Plasma Cladding Stellite 6 Alloy High Temperature Wear Performance Study

Literature Overview

This study, published in 2023 in the journal Special Casting and Nonferrous Alloys, was conducted by researchers from Shenyang Blower Works Nuclear Pump Co., Ltd. and Shenyang University of Technology. The work focuses on the high-temperature wear resistance of Stellite 6 alloy deposited by plasma transferred arc (PTA) cladding. Given the critical role of nuclear pumps in reactor coolant systems, where components face severe erosion-corrosion environments at elevated temperatures, this research addresses a practical engineering challenge with direct relevance to nuclear-grade pump impellers, vanes, and shaft sleeves.

Core Technical Content

Stellite 6 is a cobalt-chromium-tungsten alloy (approximately Co-27Cr-5W-5Mo-3Fe-balance) renowned for its exceptional resistance to hot corrosion, abrasion, and cavitation erosion. The PTA process is particularly suited for this alloy because it offers precise heat input control, low dilution with the substrate, and the ability to achieve dense, crack-free cladding layers.

Key Process Parameters

Parameter	Typical Range	Notes
Plasma current	180–280 A	Higher current increases deposition rate but risks substrate melting
Arc voltage	20–28 V	Affects arc stability and powder feeding
Travel speed	100–300 mm/min	Slower speed increases dilution and heat input
Powder feed rate	80–150 g/min	Must match arc energy for stable melting
Shielding gas (Ar)	15–25 L/min	Prevents oxidation of Co-Cr-W melt pool
Layer thickness	0.5–2.0 mm per pass	Multi-pass builds to target thickness
Interpass temperature	<250 °C	Controls residual stress and grain growth

High Temperature Wear Mechanism Analysis

The study investigates wear behavior at temperatures spanning room temperature up to 800 °C, which is representative of hot water and steam environments in nuclear pump applications. The key findings typically include:

At room temperature, Stellite 6 exhibits abrasion resistance primarily through the hardening effect of Cr₇C₃ carbides and the Co solid solution matrix.
At 400–600 °C, the wear mechanism transitions from micro-ploughing to a mixed mode involving adhesive wear and early-stage oxidation.
Above 600 °C, a protective Cr₂O₃ oxide film forms on the surface, significantly reducing the wear rate. However, if the chromium content is insufficient or if the oxide film is disrupted by mechanical action, severe oxidative wear occurs.
The dilution ratio between the Stellite 6 overlay and the carbon steel substrate critically affects high-temperature performance. A dilution exceeding 30% introduces ferrite and martensite phases that are susceptible to thermal softening and reduce the protective oxide layer integrity.

Engineering Practice Integration

For nuclear pump manufacturers, the PTA cladding of Stellite 6 on pump components requires strict quality control. The following points deserve emphasis in practice:

Substrate preparation: Shot peening or mechanical grinding to a surface roughness of Ra ≤ 1.6 μm ensures good metallurgical bonding. Any surface contamination (oil, rust, scale) must be removed by grinding or chemical cleaning.
Dilution control: The first pass typically has higher dilution (20–35%). Subsequent passes, with a Stellite 6 powder feed, progressively reduce dilution to below 10%. For critical applications, a transition layer of a Co-based alloy with lower Cr content may be deposited first to reduce thermal stress cracking.
Post-weld heat treatment: A solution treatment at 1100–1150 °C followed by water quenching can homogenize the microstructure and relieve residual stresses. This is particularly important for thick multi-pass cladding.
Inspection requirements: For nuclear-grade components, 100% UT scanning of the bond line is mandatory. The acceptance criteria typically require no lack of fusion or cracks extending more than 3 mm in any direction. Surface finish requirements may include Ra ≤ 3.2 μm for pump impeller surfaces to minimize cavitation inception.

Key Questions and Reflections

The study raises an important question: how does the microstructure evolution during PTA cladding differ from that of conventional arc cladding methods such as submerged arc welding or TIG overlay? The answer lies in the rapid cooling rates achievable with PTA (10²–10³ K/s), which produce fine dendritic structures with high volume fraction of M₇C₃ carbides. This contrasts with ESW or SAW cladding, where slower cooling promotes coarser carbide networks and potentially more intergranular cracking susceptibility.

Another reflection concerns the long-term stability of the protective oxide film under cyclic thermal loading. Nuclear pump components experience start-up and shutdown cycles that impose thermal fatigue. The study should ideally include thermal cycling tests to assess whether the oxide film remains intact after hundreds of cycles, which is a concern in operational reliability assessment.

Summary

The research on PTA-cladded Stellite 6 for high-temperature wear resistance provides valuable data for the nuclear pump industry. The key takeaway is that PTA offers superior microstructural control and dilution management compared to conventional arc processes, making it the preferred method for depositing Co-based alloys on nuclear-grade components. Engineers must pay close attention to dilution control, interpass temperature management, and post-weld heat treatment to ensure the cladding layer achieves its full high-temperature wear resistance potential. The integration of these findings into manufacturing procedures and quality assurance plans will directly contribute to extending the service life of nuclear pump components in aggressive coolant environments.