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2019 ⦽ǎႊᔍᖒ⠱ʑྜྷ⦺⫭⇹ĥ⦺ᚁݡ⫭םྙ᧞ḲMeasurement of Threshold Stress Intensity Factor of Delayed Hydride Cracking for
Unirradiated Zircaloy-4 Cladding
Jong-Dae Hong*, Euijung Kim, and Donghak Kook
Korea Atomic Energy Research Institute, 111, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, Republic of Korea
*
1. Introduction
During dry storage, the spent fuel claddings could degrade through delayed hydride cracking (DHC), which is a time dependent crack growth process resulting from stress assisted hydrogen diffusion to the crack tip. In particular, DHC could be activated within a relatively low temperature range, creep-free range. Because the estimated crack growth rate by DHC is quite high, the methodology using threshold stress intensity factor criteria is appropriate [1,2]. Nevertheless, most of the DHC study were focused on crack growth rate measurement due to pressure tube of pressurized heavy water reactor (PHWR). In this regards, the temperature dependency on DHC of Zircaloy-4 cladding were investigated by measuring threshold stress intensity factor (KIH) according to
International Atomic Energy Agency Cooperative Research Project (IAEA CRP) method [3].
2. Experimental
2.1 Test Material
The test material used in this study was unirradiated CWSR Zircaloy-4 cladding, which has been widely used in PWR fuel cladding. The initial cladding thickness (t) and outer diameter (OD) are 0.57 mm and 9.5 mm (Westinghouse 17X17 type). In addition, 200 ppm of hydrogen is charged using the gaseous method under mixed argon/hydrogen condition at 400°C, then heat-treated at 410°C for 24 h for a uniform hydride distribution (Fig. 1).
Fig. 1. Hydrided Zircaloy-4 cladding for DHC test.
2.2 Test Method
DHC test to measure KIH was performed using a
pin-loaded-tension (PLT) technique and an unloading method (Fig. 2) at 150°C to 300°C, as shown in Table 1. Test specimens were prepared as PLT-13 type with fatigue pre-cracking at RT. A linear variable differential transformer (LVDT) on the crack mouth was used for crack growth monitoring (Fig. 3). Upon detection of a certain amount of crack growth, the load was reduced by 0.5 kgf and the process repeated until no further cracking was detected for 24 h. After the DHC test, scanning electron microscope (SEM) analysis was performed to measure crack length. In addition, the applied stress intensity factor (KI) and threshold stress
intensity factor (KIH) were calculated according to
the previous research [3].
Table 1. Materials and temperature for DHC test.
Material Temp. (°C)
Unirradiated Zry-4 cladding (CWSR) ~200 ppm H 300 275 250 225 200 175 150
2019 ⦽ǎႊᔍᖒ⠱ʑྜྷ⦺⫭ ⇹ĥ⦺ᚁݡ⫭ םྙ᧞Ḳ
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Fig. 3. DHC test setup using LVDT.
3. Results and Discussion
After tests, the area identification and crack length measurements using 9-point method were performed by scanning electron microscope (SEM) analysis, as described in Fig, 4. A clear chevron pattern without striation in the DHC region was observed along the crack growth direction.
Fig. 4. Area identification and measurement of crack growth by SEM analysis.
The resulting values of KIH from tests using
unirradiated specimen are described in Fig. 5. Considering the data of the IAEA CRP and of this study, KIH increases with a significant increase
around 300°C as the temperature increases [4]. This tendency shows good agreement to a prediction based on Shi and Puls model [5].
Fig. 5. Temperature dependency of threshold stress intensity factor for unirradiated Zircaloy-4 cladding.
They are highly scattered, similar to IAEA CRP, although all of data meets crack uniformity criteria.
The high values of KIH could be contributed to
over-night unloading procedure. During over-over-night, the applied load was reduced to 30 N due to an impracticability of in-situ monitoring. For this reason, a crack could be blunted, then much load is needed to restart crack propagation. Therefore, some of data could be excluded according to an additional acceptance criterion.
4. Conclusion
KIH for unirradiated Zircaloy-4 cladding were
measured using a PLT technique and an unloading method. KIH increases with a significant increase
around 300°C as the temperature increases, although they are highly scattered.
ACKNOWLEGMENTS
This work was supported by the Radioactive Waste Management Technology Program of the Korea Institute of Energy Technology Evaluation and Planning (KETEP), granted financial resource from the Ministry of Trade, Industry & Energy, Republic of Korea. (No. 2014171020166A)
REFERENCES
[1] J.D. Hong, E. Kim, Y.S. Yang, D. Kook, ³Methodology of Delayed Hydride Cracking Assessment of Spent Fuel Cladding´, KRS Spring Meeting, Busan, May 24-26, 2017.
[2] J.D. Hong et al, ³Delayed Hydride Cracking Assessment of PWR Spent Fuel during Dry Storage´, Nucl. Eng. Des. 322 (2017), 324-330. [3] ³(YDOXDWLRQRI&RQGLWLRQVIRU+\GURJHQ,QGXFHG
Degradation of Zirconium Alloys During Fuel 2SHUDWLRQ DQG 6WRUDJH´, IAEA-TECDOC-1781, Vienna (2015).
[4] J.D. Hong et al, Zircaloy-4 Cladding Degradation Tests under Simulated Dry Storage Conditions- Creep and Delayed Hydride Cracking, IHLRWM 2019, Paper ID. 27369.
[5] S.Q. Shi, M.P. Puls ³&ULWHULD IRU IUDFWXUH initiation at hydrides in zirconium alloys-I. Sharp FUDFNWLS´-1XFO0DWHU 232-242.
