Evaluation of Residual Stress Relaxation in a Rolled Joint by Neutron Diffraction

Quantum Beam Science

RonaldB. Rogge¹  JohnH. Root²  Makoto Hayashi³  Pingguang Xu⁴ 

[1] Canadian Neutron Beam Centre, Canadian Nuclear Laboratories, Chalk River, ON K0J 1J0, Canada;Canadian Neutron Beam Centre, National Research Council, Chalk River, ON K0J 1J0, Canada;Comprehensive Research Organization for Science and Society, Tokai, Ibaraki 319-1106, Japan;Materials Sciences Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195, Japan;

关键词: neutron diffraction;    residual stress;    rolled joint;    stainless steel;    Zr alloy;    aging;   

DOI  :  10.3390/qubs2040021

来源: DOAJ

【 摘 要 】

The rolled joint of a pressure tube, consisting of three axial symmetric parts, modified SUS403 stainless steel as an inner extension, Zr–2.5Nb as the pressure tube and an Inconel-718 outer sleeve has been examined by neutron diffraction for residual stresses. It was heat treated to 350 °C for 30, 130 and 635 h to simulate thermal aging over the lifetime of an advanced thermal reactor respectively for 1, 5 and 30 years at an operating temperature of 288 °C. The crystallographic texture has been investigated from cylindric disks cut from the heat treated Zr–2.5Nb pressure tube to determine the proper sample-orientation-dependent hkl reflections for reliable residual strain measurements. Corresponding in situ tensile deformation was carried out to obtain the necessary diffraction elastic constants for the residual stress evaluation. Three-dimensional crystal lattice strains at various locations in the rolled joint before and after the aging treatments for various times were non-destructively measured by neutron diffraction and the residual stress distribution in the rolled joint was evaluated by using the Kröner elastic model and the generalized Hooke’s law. In the crimp region of the rolled joint, it was found that the aging treatment had a much weaker effect on the residual stresses in the Inconel outer sleeve and the modified SUS403 stainless steel extension. In the non-aged Zr–2.5Nb pressure tube, the highest residual stresses were found near its interface with the modified SUS430 stainless steel extension. In the crimp region of the Zr–2.5Nb pressure tube near its interface with the modified SUS430 stainless steel, the average compressive axial stress was −440 MPa, having no evident change during the long-time aging. In the Zr–2.5Nb pressure tube outside closest to the crimp region, the tensile axial and hoop stresses were relieved during the 30 h of aging. The hoop stresses in the crimp region evolved from an average tensile stress of 80 MPa to an average compressive stress of 230 MPa after the 635 h of aging, suggesting that the rolled joint had a good long-term sealing ability against leakage of high temperature water. In the Zr–2.5Nb pressure tube close to the reactor core and far away from the modified SUS403 stainless steel extension, the residual stresses near the inside surface of the pressure tube were almost zero, helping to keep a good neutron irradiation resistance.

【 授权许可】

Unknown