Delamination Mechanism Analysis of Stainless Steel Strip Electroslag Weld Overlay on Nuclear Island Main Equipment

Literature Overview

This study by Yang Chengdong, Ru Xiangkun, and Tang Weibao, published in Pressure Vessel Technology in 2023, investigates the delamination mechanism of stainless steel strip electroslag welding (ESW) overlay on nuclear island main equipment. Conducted by Shanghai Electric Nuclear Power Group Co., Ltd. and the Shanghai Nuclear Power Equipment Welding and Testing Engineering Technology Research Center, the research addresses a critical quality issue in the fabrication of nuclear-grade pressure vessels and components. The work is of particular significance given the stringent safety requirements and quality standards applicable to nuclear island equipment.

Core Technical Content

Strip electroslag welding overlay is a widely used process for applying stainless steel cladding to carbon steel or low-alloy steel pressure vessels in nuclear power plants. The process involves the use of a continuously fed strip electrode and a slag pool to achieve high deposition rates and uniform overlay thickness. However, delamination between the overlay layer and the base metal, or between successive overlay passes, is a well-documented defect that can compromise the integrity of the component.

The study identifies several mechanisms contributing to delamination:

Inclusion entrapment: Slag inclusions at the interface between overlay passes reduce the bond strength and create initiation sites for delamination.
Residual stress accumulation: The sequential deposition of multiple overlay passes generates complex residual stress fields that can exceed the interfacial shear strength.
Microstructural incompatibility: Dissimilar microstructures between the overlay passes, resulting from variations in cooling rate and thermal history, can create weak interfaces.
Hydrogen-induced cracking: Diffusible hydrogen trapped at the interface during welding can accumulate and cause delayed cracking under residual stress.

Technical Parameters and Process Control

Parameter	Typical Value	Delamination Risk Factor
Strip thickness	1.5–3.0 mm	Thicker strips increase slag inclusion risk
Welding current	1500–3000 A	Higher currents increase heat input and slag volume
Slag viscosity	8–15 Pa·s	Low viscosity promotes inclusion removal
Interpass temperature	< 250°C	Elevated temperatures increase hydrogen absorption
Number of overlay passes	3–6	More passes increase cumulative residual stress
Post-weld heat treatment	1050°C, 2 h, water quench	Dissolves intermetallic phases and relieves stress

The study employs a combination of metallographic analysis, scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS), and fracture surface analysis to characterize the delamination mechanism at the microstructural level.

Engineering Practice Integration

The findings of this study have direct implications for the fabrication of nuclear island main equipment, including:

Reactor pressure vessels: The stainless steel overlay layer provides corrosion resistance to the primary coolant while maintaining the structural integrity of the low-alloy steel base material.
Steam generators: Overlay on tube sheets and channel plates to resist corrosion by reactor coolant.
Pressurizers and spares: Overlay on pressure boundaries to extend service life in aggressive chemical environments.

The study recommends the following quality control measures to prevent delamination:

Rigorous slag removal between overlay passes using mechanical grinding or chemical cleaning.
Monitoring of interpass temperature using infrared thermography or embedded thermocouples.
Post-weld non-destructive testing using ultrasonic testing (UT) and phased array ultrasonic testing (PAUT) to detect interfacial defects.
Adherence to qualified welding procedures in accordance with NB/T 47014 and ASME IX.

Key Questions and Reflections

The delamination mechanism identified in this study raises important questions about the long-term integrity of ESW overlay layers under cyclic loading and radiation exposure in nuclear service. The combination of residual stress, interfacial inclusions, and microstructural heterogeneity creates a complex damage mechanism that may not be fully captured by conventional qualification testing. Engineers must therefore adopt a risk-based approach to the inspection and maintenance of overlay layers, incorporating advanced non-destructive testing techniques and periodic in-service inspection programs.

Another critical consideration is the effect of radiation embrittlement on the overlay layer and the interface. The neutron flux in the nuclear reactor environment can cause dislocation loop formation and precipitate coarsening in the stainless steel overlay, potentially reducing the ductility and interfacial strength. The study's findings should be interpreted in the context of the total damage mechanism, including both fabrication-related defects and in-service degradation.

Study Insights and Implications

The delamination mechanism analysis presented in this study underscores the importance of understanding the microstructural evolution at the interface between overlay layers and the base metal. The interface is not merely a mechanical boundary but a region of complex metallurgical interaction where inclusion formation, phase transformation, and stress accumulation interact to determine the long-term integrity of the component. Engineers involved in the fabrication and inspection of nuclear island equipment should prioritize interface characterization and adopt process controls that minimize interfacial defects, recognizing that even small delamination areas can serve as initiation sites for more extensive failure under service conditions.