Evaluation of Hydride Reorientation Behavior and Mechanical Properties for High-Burnup Fuel-Cladding Tubes in Interim Dry Storage

Transcription

1 Journal of ASTM International, Vol. 5, No. 9 Paper ID JAI Available online at Masaki Aomi, 1 Toshikazu Baba, 1 Toshiyasu Miyashita, 1 Katsuichiro Kamimura, 1 Takayoshi Yasuda, 2 Yasunari Shinohara, 3 and Toru Takeda 4 Evaluation of Hydride Reorientation Behavior and Mechanical Properties for High-Burnup Fuel-Cladding Tubes in Interim Dry Storage ABSTRACT: The hydride stress reorientation behavior and the mechanical properties of irradiated cladding tubes were investigated to evaluate the high-burnup fuel-cladding tube properties in interim dry storage. As for the boiling water reactor BWR Zircaloy-2 Zry-2 cladding, the hydride reorientation to the radial direction occurred at relatively low hoop stresses during the hydride reorientation treatment HRT, such as less than 70 MPa. The increase of reorientation with hoop stress was not monotonic for the specimens in which a part of the hydrides remained precipitated at the HRT temperature, such as the case for 50GWd/t type cladding at a 300 C HRT. The degree of reorientation depended on the HRT solution temperature rather than on the estimated temperature at which the hydride precipitation occurred under the relatively moderate HRT conditions. In the relatively low cooling rate HRT, the hydride preferential precipitation in the Zr liner increased for Zr lined cladding compared to that in a relatively high cooling rate. The ductility of the specimens after the 300 C HRT showed relatively good correlation to the Polymax index which reflects the length or continuity of the hydrides regardless of their orientation. The ductility of the specimens after the 400 C, 0 MPa, 30 C/h HRT increased in ring compression testing at room temperature compared to no HRT as-irradiated specimens, and it indicated recovery of irradiation damage occurred at the 400 C annealing temperature and affected the ductility of the irradiated Zry-2 cladding. As for the pressurized water reactor Zircaloy-4 cladding, little increase in the radial hydride ratio occurred in a 100 MPa, 340 C or less HRT. On the other hand, the amount and the length of the hydride in the midwall area of the cladding depended on the temperature and the cooling rate from the HRT due to hydrogen migration from the hydride rim area. It is deduced that the ductility in ring compression deformation was affected by the orientation, amount, and length of hydride in the midwall area. KEYWORDS: fuel cladding tube, Zry-2, Zry-4, MDA, ZIRLO, dry storage, high burnup, hydride reorientation, radial hydride Introduction In dry storage of spent fuel, a hoop stress is imparted on the cladding by the inner gas pressure, and the cladding temperature decreases gradually from several hundred degrees Celcius over several decades. Under such conditions, the stress reorientation of hydride to the radial direction of the cladding tube and a decrease in the cladding ductility due to an increase of radial hydride may occur, depending on the temperature and hoop stress condition. The hydride stress reorientation behavior in zirconium alloys has been studied and reported by many researchers 1 8. But considering the somewhat moderate conditions of dry storage, such as temperatures around 300 C and a hoop stress of 100 MPa, sufficient data have not been accumulated, especially for the low cooling rates of the irradiated cladding. The first objective of this study is to evaluate the correlation between the hydride morphology change including reorientation and the reorientation conditions such as temperature, hoop stress, and cooling rate for irradiated cladding tubes. Radial hydrides in cladding are well known to degrade the cladding ductility under circumferential stress 6,9,10. The second objective of this study is to examine the correlation between cladding mechani- Manuscript received May 31, 2007; accepted for publication August 4, 2008; published online October Japan Nuclear Energy Safety Organization, Toranomon, Minato-ku, Tokyo, Japan. M. Aomi is currently with Nippon Nuclear Fuel Development Co., Ltd. 2 Nippon Nuclear Fuel Development Co., Ltd., 2163 Narita-cho, Oarai-machi, Ibaraki-ken, Japan. 3 Nuclear Development Corp., Funaishikawa, Tokai-mura, Ibaraki-ken, Japan. 4 Nuclear Fuel Industries, Ltd., Muramatsu, Tokai-mura, Ibaraki-ken, Japan. Copyright 2008 by ASTM International, 100 Barr Harbor Drive, PO Box C700, West Conshohocken, PA

2 652 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM cal properties at room temperature and hydride morphology, including orientation, length, and total hydride amount. The hoop stress due to inner gas pressure in storage is estimated not to be large enough to cause the cladding to breach, which might result in some degree of fuel dispersion out of the cladding and might affect the subcriticality analysis, even for cladding with heavy radial hydrides. In this study, compression of the cladding tube in the radial direction caused by either rod rod or rod-cask interaction in a drop accident at room temperature was taken as the deformation mode to be evaluated with high priority because it might have the possibility of a cladding breach. An evaluation of the correlation between cladding circumferential ductility in ring compression deformation and the hydride morphology in the cladding is one of the main goals in this study. In addition, ring tensile tests and longitudinal tensile tests were also performed under a wide strain rate range at room temperature using a small number of the irradiated specimens to evaluate the strain rate dependence of the irradiated cladding mechanical properties. Both the hydride reorientation behavior and the fracture behavior under ring compression deformation of irradiated cladding are governed by the controlling phenomena, such as hydrogen dissolution, hydride reorientation, crack initiation, and propagation, including the effects of temperature on irradiation damage of the material. Each of these phenomenon has not been sufficiently evaluated, especially for irradiated cladding tubes to allow first principals determination. Therefore, the authors focused this study on the investigation of the macroscopic behavior of irradiated cladding tubes under the conditions expected in dry storage and discuss some elemental phenomena. Experimental Materials The cladding tube materials examined in this study are listed in Table 1. Zircaloy-2 Zry-2 cladding from boiling water reactor BWR spent fuel and Zircaloy-4 Zry-4 cladding from the pressurized water reactor PWR spent fuel are the main test materials. In addition, MDA and ZIRLO cladding prepared from PWR spent fuel were also tested. Hydride Reorientation Test The test matrix for the hydride reorientation testing is shown in Table 2 for BWR cladding and Table 3 for PWR cladding. Temperature, hoop stress, and cooling rate of the cladding specimens are the test param- BWR PWR Fuel cladding type 40GWd/t no linear type 50GWd/t type 55GWd/t A a type 55GWd/t B a type 39GWd/t type 48GWd/t type 55GWd/t A type TABLE 1 Test materials. Chemical composition of alloying elements wt % Heat treatment SPP mean diameter m Fluence b Material Sn Fe Cr Ni Nb in tube production All Zr-Fe-Cr Fr Burnup b GWd/t l/m 2, E 1 MeV Zry RXA Zry-2 with Zr liner unirradiated unirradiated N.M. c Zry SRA N.M. N.M MDA ZIRLO a A, B : different fuel maker. Rod average. Not measured.

3 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 653 TABLE 2 Test matrix for BWR cladding. (a) Irradiated BWR 50GWd/t type Zry-2 (with liner). (b) Irradiated BWR 55GWd/t (a) type Zry-2 (with liner). (c) Irradiated BWR 55GWd/t (b) type Zry-2 (with liner). (d) Irradiated BWR 40GWd/t type Zry-2 (with no liner). (e) Unirradiated BWR 50GWd/t type Zry-2 (with liner). A: hydride reorientation test (constant hoop stress); A : hydride reorientation test (gas was closed packed in the tube and hoop stress decreased with temp.); B: ring compression test; C: longitudinal tensile test (low strain rate); and D: ring tensile test, longitudinal tesile test (each test including high strain rate). a Temperature Cooling Hoop stress MPa C K rate C/h 0 heat treatment without stress As-irradiated no heat treatment A+B A+B A+B A+B A+B B A+B+C A+B A+B A+B A+B+C A+B A+B A+B A+B A A+B A+B A+B 0.6 A+B A+B b Temperature Cooling rate 0 heat treatment without stress Hoop stress MPa As-irradiated no heat treatment C K C/h A A A A A A A A A A A A c Temperature Cooling rate 0 heat treatment without stress Hoop stress MPa As-irradiated no heat treatment C K C/h A D d Temperature Cooling rate 0 heat treatment without stress Hoop stress MPa As-irradiated no heat treatment C K C/h A+B B e Temperature Cooling rate 0 heat treatment without stress Hoop stress MPa As-irradiated no heat treatment C K C/h A A+B A B A A A eters. The temperature conditions were set considering the cladding temperature in dry storage. The hoop stress conditions were set to find the threshold stress for reorientation at each temperature. The temperature and the hoop stress conditions in Tables 2 and 3 are relatively moderate conditions for hydride reorientation compared to those reported in the literature 1 7. In hydride reorientation testing, a hydride reorientation treatment HRT was performed as shown in Fig. 1. The biaxial stress in the cladding tube specimens was applied by an inner pressure of Ar gas. The specimen temperature was held for 30 for PWR cladding or 60 for BWR cladding minutes at the HRT solution temperature in the furnace to dissolve the hydrogen, then it was decreased to around room temperature to precipitate the hydride. The hoop stress in the cladding was held constant as shown in Fig. 1 b, except for one case, which is noted as A in Table 2 c. The morphology of the hydrides, including the orientation, was evaluated from metallography after the HRT. The hydride morphology of as-irradiated no HRT cladding was also evaluated as the reference. Ring Compression, Ring Tensile, and Longitudinal Tensile Test The mechanical-property change due to radial hydride reorientation was evaluated mainly using the ring compression test at room temperature. The ring compression test method is shown schematically in Fig. 2 a. For the ring compression test, ring specimens 8 mm in length were prepared from the cladding tube after the HRT. Ring specimens were compressed in the radial direction on the flat plane with a crosshead speed of about 2 mm/min at room temperature. The test was also was performed on as-irradiated specimens as the reference.

4 654 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM TABLE 3 Test matrix for PWR cladding. (a) Irradiated PWR 48GWd/t type Zry-4. (b) Irradiated PWR 39GWd/t type Zry-4. (c) Irradiated PWR 55GWd/t type MDA and ZIRLO. A: hydride reorientation test (constant hoop stress) and B: ring compression test. a Temperature Cooling Hoop stress MPa C K rate C/h 0 heat treatment without stress As-irradiated no heat treatment A+B A+B B A+B A+B A+B A A+B A A+B A+B A+B A+B A+B A+B A+B A 0.6 A+B b Temperature Cooling rate 0 heat treatment without stress Hoop stress MPa As-irradiated no heat treatment C K C/h A c Temperature Cooling rate 0 heat treatment without stress Hoop stress MPa As-irradiated no heat treatment C K C/h A A In addition, ring tensile tests and the longitudinal tensile tests were performed over a wide strain rate range s 1 for ring tensile and s 1 for longitudinal tensile at room temperature using a small number of irradiated BWR cladding specimens in order to evaluate the strain rate dependence of the mechanical properties. Both the ring tensile specimens with one side guage and the longitudinal tensile specimens were prepared from cladding tubes as shown in Figs. 2 b and 2 c. The dimension of the specimens and the details of the test method for the ring tensile test were basically the same as those described in the literature 11. The HRT conditions of the mechanical property tested specimens also are shown in Tables 2 and 3. Results for BWR Zry-2 Cladding Hydride Reorientation Test for Zry-2 Cladding Figure 3 shows the metallography of BWR Zry-2 cladding samples before and after HRT. As shown in Figs. 3 b and 3 g as compared to Fig. 3 a, a small degree of hydride reorientation to the radial direction was observed after HRT at a hoop stress of 70 MPa, temperature of 300 C, and cooling rate of 30 C/h. In Figs. 3 b and 3 g hydride reorientation was observed more clearly in the Zry-2 inner side area near the Zr liner, although little reorientation seemed to occur in the area near the outer side. On the other hand, a large degree of hydride reorientation to the radial direction was observed for specimens after an HRT at FIG. 1 Test method of hydride reorientation test.

5 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 655 FIG. 2 Test methods of ring compression test, ring tensile test, and longitudinal tensile test. 70 MPa and 400 C. At that temperature all the hydrogen was estimated to be dissolved completely for most of the specimens. Figures 3 d 3 f are the results for unirradiated cladding specimens. The hydrogen content of unirradiated cladding specimens was so low that the hydrogen dissolved completely at each HRT temperature. The reorientation for irradiated cladding seems somewhat larger than that for unirradiated cladding, comparing Figs. 3 c and 3 f. In order to quantify the hydride reorientation dependence on the HRT temperature, hoop stress and cooling rate, the degree of radial hydride orientation was evaluated using Fn 40 and Fl 45 12, which are defined in Eqs 1 and 2 below. Here, hydrides in the metallography, except not in the Zr liner nor at the interface between the Zr liner and Zry-2 matrix area, were analyzed for both Fn 40 and Fl 45. The hydrides with less than 16 m in length were excluded for Fn 40. Fn 40 is the conventionally used definition, and Fl 45 is a second indicator of the amount of radial hydride Fn 40 = Sum of the number of hydrides in radial direction 40, 1 Sum of the number of all hydrides Sum of the length of hydrides in radial direction 45 Fl 45 =. 2 Sum of the length of all hydrides In Fig. 4, Fn 40 and Fl 45 are plotted versus HRT hoop stress for BWR Zry-2 claddings with Zr liner. For 300 C HRT specimens, a slight increase in Fn 40 and Fl 45 was observed at HRT hoop stresses of 40 and 70 MPa, and significant hydride reorientation occurred at HRT hoop stress of 100 MPa. In 400 C HRT specimens, both Fn 40 and Fl 45 increased with the hoop stress within the tested hoop stress range. Figure 5 shows the effect of cooling rate on Fn 40, Fl 45, and the sum of the length of hydrides per area, which reflects the total hydride amount in the Zry-2 cladding. As shown in Figs. 3 h and 3 i, the amount of hydride in the inner Zry-2 area near the Zr liner decreased and the amount of hydride in the Zr liner increased with decreasing cooling rate. The effect of cooling rate on hydride morphology described above was confirmed quantitatively in Fig. 5. This influence of cooling rate on the Zr lined cladding is

6 656 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM FIG. 3 Metallography of BWR Zry-2 cladding specimens in hydride reorientation test [Radial cross section except for b ]. easily understood as the result of larger hydrogen migration in the lower cooling rate samples from the Zry-2 matrix to the Zr liner due to the terminal solid solubility precipitation TSSp difference between Zr and Zry Zry-2 cladding with no Zr liner showed a relatively large degree of hydride reorientation as observed in Fig. 3 j, which was a different behavior from that for Zr lined cladding.

7 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 657 FIG. 4 Correlation between the degree of reorientation and the HRT conditions for irradiated BWR Zry-2 cladding with Zr liner (cooling rate: 30 C/ h). Ring Compression Test for Zry-2 Cladding The metallography of BWR Zry-2 specimens after ring compression testing is shown in Fig. 6. Cracking was observed in the normal direction to the ring compression load axis in the ring specimens with a Zr liner Figs. 6 a 6 e and 6 g. On the other hand, cracking was observed on the ring compression load axis in the ring specimen of the cladding with no Zr liner Fig. 6 f. In the former case, the crack seemed to propagated from the outer surface to the inner surface along the hydrides. In the latter case, the crack seemed to propagate from the inner surface to the outer surface. A circumferential tensile stress is applied at the inner surface along the ring compression load axis in the early stage of the ring compression deformation. It is suggested that the crack was not initiated in the Zr liner inner surface because the Zr liner has high ductility, even after irradiation and the precipitation of a high content of hydrides. The ring compression test results were evaluated from the crosshead displacement ratio as defined in Fig. 2 a. Figure 7 summarizes the results for BWR Zry-2 cladding with a Zr liner. The crosshead displacement ratio did not show a hoop stress dependence for the 300 and 250 C HRT specimens, although some degree of reorientation was observed in the 300 C, MPa HRT specimens. On the FIG. 5 Effects of HRT cooling rate on hydride morphology for irradiated BWR Zry-2 cladding with Zr liner. [ Sum of the length of hydrides per area 1/mm Sum of the length of hydrides those are measured in the observed area (mm) / The area where the measured hydrides are observed 1/mm 2. All hydrides marks represent the value for all hydrides regardless of the orientation. Radial hydrides marks represent the value for the hydrides in radial direction 45.]

8 658 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM FIG. 6 Metallography of BWR Zry-2 specimens after ring compression test (Room temperature. Crosshead speed 2 mm/ min. (a) (e), (g): 50GWd/t type, and (f): 40GWd/t type). other hand, the crosshead displacement ratio increased for the 400 C, 0 MPa HRT heat treatment only with no stress specimens compared to as-irradiated specimens, and it decreased with an increase of the hoop stress in HRT or with the amount of radial hydride. The effect of cooling rate on ductility is shown in Fig. 8. The crosshead displacement ratio increased with a decrease of cooling rate for the Zr liner cladding. It is consistent with the decrease the amount of radial hydride at low cooling rate as shown in Fig. 5. The crosshead displacement ratio after a 300 C, 70 MPa, 3 C/h HRT for cladding with no Zr liner indicates a low ductility cross head displacement ratio of 0.3 % 0.4 % among the BWR specimens in the ring compression test, although the hydrogen content of the no liner specimens was the lowest 40 ppm H among the materials tested. It is deduced that this result was due to both the large ratio of radial hydride and the crack growth initiating from the inner surface, which is different from the case for Zr lined cladding. Ring Tensile Test and Longitudinal Tensile Test for BWR Zry-2 Cladding Figure 9 provides the results of longitudinal tensile tests and ring tensile tests at room temperature for irradiated BWR Zry-2 cladding with Zr liner. The effects of HRT on longitudinal mechanical properties were not observed at the strain rate of s 1 in Fig. 9, which is consistent with the fact that most of the hydrides precipitated along the longitudinal axis and few hydrides orientated perpendicular to longitudinal direction in Fig. 3 b. In Fig. 9, it also is suggested that the circumferential strength was larger than the longitudinal strength, and the strength was slightly increased with the strain rate both in

9 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 659 FIG. 7 Correlation between the ductility of the specimens and HRT conditions for irradiated BWR Zry-2 cladding with Zr liner. longitudinal and circumferential properties. It is well known that the plastic anisotropy of cladding tube formed in tube fabrication is reduced after irradiation because the irradiation defects affect the prism slip. Figure 9 indicates that some degree of anisotropy at room temperature remained after the irradiation to a fast neutron fluence of the approximate /m 2. Results for PWR Zry-4 and Improved Alloy Cladding Hydride Reorientation Test The results of hydride reorientation tests for irradiated PWR 48GWd/t type Zry-4 cladding were evaluated in the same way as for BWR cladding, and the results are shown in Figs. 10 and 11. The hydrides in the midwall area area except for hydride rim area were used in the image analysis to calculate Fl 45. The radial hydride ratio increased for the specimens after a 115 MPa, 300 C, 30 C/ h HRT compared to the as-irradiated specimens, as shown in Figs. 10 and 11. In addition, the sum of the length of radial hydrides per unit area in the midwall area, which reflects the total amount of hydride in the midwall area, increased with HRT temperature as shown in Fig. 12. Here, terminal solid solubility dissolution TSSd was calculated using the equation by Ogata et al. 14. The increase in the hydride amount over that for as-irradiated specimens in the midwall area was almost proportional to the TSSd at each HRT temperature. The effects of cooling rate on hydride morphology is shown in Fig. 13. Fl 45 in 3 C/h cooled specimens seemed to be larger than that in 30 C/h cooled specimens, while it was almost the same in 0.6 C/h cooled specimens Fig. 13 a. The maximum in the length of radial hydride increased with the decrease in cooling rate between 30 and 3 C/h, while the difference between 3 and 0.6 C/h became smaller, considering the scattering of the data in Fig. 13 b. The Fl 45 after a 300 C, 30 C/h HRT are compared among each cladding material in Figs. 10 and GWd/t type Zry-4 cladding showed a relatively high value compared to 48GWd/t type Zry-4 cladding. It is consistent with the fact that the Kearns factor Fr for the 48GWd/t type cladding is larger than that for 39GWd/t type cladding. As for the 55GWd/t type cladding materials, the Fl 45 for ZIRLO cladding were relatively large among the PWR cladding materials tested in this study, while MDA showed the same level of Fl 45 as 48GWd/t type Zry-4.

10 660 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM FIG. 8 Effects of HRT cooling rate on ductility in ring compression deformation for irradiated BWR Zry-2 cladding with Zr liner (50GWd/t type). FIG. 9 Correlation between strength and strain rate in ring tensile test and longitudinal tensile test for irradiated BWR Zry-2 cladding with Zr liner.

11 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 661 FIG. 10 Metallography of PWR cladding specimens before and after HRT. Ring Compression Test The metallography of the specimens after ring compression testing is shown in Fig. 15. Cracking was observed on the ring compression load axis in the ring specimen, and the crack seemed to have propagated from the inner surface where the circumferential tensile stress was relatively high at the early stage of ring compression deformation. The results of the ring compression test for 48GWd/t type Zry-4 cladding are summarized in Fig. 16. The crosshead displacement ratio was almost the same level for the specimens after a 250 C, 100 MPa and a 275 C, 100 MPa HRT compared to as-irradiated specimens. The specimens after a 340 C, 100 MPa and a 300 C, 100 MPa HRT showed a lower crosshead displacement ratio compared to as-irradiated specimens, although Fl 45 did not show an increase after these HRT conditions. As shown in Figs. 12 and 13, the length and the amount of hydride changed depending on the HRT temperature and the cooling rate. It is supposed that the morphology change including the orientation, length, and the amount of hydride affected the ductility. The effect of HRT cooling rate on ductility in ring compression deformation is shown in Fig. 17. The crosshead displacement ratio in 0.6 C/h cooling rate specimens seemed to be the same as that in 3 C/h specimans, while it decreased in 3 C/ h specimens compared with that in 30 C/ h cooling rate specimens. As shown in Fig. 13, an increase in both the degree of reorientation Fl 45 and the length of hydride was

12 662 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM FIG. 11 Correlation between the degree of reorientation and the HRT conditions for irradiated PWR Zry-4 cladding. observed in 3 C/h compared to that in 30 C/h specimens, while a remarkable increase was not observed in either factor in 0.6 C/h specimens compared to that in 3 C/h specimens. It is deduced that the effect of cooling rate on ductility is in agreement with the change in hydride morphology as illustrated in Fig. 13. Discussion Hydride Reorientation Behavior In this section, the observed stress and temperature dependence of hydride reorientation in BWR cladding is discussed. A fraction of hydrides present in the tested BWR cladding is estimated to remain precipitated at the HRT temperature except for the HRT 400 C condition because the hydrogen contents of the irradiated BWR specimens were over 150 ppm and higher than the TSSd at the HRT temperatures,. Therefore, the index of hydride reorientation Fne and Fle were defined as described in Eqs 3 and 4 for a quantitative understanding of the reorientation ratio of dissolved hydrogen at the HRT temperature FIG. 12 Correlation between the sum of the length of all hydrides per area in midwall area of the cladding and the estimated TSSd for irradiated PWR Zry-4 cladding. [The definition of vertical axis is the same as that in Fig. 5(b) for all hydrides.]

13 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 663 FIG. 13 Effects of HRT cooling rate on hydride morphology for irradiated PWR 48GWd/t type Zry-4 cladding. [The definition of vertical axis in (b) is the maximum value among the length of hydride in radial direction 45. ] Fne 40 = C H t Fn 40 C H t C H d Fn 40 0 C H d, 3 Fle 45 = C H t Fl 45 C H t C H d Fl 45 0 C H d, 4 where: C H t : total hydrogen content of the specimen, C H d : dissolved hydrogen content at the HRT temperature, If C H t is more than C H d in Eqs 3 and 4, C H d is equal to TSSd on of the cladding material, If C H t is less than C H d in Eqs 3 and 4, C H d is equal to C H t, Fne 40 equals Fn 40 and Fle 45 equals Fl 45, TSSd was calculated using the equation reported by Ogata et al. 14, Fn 40 0 :Fn 40 for as-irradiated specimen, and Fl 45 0 :Fl 45 for as-irradiated specimen. In Fne 40 and Fle 45, the difference in the amount of dissolved hydrogen at each HRT temperature is normalized using C H t and C H d. Here, Fle 45 is defined by a small modification of the index as FIG. 14 Comparison of reorientation behavior among irradiated PWR cladding materials (HRT 300 C, 30 C/h).

14 664 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM FIG. 15 Metallography of PWR Zry-4 specimens after ring compression test (irradiated PWR 48GWd/t type Zry-4). suggested by Oohama et al. 12. Fne 40 was defined by Eq 3 to describe the trend of the temperature and stress dependence of reorientation from the view point of the number of radial hydride. Fn 40 has been applied widely as the hydride orientation index, and Fne 40 could be another index considering the amount of dissolved hydrogen at each HRT temperature, although it is not a strict physical quantitative index. In Fig. 18, the reorientation behavior at relatively low HRT temperatures, such as 300 C, is clearer in both Fne 40 and Fle 45 as compared to that in Fig. 4. Ells suggested an equation which gives the stress and temperature dependence of the fraction R ratioing the number of radial hydrides to the number of circumferential hydrides 15. In the Ells s equation the effect of stress was described as an exponential relationship, and the effect of temperature was described as an Arrhenius type relation. The authors of this paper attempted to discuss the temperature and stress dependences of hydride reorientation, based on Eqs 5 and 6 below, which were defined considering the Ells s equation 15, R Fne = Fne 40 1 Fne 40, 5 FIG. 16 Correlation between the ductility of the specimens and the HRT conditions for irradiated PWR 48GWd/t type Zry-4 cladding.

15 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 665 FIG. 17 Effects of HRT cooling rate on ductility in ring compression deformation for irradiated PWR Zry-4 cladding. R Fne = A exp B T, 6 where: : hoop stress MPa, T: temperature K, and A, B: constants. The logarithm of R Fne is plotted versus hoop stress in Fig. 19 a. It shows that relatively good linearity is observed for the 400 C HRT specimens, but the reorientation behavior of the 300 C HRT specimens does not show a monotonic increase with hoop stress. Considering the trend of Fl 45 in the outer side cladding area illustrated in Fig. 19 b, the stress dependence of reorientation of the 300 C HRT specimens in Fig. 19 a can be interpreted as illustrated in Fig. 19 c. As indicated in Figs. 3 b, 3 g, and 19 b, a difference exists in the reorientation behavior between the inner and outer surfaces as suggested in Fig. 19 c. In the cooling stage of the relatively moderate HRT conditions such as 300 C, 70 MPa, new hydrides precipitate in the radial direction in the Zry-2 inner area near the Zr liner where the amount of FIG. 18 Reorientation behavior in dissolved hydrogen at HRT temperature for irradiated BWR Zry-2 cladding with liner (HRT 30 C/ h).

16 666 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM FIG. 19 Stress dependence of reorientation for irradiated BWR Zry-2 cladding with liner (HRT 30 C/h). hydride is sufficiently small that the hydride content completely dissolves at the HRT temperature. On the other hand, new hydrides precipitate in the circumferential direction in the area near the outer surface where some circumferential hydrides remain precipitated at the HRT temperature. Under relatively moderate hoop stress conditions it is hypothesized preferential hydride precipitation occurs along the precipitated not dissolved at the HRT temperature hydrides due to the effects of local stress/strain caused by the precipitated hydride. Under high hoop stress HRT condition such as 100 MPa, it is suggested the effect of the hoop stress surpasses the effect of the pre-existing precipitation explained above, and the new hydride then precipitates in the radial direction near the outer surface area. The logarithm of R Fne is plotted versus 1/T in Fig. 20 based on Eq 6. Here, the HRT temperature was used as T in Fig. 20 a, and the hydride precipitation temperature calculated from TSSp 14 and the hydrogen content of the specimen was used as T in Fig. 20 b. The estimated T precipitate is considered sufficient to discuss the relative tendency, although it is recognized the value of T precipitate has some degree of error due to the difficulty in the local hydrogen content evaluation. In Fig. 20 b, T precipitate and R Fne have little correlation. On the other hand, T HRT and R Fne have significant correlation in Fig. 20 a, at the 100 MPa HRT condition. As for the effect of temperature on reorientation, besides the relation suggested by Ells et al. 15, the temperature dependence of the material strength 4 and the memory effect 3,7 have been reported as significant factors. The memory effect is related to the effect hydride precipitation prior to dissolution during the HRT has on subsequent precipitation. It is supposed that new hydrides tend to precipitate easier on sites with local strain or dislocations caused by precipitated hydride before dissolution. In addition, the existence of precipitated hydride at the HRT temperature is considered to affect the hydride precipitation as mentioned above. Figure 21 gives some indication of the effects of temperature. Figures 21 a 21 c show the metallography of as-irradiated material, post-hrt material near the center position in the longitudinal direction far from the end plug weld, and post-hrt material near the end plug position, respectively. The HRT condition is A in Table 2 c. Some decrease in Vickers hardness is observed in the position corresponding to Fig. 21 c, probably due to irradiation hardening recovery as a result of heat input in the end plug weld 208 Hv for Fig. 21 c compared to 276 Hv for Fig. 21 b. There were, however, neither metal microstructure changes such as grain growth nor end plug binding effects on the stress condition. In the Fig. 21 c position, the dislocations due to hydrides precipitated in the as-irradiated condition were likely recovered by the high temperature annealing during the end plug welding. In addition, the fine hydrides precipitated during the rapid cool after welding were considered not to have a large memory effect in the

17 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 667 FIG. 20 Temperature dependence of reorientation in dissolved hydrogen for BWR Zry-2 cladding with liner (HRT 30 C/h). HRT. The circumferential hydrides near the outer surface were observed in the as-irradiated material in Fig. 21 a. The same type of circumferential hydrides near the outer surface were also observed in Fig. 21 b, but few circumferential hydrides were observed in Fig. 21 c. The difference between Figs. 21 b and 21 c in the existence of hydride with the same type morphology as those shown in the as-irradiated specimen may indicate a memory effect at the 300 C HRT. An effect of temperature is not so clear in Fig. 20, and it is suggested that some effects of temperature, including the precipitated not dissolved hydride at the HRT solution temperature and the memory effect, are mixed in at the relatively moderate HRT conditions. This experiment also revealed the different reorientation behavior among cladding materials. As indicated in the literature 16, the difference in reorientation behavior between Zry-2 RXA and PWR cladding materials SRA is attributed to the effect of heat treatment, which affects the material properties such as the grain structure including grain size, grain boundary orientation, the internal stress/strain, and so on. As for the difference in reorientation behavior among PWR cladding materials SRA, it is supposed that the factor which caused the large reorientation for ZIRLO cladding was not the texture represented by the Fr value, because the Fr for ZIRLO cladding was the same as that for 48GWd/t type Zry-4. The effect of alloying element seemed not to be the dominant factor because, for example, the Zry-4 and MDA alloy containing Nb showed relatively similar behavior, although ZIRLO containing Nb showed a different behavior. It was indicated, however, that the annealing temperature in the tube production process for FIG. 21 Effects of heat treatment before HRT on reorientation for irradiated BWR 55GWd/t type cladding with liner [HRT 300 C, 100 MPa (decreasing with temp.), 3 C/h).

18 668 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM ZIRLO cladding was lower than that for Zry-4 cladding 17. Hence, properties which are affected by annealing temperature, grain size for example 18, might be the possible factors. In this report, the effects of stress, temperature, and cooling rate on reorientation were discussed for each material. In addition, it is suggested that the morphology of precipitated hydride before and during the HRT is one factor which affects reorientation in relatively moderate conditions. The comparison of reorientation behavior between unirradiated and irradiated specimens is shown in Figs. 3 and 20, but it is difficult to discuss the effect of irradiation because the initial hydride amount and distribution for unirradiated specimens were not the same as those for irradiated specimens. An additional investigation is desirable about the effects of the material properties such as grain structure, internal stress/strain, alloying element, SPP, and the irradiation defects. Fracture Behavior in Ring Compression Test In Fig. 22, the crosshead displacement ratio values are plotted versus each index concerning hydride morphology. In the results for PWR cladding, the crack initiated at the inner side area of the wall where the circumferential tensile stress was loaded at the early stage of ring compression. The fracture near the inner surface seems to be caused by shear, which indicates that the metal matrix has some ductility. It is deduced that the crack initiated along the radial hydrides near the inner side area and then propagated to join other crack or to the inner surface at the stage that the first load drop was observed. Therefore, it is suggested changes of orientation, amount, and the length of hydride in the midwall near the inner side area is responsible for the HRT affect on the ductility in ring compression deformation as indicated in Figs. 22 a 22 c. Especially, the relatively good correlation between the maximum length of radial hydride and the ductility as shown in Fig. 22 c as compared to Figs. 22 a or 22 b. This indicates that long cracks caused by long hydride brittle fracture occurs easier than cracks caused by small hydride crack connection because crack connection is not easy due to the relatively sparse hydride distribution in the somewhat ductile metal matrix. In this report, the post-hrt ductility of both ZIRLO and MDA was not evaluated. It is desirable to accumulate mechanical property data for these new alloys and discuss them from the view point of the effects of hydride and the bulk metal properties in the next stage of the investigation. For BWR Zry-2 cladding with a Zr liner, it is supposed that the cracks initiated along the radial hydrides near the outer surface area, and then the cracks propagate to join other cracks or to the outer surface at the first drop of the load detected in the ring compression test. From Figs. 22 d and 22 e, it seems that the ratio and the amount of radial hydride in the whole Zry-2 area do not have clear a effect on the ductility in ring compression deformation for the specimens receiving a 300 C or less HRT. One of the reasons for this is the difference in hydride morphology between the inner and outer side area of the cladding as indicated in Fig. 19. In addition, it may be attributed to the close hydride distribution in the outer side area of the BWR cladding with Zr liner. In Fig. 22 f, the correlation between ductility and the Polymax is shown. Here, Polymax is defined in Fig. 22 g as the maximum value among the integrated length of interconnecting hydrides regardless of the orientation measured in the metallography. This reflects the length or continuity of the hydrides. The correlation of Polymax with ductility in Fig. 22 f seems somewhat higher than those of other indexes in Figs. 22 d and 22 e for the 300 and 250 C HRT specimens. It is suggested that crack propagation in a Zry-2 matrix or along circumferential hydrides may occur easier than for PWR specimens due to the relatively closely spaced hydride distribution. In this case the fracture behavior depends more on the initial hydride network than on the radial hydrides. The fracture behavior after a 400 C HRT is explained as follows. After a 400 C, 0 MPa HRT, some degree of Vickers hardness decrease was measured compared to as-irradiated material from 286 to 279 Hv, and little change was measured after a 300 C, 0 MPa HRT from 286 to 287 Hv. Itis hypothesized that the length or the amount of radial hydride has a larger effect on the fracture behavior after a 400 C HRT as compared to that after a 300 C or less HRT as shown in Fig. 22 e because crack propagation in the metal matrix is suppressed due to irradiation hardening recovery.

19 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 669 FIG. 22 Correlation between the ductility of the specimens and the factors concerning hydride morphology for irradiated cladding [(a), (b), and (c) for irradiated PWR 48GWd/t type Zry-4. (d), (e), and (f) for irradiated BWR 50GWd/t type Zry-2.] [The definition of horizontal axis in (b) and (e) is the same as that in Fig. 5(b) for radial hydrides. The definition of horizontal axis in (c) is the same as that in Fig. 13(b). The definition of horizontal axis in (f) is described in the text and (g).]

20 670 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM Conclusions Test results in this study have led to the following conclusions about hydride stress reorientation behavior and mechanical properties for irradiated BWR and PWR fuel cladding tubes. Irradiated BWR Zry-2 Cladding Hydride reorientation to the radial direction occurred at a relatively low HRT hoop stress, such as less than 70 MPa. The increase of reorientation with hoop stress was not monotonic for specimens in which a part of the hydrides remained precipitated at the HRT temperature, such as the case for 50GWd/t type cladding at a 300 C HRT. The degree of reorientation under relatively moderate HRT conditions depends on the HRT solution temperature rather than on the estimated temperature at which the hydride precipitation occurs. It is suggested that reorientation behavior under relatively moderate HRT conditions such as 300 C, 70 MPa, 30 C/h is a complex function of several effects, including the effects of precipitated not dissolved at the HRT temp. hydride and the memory effect. Under a relatively low cooling rate HRT, preferential hydride precipitation in the Zr liner increased for Zr lined cladding compared to that at a relatively high cooling rate. This phenomenon is easily understood as the results of hydrogen migration from the Zry-2 matrix to the Zr liner due to the TSSp difference between Zr and Zry-2. The ductility of specimens after a 300 C HRT showed relatively good correlation to the Polymax index which reflects the length or continuity of hydrides regardless of their orientation. The ductility of specimens after a 400 C, 0 MPa, 30 C/ h HRT increased in ring compression testing at room temperature as compared to the no HRT as-irradiated specimens. It is hypothesized that the recovery of irradiation damage occurred at the 400 C anneal and affected the ductility of the irradiated Zry-2 cladding. Irradiated PWR Zry-4 and Improved Zirconium Alloy Cladding Little increase in the radial hydride ratio occurred following a 100 MPa, 340 C, or less. HRT. On the other hand, the amount and the length of hydride in the midwall area of the cladding depended on the temperature and cooling rate in the HRT due to the hydrogen migration occurring from the hydride rim area. It is deduced that ductility in ring compression deformation was affected by the orientation, amount, and length of hydride in the midwall area. Acknowledgments The writers are grateful to Professor S. Iwata and the members of the Committee of Long-Term Fuel Safety, who gave the writers useful advice and comments about the plans, implementation and evaluation of this testing. References Hardie, D. and Shanahan., M. W., Stress Reorientation of Hydrides in Zirconium-2.5 % Niobium, J. Nucl. Mater., Vol. 55, 1975, pp Mishima, Y. and Okubo, T., Effect of Thermal Cycling on the Stress Orientation and Circumferential Ductility in Zircaloy-2, Can. Metall. Q., Vol. 11, 1972, pp Northwood, D. O. and Kosasih, U., Hydrides and Delayed Hydrogen Cracking in Zirconium and Its Alloys, Int. Metall. Rev., Vol. 28, 1983, pp Singh, R. N., Kishore, R., Singh, S. S., Sinha, T. K., and Kashyap, B. P., Stress-Reorientation of Hydrides and Hydride Embrittlement of Zr-2.5 wt % Nb Pressure Tube Alloy, J. Nucl. Mater., Vol. 325, 2004, pp Chung, H. M., Understanding Hydride- and Hydrogen-Related Processes in High-Burnup Cladding in Spent-Fuel-Storage and Accident Situations, Proceedings of the 2004 International Meeting on

21 AOMI ET AL. ON HYDRIDE STRESS REORIENTATION AND IRRADIATED CLADDING TUBES 671 LWR Fuel Performance, 2004, pp Daum, R. S., Majumdar, S., Liu, Y., and Billone, M. C. Mechanical Testing of High-Burnup Zircaloy-4 Fuel Cladding Under Conditions Relevant to Drying Operations and Dry-Cask Storage, Proceedings of the 2005 Water Reactor Fuel Performance Meeting, 2005, pp Sakamoto, K. and Nakatsuka, M., Stress Reorientation of Hydrides in Recrystallized Zircaloy-2 Sheet, Proceedings of the 2005 Water Reactor Fuel Performance Meeting, 2005, pp Chung, H. M., Daum, R. S., Hiller, J. M., and Billone, M. C., Characteristics of Hydride Precipitation and Reorientation in Spent-Fuel Cladding, Zirconium in the Nuclear Industry: Thirteenth International Symposium, ASTM STP 1423, ASTM International, West Conshohocken, PA, 2002, pp Marshall, R. P. and Louthan, Jr., M. R., Tensile Properties of Zircaloy with Oriented Hydrides, Trans. ASM, Vol. 56, 1963, pp Yagnik, S. K., Kuo, R.-C., Rashid, Y. R., Machiels, A. J., and Yang, R. L., Effect of Hydrides on the Mechanical Properties of Zircaloy-4, Proceedings of the 2004 International Meeting on LWR Fuel Performance, 2004, pp Sugiyama, T., Nagase, F., and Fuketa, T., Modification of Ring Tensile Test for LWR Fuel Cladding, Proceedings of the 2005 Water Reactor Fuel Performance Meeting, 2005, pp Oohama, T., Okunishi, M., Senda, Y., Murakami, K., and Sugano, M., Study on Hydride Re- Orientation Properties in Zircaloy-4 Cladding Tube, The 6th International Conference on Nuclear Thermal Hydraulics, Operations and Safety NUTHOS-6, Nara, Une, K. and Ishimoto, S., Terminal Solid Solubility of Hydrogen in Unalloyed Zirconium by Differential Scanning Calorimetry, J. Nucl. Sci. Technol., Vol. 41, 2004, pp Ogata, K., Baba, T., Kamimura, K., Kubo, T., Une, K., and Etoh, Y., Study on the Hydrogen Assisted Cracking of Zircaloy Fuel Cladding at High Burnup, 15th International Symposium on Zirconium in the Nuclear Industry, Sunriver, June 2007, pp Ells, C. E., The stress orientation of hydride in zirconium alloys, J. Nucl. Mater., Vol. 35, 1970, pp Bai, J. B., et al. Effect of Microstructure Factors and Cold Work on the Hydride Precipitation in Zircaloy-4 Sheet, J. Adv. Sci., Vol. 3, 1991, p Sabol, G. P., In-Reactor Corrosion Performance of ZIRLO and Zircaloy-4, Zirconium in the Nuclear Industry: Tenth International Symposium, ASTM STP 1245, ASTM International, West Conshohocken, PA, 1994, pp Kearns, J. J., et al., Effect of Texture, Grain Size and Coldwork on the Precipitation of Oriented Hydrides in Zircaloy Tubing and Plate, J. Nucl. Mater., Vol. 20, 1966, p. 241.

22 672 ZIRCONIUM IN THE NUCLEAR INDUSTRY: 15TH SYMPOSIUM DISCUSSION Question 1, Paul Cantonwine, GNF, USA: Your threshold stresses were determined from constant stress tests. How might your results change if the stress decreased linearly with temperature? Authors Reply: The degree of reorientation increases with the stress. Therefore it is qualitatively expected that the degree of reorientation decreases in decreasing stress conditions. In our test, constant stress conditions were applied in order to acquire the correlation between the degree of reorientation and the stress. Question 2, Djamshid Khatamian, AECL, Canada: What is the reason for the preferential accumulation of hydrogen in the Zr liner? Authors Reply: As reported in the literature, the terminal solid solubility TSS of Zr is smaller than that of Zircaloy. The hydride precipitation in the Zr liner occurs prior to the precipitation in the Zircaloy in the cooling stage of the HRT due to the TSS difference between Zr and Zircaloy. This is supported by the simulation using a calculation code based on the diffusion equation. Question 3, Bruce Kammenzind, Bettis Laboratory, USA: You show ZIRLO to have a higher propensity for hydride re-orientation in the radial direction than Zircaloy. Do you have any insights for this? Are these due to differences in texture or degree of recrystallization? Authors Reply: Unfortunately, we do not have a good insight about this. As reported in the literature, reorientation susceptibility depends on the microstructure including grain size, grain boundary orientation and so on. The difference in heat treatment between ZIRLO and Zircaloy may result in a difference in the microstructure. Question 4, Young Suk Kim, KAERI, Korea: Your critical stress for the reorientation of hydrides, 28MPa is lower than other reported values. Could you comment on the cause of your lower critical stresses for the reorientation? When 28MPa was applied, this stress must be elastic, and not cause any plastic deformation. How could you explain that hydride precipitation is affected in the elastic stress range? Authors Reply: From our test results, the hydride reorientation for BWR cladding increased in relatively low hoop stress conditions. The literature data for SRA cladding suggest a larger threshold stress values than our results for RXA cladding. But, considering the literature data for RXA Zircaloy and the HRT conditions, our test results are not that strange. The hydride reorientation occurs in the elastic stress range because it is considered that the tensile stress affects the free energy change in hydride precipitation with volume expansion. Question 5, John Foster, Westinghouse, USA: What are the PWR Zry-4, MDA, and ZIRLO Fne values? Authors Reply: The Fn 40 for Zry-4 48GWd/t type, MDA, and ZIRLO are 0.21, 0.38, and 0.89 respectively. The hydrides in the cladding accumulated in the hydride rim area before HRT, thus the hydrogen content in the area except for the hydride rim is evaluated to be sufficiently small that almost all the hydrogen dissolved at the HRT temperature. Therefore the Fne 40 is almost equivalent to the Fn 40. The reorientation behavior for PWR materials was evaluated from Fl 45 in our study, because the number of hydrides was not large and it was suggested the error in Fl 45 is smaller than Fn 40. Question 6, Karl-Fredrik Nilsson, Institute For Energy Nuclear, Netherlands: Have you considered modeling the variation in ductility versus reorientation Fn 45? This could for instance be done by probabilistic fracture mechanics given that you have orientation and length. Authors Reply: We have not modeled the correlation between ductility and hydride morphology.