Effect of Hydrogen on Irradiation Creep and Growth for ZIRLO Alloy and Zr-1.0Nb PDF Download
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Author: John Paul Foster
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 23
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Book Description
The impact of hydrogen on the irradiation growth and creep of stress-relief annealed (SRA) ZIRLO® alloy and recrystallized annealed (RXA) Zr-1.0Nb cladding tubes is evaluated in this paper. The samples were charged with hydrogen in the range of approximately 160-720 ppm using the gaseous method. Biaxial in-reactor creep tests were performed after Cycle 1, Cycle 2, Cycle 3, and Cycle 4, on both as-received and precharged hydrogen cladding tubes. Outside diameter and axial length measurements were performed on the samples. The results showed that hydrogen had no effect on the axial irradiation creep but a relatively large effect on the axial irradiation growth. Increasing hydrogen decreased the axial irradiation growth in RXA Zr-1.0 Nb, which was opposite from the behavior of SRA ZIRLO cladding. This unique hydrogen effect on the irradiation axial growth of RXA Zr-1.0Nb could be due to the different intergranular stress resulting from the fabrication process or the absence of alloying elements, including tin. The total axial strain of both Zr-1.0Nb and SRA ZIRLO cladding increased with increasing fast fluence, and Zr-1.0Nb increased at a faster rate relative to SRA ZIRLO cladding. In the diameter direction, hydrogen had a minimal effect on the total diameter strain and the diameter irradiation creep strain for both SRA ZIRLO samples and the RXA Zr-1.0Nb sample. This finding from in-reactor test is contrary to the out-reactor tests results from the literature that have shown that hydrogen significantly decreases thermal creep. The total diameter strain and diameter irradiation creep behavior for the SRA ZIRLO samples and RXA Zr-1.0Nb were similar.
Author: John Paul Foster
Publisher:
ISBN:
Category : Irradiation
Languages : en
Pages : 23
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Book Description
The impact of hydrogen on the irradiation growth and creep of stress-relief annealed (SRA) ZIRLO® alloy and recrystallized annealed (RXA) Zr-1.0Nb cladding tubes is evaluated in this paper. The samples were charged with hydrogen in the range of approximately 160-720 ppm using the gaseous method. Biaxial in-reactor creep tests were performed after Cycle 1, Cycle 2, Cycle 3, and Cycle 4, on both as-received and precharged hydrogen cladding tubes. Outside diameter and axial length measurements were performed on the samples. The results showed that hydrogen had no effect on the axial irradiation creep but a relatively large effect on the axial irradiation growth. Increasing hydrogen decreased the axial irradiation growth in RXA Zr-1.0 Nb, which was opposite from the behavior of SRA ZIRLO cladding. This unique hydrogen effect on the irradiation axial growth of RXA Zr-1.0Nb could be due to the different intergranular stress resulting from the fabrication process or the absence of alloying elements, including tin. The total axial strain of both Zr-1.0Nb and SRA ZIRLO cladding increased with increasing fast fluence, and Zr-1.0Nb increased at a faster rate relative to SRA ZIRLO cladding. In the diameter direction, hydrogen had a minimal effect on the total diameter strain and the diameter irradiation creep strain for both SRA ZIRLO samples and the RXA Zr-1.0Nb sample. This finding from in-reactor test is contrary to the out-reactor tests results from the literature that have shown that hydrogen significantly decreases thermal creep. The total diameter strain and diameter irradiation creep behavior for the SRA ZIRLO samples and RXA Zr-1.0Nb were similar.
Author:
Publisher: ASTM International
ISBN:
Category :
Languages : en
Pages : 308
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Book Description
Author: Suresh Yagnik
Publisher:
ISBN:
Category : Nuclear reactors
Languages : en
Pages : 48
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Book Description
In-reactor dimensional changes in zirconium-based alloys result from a complex interplay of many factors, such as (1) alloy type and composition, including the addition of elements such as niobium, iron, and tin; (2) fabrication process, including cold work, texture, and residual stresses; (3) irradiation temperature; and (4) hydrogen levels. In many cases, the observed dimensional changes in light water reactor fuel-assembly components--especially at high exposures--cannot be fully explained based on current growth and creep models. Therefore, a systematic approach was taken in this multiyear (2005-2011) Nuclear Fuel Industry Research Program investigation. The objective was to measure stress-free irradiation-induced growth (IIG) of specially fabricated alloys through irradiation under controlled conditions in the BOR-60 fast-flux test reactor up to a high fluence of approximately 2 x 1026 m-2 (E > 1 MeV)--equivalent to maximum of approximately 37 dpa exposure--followed by postirradiation examinations (PIEs). Irradiation temperature was within a narrow temperature range (320 ± 10°C). The PIEs included dimensional-change and microhardness measurements, metallography and hydride etching, and scanning transmission electron microscopy (STEM) or transmission electron microscopy (TEM).
Author: D. G. Franklin
Publisher: ASTM International
ISBN: 9780803102590
Category : Science
Languages : en
Pages : 322
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Book Description
Author: R. A. Holt
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Book Description
Author: Holt, R. A
Publisher: [London] : Pergamon Press
ISBN:
Category :
Languages : en
Pages : 10
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Book Description
Author: E. F. Ibrahim
Publisher:
ISBN:
Category :
Languages : en
Pages : 0
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Book Description
Author: A. Soniak
Publisher:
ISBN:
Category : Creep
Languages : en
Pages : 23
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Book Description
This paper deals with the irradiation-induced changes in the microstructure of FRAMATOME advanced Zr-base alloys and the correlations with their irradiation creep and growth behavior. The first part is dedicated to experimental irradiations performed at 280 and 350°C in a CEA metallurgical test reactor (Siloé, 1 to 1018 n/m2 s, E > 1 MeV) on stress-relieved (SRA) and recrystallized (RXA) low-tin Zircaloy-4 and two advanced RXA materials (Alloy 4 (M4): Zr-SnFeV, and Alloy 5 (M5): Zr-NbO), which are proposed for fuel rod cladding applications in PWRs. The irradiation creep results confirm the improved behavior of RXA Zy-4, M4, and M5 in comparison to that of SRA Zy-4. Similar trends are observed for irradiation growth, the SRA Zy-4 exhibiting a quasi-linear behavior with increasing fluence while RXA alloys undergo an early saturation phenomenon. Among RXA materials, M5 has the higher irradiation growth resistance. These creep and growth results at moderate neutron fluences (
Author: D. Faulkner
Publisher:
ISBN:
Category : Damage rate
Languages : en
Pages : 17
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Book Description
The irradiation creep and growth behavior of zirconium alloys has been studied during irradiation with 3.5-MeV protons. Irradiations were carried out at temperatures in the range 423 to 623 K and strain measurements were recorded up to displacement levels of 0.03 displacements per atom (dpa). In annealed materials, a significant portion of the measured strain could be attributed to the presence of dislocation loops. The measured growth strain was found to be dependent on texture, grain dimensions, network dislocation structure (cold work), and temperature. Experiments to separate the irradiation creep and growth components of the total strain revealed that irradiation growth was by far the most significant component in cold-worked zirconium-niobium alloys but that the two components were approximately equal in annealed crystal bar zirconium specimens. An investigation of transient effects revealed that no strain transient was observed when the irradiation flux was removed. The strain rate was found to be proportional to the applied stress (at low stresses) and to the damage rate.
Author: Manfred P. Puls
Publisher: Springer Science & Business Media
ISBN: 1447141954
Category : Science
Languages : en
Pages : 475
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Book Description
By drawing together the current theoretical and experimental understanding of the phenomena of delayed hydride cracking (DHC) in zirconium alloys, The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Components: Delayed Hydride Cracking provides a detailed explanation focusing on the properties of hydrogen and hydrides in these alloys. Whilst the emphasis lies on zirconium alloys, the combination of both the empirical and mechanistic approaches creates a solid understanding that can also be applied to other hydride forming metals. This up-to-date reference focuses on documented research surrounding DHC, including current methodologies for design and assessment of the results of periodic in-service inspections of pressure tubes in nuclear reactors. Emphasis is placed on showing how our understanding of DHC is supported by progress in general understanding of such broad fields as the study of hysteresis associated with first order phase transformations, phase relationships in coherent crystalline metallic solids, the physics of point and line defects, diffusion of substitutional and interstitial atoms in crystalline solids, and continuum fracture and solid mechanics. Furthermore, an account of current methodologies is given illustrating how such understanding of hydrogen, hydrides and DHC in zirconium alloys underpins these methodologies for assessments of real life cases in the Canadian nuclear industry. The all-encompassing approach makes The Effect of Hydrogen and Hydrides on the Integrity of Zirconium Alloy Component: Delayed Hydride Cracking an ideal reference source for students, researchers and industry professionals alike.
