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

Aomi, Masaki; Baba, Tetsuya; Miyashita, Toshiyasu; Kamimura, Kazuo; Yasuda, Takayoshi; Shinohara, Yasunari; Takeda, Takeshi

doi:10.1520/stp48161s

Cited by 12 publications

(18 citation statements)

References 10 publications

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“…The absence of radial hydrides is a demonstration that the hoop stress produced by inner pressure added to any other circumferential stress (if any) do not overpass the threshold to produce hydride reorientation; this is in agreement with the results obtained by Aomi et al [10]. Blisters in samples 5.5-Al-a and 5.5-Al-b were grown heating until 420°C.…”

Section: Resultssupporting

confidence: 89%

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Effect of blisters and rims on radial hydride precipitation in fuel elements under dry storage conditions

et al. 2018

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Hydrogen enters in Zry clads during fuel element life in service; when hydrogen exceeds the terminal solid solubility, circumferential hydrides are precipitated. In the presence of thermal gradients, produced for instance by oxide spallation, localized concentration of hydrides -known as blisters -can be formed. Due to the fact that zirconium hydrides have a lower density that zirconium alloys, stresses are produced around blisters. These stresses may overcome the necessary threshold to precipitate radial hydrides. Moreover, cracks inside blisters may act as initiators of delayed hydride cracking (DHC). In the dry storage facilities developed for Atucha 1, Zry -4 clads of spent fuel elements (SFE) must withstand environmental conditions which include maximum temperature of 200°C and internal SFE pressure of ~ 75 bar (value at end of cycle). During dry storage, if blisters are present, stresses around them are superposed to circumferential stresses generated by the internal pressure due to fission gasses helping precipitation and/or growth of radial hydrides and DHC process. In order to assess this possibility, in the present work sections of Zry-4 cladding were hydrided. Blisters have been grown by thermal gradient under different conditions of temperature and shapes of thermal contact in order to obtain different blister dimensions and fraction of radial hydrides. After blister formation, some specimens were subjected to Hydride Reorientation Test (HRT) by applying an internal pressure of 75 bar at 200°C. Finally, tubes were sectioned and metallographically analyzed looking for radial hydrides. In some samples radial hydrides were observed, which effectively grew by the internal pressure of Zrycladding during HRT.

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Section: Resultssupporting

confidence: 89%

“…VALANCE, et al [9] reported 50 MPa for irradiated Zry-2 and 65-74 MPa for the same unirradiated material. Values above 100 MPa have been reported in Zry-4 [10] and less than 70 MPa in Zry-2 cladding of SFE [10]. Finally, the radial hydrides fraction increases inversely with cooling speed.…”

Section: Hydride Embrittlement (He)mentioning

confidence: 90%

Effect of blisters and rims on radial hydride precipitation in fuel elements under dry storage conditions

et al. 2018

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“…In addition to the Argonne datasets for PWR cladding alloys, Aomi et al [8] have generated data for Zry-2 and Zry-4 by using test methods similar to the ones used by Argonne. However, RHT samples were cooled under constant stress (vs. decreasing stress, as used by Argonne), and RCTs were conducted only at room temperature (RT) and a slow displacement rate (0.033 vs. 5 mm/s at Argonne).…”

Section: Baseline Studies For Ring Compression Testing Of High-burnupmentioning

confidence: 99%

Used Fuel Disposition Campaign - Baseline Studies for Ring Compression Testing of High-Burnup Fuel Cladding

Billone

Burtseva

Liu

et al. 2012

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SUMMARYStructural analyses of high-burnup fuel require cladding mechanical properties and failure limits to assess fuel behavior during long-term dry cask storage and transportation. Pre-storage drying-transfer operations and early stage storage subject cladding to higher temperatures and much higher pressure-induced tensile hoop stresses relative to in-reactor operation and pool storage. Under these conditions, radial hydrides may precipitate during slow cooling and provide an additional embrittlement mechanism as the cladding temperature decreases below the ductile-to-brittle transition temperature (DBTT). On the basis of previous test results, susceptibility to radial-hydride precipitation depends on cladding material, microstructure, and pre-drying distribution of hydrides across the cladding wall, as well as peak hoop stresses and temperatures during drying operations and storage. Susceptibility to embrittlement depends on the extent of radial-hydride precipitation and the thickness of the outer-surface hydride rim.Consistent with the Argonne Test Plan (December 31, 2011), baseline studies were conducted with asirradiated cladding to determine hydrogen distribution and hydride morphology across the cladding wall; strain-rate sensitivity; and temperature sensitivity of high-burnup M5 ® , ZIRLO™, and Zircaloy-4 (Zry-4) cladding subjected to ring compression test (RCT) loading. The results also serve as the baseline for highburnup cladding exposed to drying-storage conditions that do not lead to radial-hydride precipitation.Baseline mechanical properties and failure limits for irradiated M5 ® and ZIRLO™ are particularly important because they are not publicly available. Cladding samples used for baseline studies were from sibling high-burnup rods irradiated to high-burnup in the same assembly as cladding samples used to study radial-hydride-induced embrittlement. RCT displacement rates were 0.05-50 mm/s (1000-fold increase in elastic strain rate). High-burnup M5 ® with <100-wppm hydrogen exhibited high strength (based on maximum load) and high ductility (>10%, based on offset strain) with relatively low strain-rate and temperature (20-90°C) sensitivity. High-burnup ZIRLO™ samples had 530-wppm hydrogen, a welldeveloped hydride rim, and only 140-wppm hydrogen within the inner two-thirds of the cladding wall. This ZIRLO™ also exhibited low strain-rate and temperature sensitivity. Cracking initiated in the hydride rim. Room-temperature (RT) offset strains were 7±1% for the three displacement rates. Offset strains increased from 7% to >10% with the 20-150°C increase in RCT temperature. High-burnup Zry-4 samples had 640-wppm hydrogen, a well-developed hydride rim, and 250-wppm hydrogen within the inner twothirds of the cladding wall. Based on maximum load, the material appeared to have low strain-rate sensitivity. However, offset strains were too low (<2%) to assess strain-rate and temperature (20-90°C) effects on ductility. Cracking within the outer half of the cladding wall occurred concurrently with plastic defor...

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“…사용후핵연료의 건전성에 영향을 미칠 수 있는 요인 들로는 크립(creep) [5][6][7]과 수소화물재배열 [8][9][10], 지연수 소화균열(Delayed Hydride Cracking, DHC) [11][12][13], 응력 부식균열 [14,15], 산화 및 부식에 의한 파괴 등을 들 수 있 다 [16]. 미국 원자력안전규제위원회(U.S. Nuclear Regula- …”

Section: 서 론unclassified

Review on Spent Nuclear Fuel Performance and Degradation Mechanisms under Long-term Dry Storage

Kim¹,

Kook²,

Sim³

et al. 2013

Journal of the Nuclear Fuel Cycle and Waste Technology

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As the capacity of spent nuclear fuel storage pool at reactor sites becomes saturated in ten years, long term dry storage strategy has been recently discussed as an alternative option in Korea. In this study, we reviewed safetycriteria-related research results on spent nuclear fuel performance and integrity under long-term dry storage and proposed the direction and the scope of future domestic research and development. Creep and hydride effect in relation to the embrittlement are known to be the major degradation mechanisms of the spent fuels during the long term dry storage. However, recent research results showed that hydride reorientation and hydride embrittlement are one of the most critical factors to the spent fuel integrity. Accordingly safety criteria of US and Japan for the storage system are basically founded on those mechanisms. However, in Korea, not only in-pile but out-of-pile experimental data have not been generated to understand fuel cladding degradation and to determine the criteria to ensure the safety. In addition, the transient behavior of the spent fuel during transportation also needs to be thoroughly examined. Therefore, various experimental research and development will be required to establish our own safety criteria for future long-term dry storage of domestic spent fuels.

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

Cited by 12 publications

References 10 publications

Effect of blisters and rims on radial hydride precipitation in fuel elements under dry storage conditions

Effect of blisters and rims on radial hydride precipitation in fuel elements under dry storage conditions

Used Fuel Disposition Campaign - Baseline Studies for Ring Compression Testing of High-Burnup Fuel Cladding

Review on Spent Nuclear Fuel Performance and Degradation Mechanisms under Long-term Dry Storage

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