Author: R. A. Holt
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

Get Book Here

Book Description

Author: M. Griffiths
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

Get Book Here

Book Description

Author:
Publisher:
ISBN:
Category : Science
Languages : en
Pages : 1120

Get Book Here

Book Description

Author: A. Soniak
Publisher:
ISBN:
Category : Creep
Languages : en
Pages : 23

Get Book Here

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: R. A. Holt
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

Get Book Here

Book Description

Author: T. Andersson
Publisher:
ISBN:
Category : Congress
Languages : en
Pages : 21

Get Book Here

Book Description
The creep behavior of Zircaloy cladding materials depends on materials texture, degree of recrystallization, and chemical composition. This study is devoted mainly to the analysis of the effect of the final annealing (i.e., the degree of recrystallization) on the creep characteristics. For this purpose, data from a series of thermal creep tests are presented and evaluated. In addition, the in-reactor creep data presented by Franklin et al. are used to evaluate the effect of irradiation on cladding creep performance. The out-of-reactor tests are performed under internal pressurization, and the test matrix covers seven conditions with temperatures from 330 to 400°C and hoop stresses between 80 and 160 MPa. Three lots of Zircaloy-2 claddings and one lot of Zircaloy-4 are considered. The difference between the three Zircaloy-2 lots is in their final annealing conditions. The claddings are either stress relief annealed (SRA), recrystallization annealed (RXA), or partially recrystallization annealed (PRXA). The materials used when fabricating the Zircaloy-2 claddings are from the same ingot, and the chemical compositions of the three types of claddings are almost identical. The Zircaloy-4 cladding included in the test is SRA, and the tin content in this material is similar to that in the Zircaloy-2 materials. The creep data are analyzed by separating the primary (transient) and the secondary (steady-state) creep. In this analysis, the Matsuo creep model, which accounts for both primary and secondary creep, is modified, calibrated, and verified using the new thermal creep data. Based on in-reactor data, the thermal creep model is extended to cover also the creep behavior under irradiation. The claddings considered in the in-reactor test were of both SRA and RXA types, and the experiments were made under external pressure. It is observed that for moderate hoop stresses (

Author: Robert Odette
Publisher: Newnes
ISBN: 012397349X
Category : Technology & Engineering
Languages : en
Pages : 673

Get Book Here

Book Description
High-performance alloys that can withstand operation in hazardous nuclear environments are critical to presentday in-service reactor support and maintenance and are foundational for reactor concepts of the future. With commercial nuclear energy vendors and operators facing the retirement of staff during the coming decades, much of the scholarly knowledge of nuclear materials pursuant to appropriate, impactful, and safe usage is at risk. Led by the multi-award winning editorial team of G. Robert Odette (UCSB) and Steven J. Zinkle (UTK/ORNL) and with contributions from leaders of each alloy discipline, Structural Alloys for Nuclear Energy Applications aids the next generation of researchers and industry staff developing and maintaining steels, nickel-base alloys, zirconium alloys, and other structural alloys in nuclear energy applications. This authoritative reference is a critical acquisition for institutions and individuals seeking state-of-the-art knowledge aided by the editors’ unique personal insight from decades of frontline research, engineering and management. Focuses on in-service irradiation, thermal, mechanical, and chemical performance capabilities. Covers the use of steels and other structural alloys in current fission technology, leading edge Generation-IV fission reactors, and future fusion power reactors. Provides a critical and comprehensive review of the state-of-the-art experimental knowledge base of reactor materials, for applications ranging from engineering safety and lifetime assessments to supporting the development of advanced computational models.

Author: R. A. Holt
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

Get Book Here

Book Description

Author: R. A. Holt
Publisher:
ISBN:
Category :
Languages : en
Pages : 0

Get Book Here

Book Description

Author: George P. Sabol
Publisher: ASTM International
ISBN: 0803124066
Category : Nuclear fuel claddings
Languages : en
Pages : 907

Get Book Here

Book Description