactor (BWR) environment under normal water chemistry (NWC) using constant extension rate tensile (CERT) test in which samples were strained up to 6% at a strain rate of 3×10-7 s-1 at 288 °C. Scanning electron microscopy (SEM) after the test revealed intergranular cracking in the irradiated region but not in the un-irradiated region. Dislocation channels were observed in the irradiated region, which were absent in the un-irradiated region. However, several ruptures of the oxide film were observed in the un-irradiated region, but not in the irradiated region. Raman analysis of the irradiated as well as un-irradiated zones indicated that the oxide-film consisted mainly of NiFe2O4 and FeCr2O4. No qualitative difference in the oxide film was detected between the un-irradiated and irradiated zones.

1. INTRODUCTION

Alloy 33 is an austenitic alloy with nominal composition (in wt %) 33 Cr, 32 Fe, 31 Ni, 1.6 Mo, 0.6 Cu, and 0.4 N. This alloy exhibits an excellent corrosion resistance in various corrosive media e.g. oxidizing and chloride containing environments [1-3]. The resistance of Alloy 33 in boiling 67% HNO3 is 6 and 2 times superior to AISI 304L and 310L, respectively [1]. In aggressive mixture of 20% HNO3 and 7% HF at 50 °C, the resistance of Alloy 33 is 200 and 2.4 times better than AISI 316Ti and Alloy 28, respectively [1]. The alloy also exhibited excellent resistance to localized corrosion in chloride-containing media [2]. For example, the critical pitting temperature in 10% FeCl3·6H2O is 85 °C [2]. The resistance to stress corrosion cracking (SCC) in saturated CaCl2 at 120 °C is even superior to Alloy 926 (UNS N80926) [2]. No sensitization could be detected for Alloy 33 when tested in boiling 65% HNO3 after ageing between 600 and 1000 °C [3]. In addition, Alloy 33 also has an excellent resistance to 96-98.5 H2SO4 up to a temperature of 150 °C. Alloy 33 has been approved by ASME for the use as pressure vessels the temperature range of -196 to 450 °C [2].

Alloy 33 is now being considered as a candidate cladding material in for accident tolerant fuel in light water nuclear reactors because of excellent corrosion resistance and low reactivity as compared to zirconium alloys. This paper focuses on the resistance of Alloy 33 to SCC in the proton-irradiated condition under simulated boiling water reactor (BWR) environment under normal water chemistry (NWC). The SCC resistance was evaluated using Constant Extension Rate Tensile (CERT) test at 288 °C with a strain rate of 3×10-7 s-1.

17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

2. EXPERIMENTAL

2.1 Material and Sample Preparation

The alloy used in this investigation was in the form of a rolled plate of size 110 mm × 50 mm × 16 mm. The chemical composition (in wt. %) of the alloy was 33 Cr, 32 Fe, 31 Ni, 1.6 Mo, 0.6 Cu, and 0.4 N.

Tensile specimens with a total length of 41 mm and a gage length of 21 mm and square cross-section of 2 mm × 2 mm were prepared using Electrical Discharge Machining (EDM). Tensile specimens were mechanically polished using different grades of SiC paper followed by electropolishing in methanol (90%) and perchloric acid (10%) solution at a minimum temperature of -30 °C with an applied voltage of 30 V for ~ 30 s. Electropolishing was done to ensure removal of plastic deformation on the surface introduced during mechanical polishing.

2.2 Proton Irradiation

A tensile specimen of Alloy 33 in the as-received condition was irradiated to a dose of 5 dpa using 2 MeV protons in the Michigan Ion Beam Laboratory using a stage attached to a 1.7 MV Tandem accelerator. Irradiation was conducted at a dose rate of ~ 2 × 10-5 dpa/s at 360 °C. The dose was calculated using SRIM2013 [4] in full cascade mode. The irradiation parameters produced a nearly uniform damage rate through the first 15 µm of the total proton range of 20 µm. The dpa was calculated using a displacement energy of 40 eV.

2.3 CERT Test

The CERT test was conducted on the proton-irradiated specimen (5 dpa) in a multi-specimen recirculating autoclave system in the High Temperature Corrosion Laboratory (HTCL) at the University of Michigan. The specimen was strained to a total strain of ~ 6% at 288 °C using a strain rate of 3×10-7 s-1. The temperature within the autoclave was maintained at 288 ± 1 °C and the pressure within the autoclave was maintained at 9.60 ± 0.30 MPa. The conductivity of the water at the inlet was < 0.06 µS/cm and at the outlet was < 0.1 µS/cm throughout the experiment. The dissolved oxygen level was maintained at 2 ppm by purging a mixture of argon and oxygen throughout the experiment and during the heating of the vessel.

2.4 Microhardness Measurements

The hardness of the irradiated material was measured using a Vickers hardness tester capable of applying load as low as 25 g. All hardness measurements were carried out using 25 g force to confine the indentation within the irradiated zone. As mentioned above, the depth of uniform irradiation is ~ 15 µm,

and the indent depth should be less than 5 m. The hardness of the irradiated material was measured on a rectangular bar of 21 mm × 2 mm × 2 mm prepared from the as-received block. The rectangular bar was irradiated simultaneously with the tensile specimen of the same alloy. The hardness was measured at two different locations (with a minimum of 10 indents) in the un-irradiated and the irradiated zones.

2.5 SEM Examination

After fracture during CERT tests, the gage surface and fracture surface were examined using a scanning electron microscope (JEOL-JSM-6480). Examination of the gage surface was conducted to determine the presence of cracks in the un-irradiated and the irradiated zones after the CERT tests. After SEM examination, oxide films were removed from the specimens to better resolve small surface cracks that

17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

might be masked due to the presence of the oxide film formed during the exposure to a simulated BWR environment.

2.5 Characterization of Surface Oxidation

Oxide films formed in irradiated and un-irradiated zones after the CERT test in simulated BWR-NWC environment were characterized using Raman spectroscopy. Raman analysis was done using a Renishaw inVia Raman Microscope equipped with a RenCam CCD detector with resolution of 1 cm-1, 1800 lines/nm grating, and 50 µm slit. An objective lens of 20x was used to give a laser beam spot size of ~ 1 µm. The spectra were excited using 633 nm red laser with power of 9 mW. The incident angle for the laser beam on specimen was 90°. The acquisition time was 15 s and each spectrum was collected twice. Each spectrum was scanned between wavenumbers 1450 to 150 cm-1. All experimental parameters were kept identical between different measurements.

3. RESULTS AND DISCUSSION

3.1 Stress-strain Behavior

The stress-strain curve for the CERT test on the proton-irradiated specimen up to a total strain of ~ 6% is shown in Fig. 1. The 0.2% offset yield strength was estimated to be 240 MPa. The depth of the irradiation

was about 20 m and accounted for only 1% of the sample cross-sectional area. Therefore, the presence of the proton-irradiated zone is not expected to affect the overall stress-strain behavior of the sample.

3.2 Microhardness Measurements

The average hardness in the un-irradiated zone was 287 HV and the average hardness of the irradiated zone was 567 HV, resulting in a 97% increase in the hardness due to irradiation. The hardness of an austenitic material is expected to increase due irradiation hardening [5]. It was shown that the increase in the hardness after irradiation is due to formation of dislocation loops during irradiation [6].

3.3 SEM Examination

SEM examination of the irradiated zone after the CERT test in a simulated BWR-NWC revealed several cracks. No such cracks were observed in the un-irradiated zones of the specimen. Cracks in the irradiated zone are shown in Figures 2a through 2c. The morphology of cracks indicated that they are intergranular in nature. In addition to cracks, oxide particles were observed over the entire gage section. The oxide film formed in the un-irradiated zone ruptured at many locations as shown in Figure 2d. No such rupturing of the oxide film was observed in the un-irradiated zone.

Figure 3a and 3b are micrographs of the irradiated zones after removal of the oxide film; intergranular cracks are clearly visible in the micrographs. Further, deformation bands (as marked by arrows in the Figure 3a and 3b) were observed within the matrix in almost all grains. It has been shown that irradiation leads to localized deformation in the form of dislocation channels during tensile loading of irradiated austenitic alloys [7, 8]. The deformation bands shown in Figure 3 are morphologically similar to dislocations channels reported earlier for austenitic stainless steels [7, 8]. No such localized deformation regions were observed in the un-irradiated zone.

It is suspected that localized deformation plays a primary role in increasing susceptibility to Irradiation Assisted Stress Corrosion Cracking (IASCC) of AISI 304 in simulated BWR environment [8]. The presence of intergranular cracks in the irradiated zone and the absence of the same in the un-irradiated

17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

zone indicate that the irradiation has enhanced susceptibility of Alloy 33 to intergranular stress corrosion cracking. The presence of localized deformation in the irradiation zone is consistent with the large increase in the hardness of the irradiated region.

3.4 Characterization of Oxide Film

Raman spectra obtained from the un-irradiated and the irradiated zones on the tensile specimen are shown in Figure 4a and Figure 4b, respectively. Both spectra are identical in nature with respect to positions of peaks as well as corresponding intensity. The identification of the various peaks indicated that the oxide film consisted of NiFe2O4 and FeCr2O4 in both, the un-irradiated and irradiated zones.

4. CONCLUSIONS

The presence of intergranular stress corrosion cracks in only the irradiated zone of the Alloy 33 (proton-irradiated to 5 dpa) in a simulated BWR-NWC indicated that irradiation enhances the susceptibility to SCC. This was further confirmed by the presence of localized deformation as well as a large increase in the hardness in the irradiated zone.

References:

1. M. Kohler, U. Heubner, K.-W. Eichenhofer, M. Renner, “Alloy 33: A new material for handling of HNO3/HF media in reprocessing of nuclear fuel”, CORROSION/ 97, Paper No. 115, Houston, Texas, NACE International, 1997.

2. D.C. Agrawal, M. Kohler, “A new material resisting marine corrosion”, CORROSION/ 97, Paper No. 424, Houston, Texas, NACE International, 1997.

3. M. Kohler, U. Heubner, K.-W. Eichenhofer, M. Renner, “Progress with Alloy 33 (UNS R20033), A new corrosion resistant chromium-based austenitic alloy”, CORROSION/ 96, Paper No. 428, Houston, Texas, NACE International, 1996.

4. J.F. Zeigler, SRIM 2013 Program, http://www.srim.org/SRIM/SRIMLEGL.htm.

5. J. Gan, G.S. Was, “Microstructure evolution in austenitic Fe-Cr-Ni alloys irradiated with Protons: comparison with neutron-irradiated microstructures”, J. Nucl. Mater. 297 (2001) 161-175.

6. G. S. Was “Fundamentals of Radiation Materials Science”, Springer Verlag, First Edition, 2007.

7. E. West, M. McMurtrey, Z. Jiao, G.S. Was, Metall. Mater. Trans. A 43 (2012) 136-146.

8. M. McMurtrey, G.S. Was, L. Patrick, D. Farkas, Mater. Sci. Eng., A 528 (2011) 3730.

17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

Figure 1: Stress-strain curve for the irradiated Alloy 33, strained up to ~ 6% at a strain rate of 3×10-7 s-1 at 288 °C in simulated BWR environment.

17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

Figure 2: SEM micrographs of the tensile specimen of Alloy 33 (irradiated to 5 dpa) after the CERT test in a simulated BWR-NWC environment showing (a), (b), (c) intergranular cracks in the irradiated zone and (d) rupture in the oxide film at several locations in the un-irradiated zone. The loading direction is in the vertical direction.

(b) (a)

(c) (d)

17th International Conference on Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors August 9-13, 2015, Ottawa, Ontario, Canada

Figure 3: SEM micrographs of the irradiated zone of the irradiated Alloy 33 after the CERT test showing intergranular cracks and deformation bands. The loading direction is in the vertical direction.

Figure 4 Raman spectrum for the tensile specimen of the Alloy 33 (5 dpa) taken from the (a) un-irradiated zone and (b) irradiated zone. Both spectra are identical in nature.

(a) (b)

2 µm 5 µm

(a) (b)

NiFe2O

4

FeCr2O

4

NiFe2O

4 NiFe2O

4

NiFe2O

4

FeCr2O

4