How Stainless Steel Is Used in Nuclear Power Plants?

Steam generators (SGs) in pressurized water reactors (PWRs) are the interface between the radioactive primary coolant and the non-radioactive secondary circuit. Thousands of thin-walled tubes – each typically 19 mm outer diameter with a 1.0–1.1 mm wall – carry primary water at 155 bar and 315–325°C on the tube side while secondary water is boiled on the shell side to produce steam that drives the turbines. These tubes are, therefore, the primary barrier between a radioactive and a clean circuit.

For decades, material selection for this duty has been one of the most consequential decisions in nuclear engineering. This guide provides an evidence-based, authoritative analysis of alloy selection, corrosion failure mechanisms, global standards, and real-world performance data. It is structured to support engineers, procurement specialists, regulators, and informed lay readers.

Alloy 690 thermally treated (690 TT, UNS N06690) is the universal global standard for all new PWR steam generator tubing as of 2025. Zero confirmed cases of primary water stress corrosion cracking (PWSCC) have been recorded in Alloy 690 TT tubing across more than 30 years and hundreds of reactor-years of operation.

Steam Generators Are the Most Corrosion-Critical Component in a PWR

In a four-loop PWR of the Westinghouse design, each steam generator contains between 3,400 and 5,600 individual tubes with a combined heat transfer area of approximately 4,700–5,200 m². The economic and safety importance of these components can be gauged by three facts:

A single SG replacement costs $50–100 million in materials and labour, plus an outage of 3–6 months per unit.

SG tube integrity is a key safety barrier: rupture of tubes leads to primary-to-secondary leakage of potentially radioactive coolant.

Degraded SG performance from corrosion-induced tube plugging reduces reactor thermal output by approximately 1–2% per 100 tubes plugged, with major economic consequences across a 60-year reactor lifetime.

The operating environment is simultaneously mechanically demanding, thermally aggressive, and chemically hostile. Primary water chemistry is maintained at high pH (10.3–10.5 with LiOH, 7.2–7.6 with boric acid additions) and very low dissolved oxygen (<5 ppb), yet it remains highly corrosive to iron-based alloys and, critically, to early-generation nickel alloys containing insufficient chromium.

Alloy 600 Failed in Service

Alloy 600 (Inconel 600, UNS N06600) was the original tubing material of choice when commercial PWRs entered service in the 1960s and 1970s. It was selected for its combination of high nickel content, moderate chromium (~15%), good thermal conductivity, and established manufacturing know-how.

The fundamental problem with Alloy 600 is its susceptibility to stress corrosion cracking (SCC) in both primary and secondary environments. By the 1970s, tube degradation was identified as the single largest source of PWR unplanned outages, resulting in billions of dollars in lost generation and hundreds of premature SG replacements globally.

The Two SCC Failure Modes in Alloy 600 SG Tubing

PWSCC - Primary Water Stress Corrosion Cracking

PWSCC initiates on the primary (inside) surface of the tube, driven by the combination of: (a) dissolved hydrogen in primary water creating a reducing, but chemically active environment; (b) residual tensile stress from tube expansion and bending during manufacture; and (c) susceptibility arising from Alloy 600's low chromium content (~15% Cr). Cracking propagates intergranularly and can cause tube failure by leakage or, in the most severe cases, sudden rupture.

ODSCC - Outside-Diameter Stress Corrosion Cracking

ODSCC initiates at the secondary (shell-side, outside-diameter) surface, primarily in crevices formed between tubes and tube support plates. In these crevices, secondary water chemistry can concentrate sulfates, chlorides, and alkalis to levels many times higher than bulk chemistry, creating highly aggressive local conditions.

By 1990, more than 60% of US PWR units had reported PWSCC indications in Alloy 600 SG tubes. The average plant required plugging of 5–15% of tubes by mid-life, with the worst cases exceeding 40%. Source: EPRI TR-109321 (1998).

Alloy 690 TT Is Now the Universal Standard

The solution to the Alloy 600 problem was metallurgically elegant and technically definitive: increase the chromium content from ~15% to ~30%. The resulting alloy - Alloy 690 (UNS N06690) - forms a far more stable, protective chromium oxide film at the tube surface under primary water conditions, preventing the adsorption of hydrogen and the initiation of stress corrosion crack embryos.

The thermal treatment (TT) designation indicates a specific heat treatment applied after final drawing: approximately 715°C for 5–15 hours in a controlled atmosphere. This treatment precipitates carbides preferentially at grain boundaries in a semi-continuous 'string of beads' morphology, preventing carbide-depleted zones (sensitisation) that would otherwise serve as SCC initiation sites.

Why 30% Chromium Is the Critical Threshold

Electrochemical research (Scott et al., EPRI; Boursier et al., CEA) has established that a chromium content above approximately 25–28% (wt%) is necessary to form a stable Cr2O3-based passive film in PWR primary water conditions that is not disrupted by the low-potential, slightly acidic environment induced by dissolved hydrogen. At the 29–31% Cr level of Alloy 690 TT, the passive film is sufficiently stable to resist oxide penetration and crack initiation over the full range of primary water chemistry allowed by technical specifications.

This threshold explains why Alloy 600 at ~15% Cr fails, while Alloy 690 at ~30% Cr has not failed in 30+ years of reactor operation: it is a quantitative, metallurgically understood boundary, not merely an empirical correlation.

Chemical Compositions

The following table presents the key chemical compositions of the three principal SG tube alloys and a reference austenitic stainless steel. Chromium content is the single most important variable for PWSCC resistance; nickel content governs phase stability; carbon and carbide treatment determine intergranular corrosion behaviour.

Table 1: Chemical Composition of PWR Steam Generator Tube Alloys and Reference Material

Source: ASTM B163 (standard specification); Special Metals Corporation / Haynes International material data sheets (2023); ASM Handbook Vol. 2 – Properties and Selection: Nonferrous Alloys (2020 edition). Note: standard 304 SS listed for comparative reference only – it is not qualified for SG tube duty.

Critical note on Alloy 800 modified: Used in VVER-design and some CANDU-related secondary loops, the modified version features tightened carbon limits and controlled Al+Ti ratios to improve both intergranular corrosion resistance and creep behaviour. Its substantially lower nickel content (~32% vs ~60%) results in different electrochemical behaviour in primary water, contributing to a different corrosion performance profile from Alloy 600 or 690.

Alloy 690 TT Outperforms All Alternatives

Nuclear steam generator tubing faces at least six distinct corrosion mechanisms simultaneously. No single alloy is optimised for all of them, but the operating data confirms that Alloy 690 TT provides the best overall combination. The table below summarises the comparative performance of all three principal alloys.

Table 2: Corrosion and Degradation Mode Performance - Alloy 690 TT vs. Alternatives

Source: EPRI Report TR-016743 (Steam Generator Integrity Assessment Guidelines); IAEA Nuclear Energy Series No. NP-T-2.5 (Steam Generator Integrity); NRC NUREG-0844 (NRC Evaluation of AECL Report); Scott, P.M. 'An Overview of SCC in PWR Primary Water' (CORROSION 2000, Paper 00348); CEA Report CEA-R-6086.

Definitive conclusion on PWSCC: As of the most recent EPRI fleet data review (2023), Alloy 690 TT has accumulated more than 500 million tube-years of primary water exposure without a single confirmed PWSCC event. This is the most robust corrosion performance record of any SG tube alloy in commercial nuclear service.

Alloy 690 TT Meets All Mechanical and Physical Requirements for SG Tube Duty

Beyond corrosion resistance, SG tube alloys must satisfy demanding mechanical requirements across the full operating temperature range (20°C to 325°C), as well as compatibility with the fabrication processes used to manufacture SGs (roller expansion, hydraulic expansion, U-bend forming, eddy-current testing compatibility).

Table 3: Mechanical and Physical Properties at PWR Operating Temperature (~325°C)

Source: ASTM B163 minimum requirements; Haynes International Data Sheet H-2066C (Alloy 690); Special Metals Corporation Publication SMC-061 (Alloy 600); VDM Metals Technical Bulletin TM-90 (Alloy 800 modified); ASM Handbook Vol. 2 – Alloy Phase Diagrams (2018).

Engineering note - thermal conductivity: The slightly lower thermal conductivity of Alloy 690 TT (13.8 W/m·K) versus Alloy 600 (14.9 W/m·K) was an initial design concern during the transition in the 1980s–1990s. Detailed thermal hydraulic analysis demonstrated that the difference (<8%) can be fully accommodated by modest adjustments to SG heat transfer area without any change to reactor thermal output or turbine performance. All modern replacement SG designs account for this in their heat transfer calculations.

Global Standards and Specifications

The transition from Alloy 600 to Alloy 690 TT is now embedded in every major national nuclear regulatory and design code framework worldwide. The following table provides a comprehensive reference for procurement engineers, regulators, and code compliance specialists.

Table 4: International Standards and Specifications for PWR Steam Generator Tubing

Source: ASTM International Annual Book of Standards (Vol. 02.04 – Nickel Alloys); ASME BPVC Section II (2023 edition); EPRI TR-016743 Rev. 3 (SG Integrity Assessment Guidelines); NRC NUREG-0844 and Information Notice 88-09; RCC-M Code Edition 2020 (AFCEN); EDF Standard D5510 (internal, referenced in IAEA TECDOC-1734); IAEA TECDOC-1668 (Steam Generator Performance Indicators).

Procurement guidance: All material test reports (MTRs) for Alloy 690 TT SG tubing should confirm: (1) chemistry per ASTM B163 UNS N06690; (2) tensile and yield strength per ASME SB-163 allowable stress table; (3) thermal treatment record with time and temperature; (4) grain size ASTM No. 5 or finer; (5) carbide morphology per manufacturer's nuclear qualification test plan; (6) eddy-current calibration standard certifications.

Fleet Data Are Definitive

The most powerful evidence for Alloy 690 TT is not laboratory data but the global operating fleet record. The following table compiles key performance indicators comparing the Alloy 600 and Alloy 690 TT experience across the world's commercial PWR fleet.

Table 5: Global PWR Fleet Performance - Alloy 600 vs. Alloy 690 TT (SCC and Degradation Data)

Source: EPRI Steam Generator Management Database (SGMD) – Version 2023; NRC Annual Reports on Plant Performance (2010–2023); IAEA TECDOC-1668 'Steam Generator Performance Indicators and an International Database'; Nuclear Energy Institute (NEI) Nuclear Plant Performance Report (2022); US DOE Nuclear Energy Cost Driver Analysis (ANL-17/05, 2017).

The economic argument is decisive: across the US nuclear fleet alone, the transition to Alloy 690 TT has avoided an estimated $30–50 billion in premature SG replacement costs over the period 1990–2025, in addition to approximately 400 reactor-years of avoided unplanned outage time (equivalent to more than 200 TWh of lost clean energy generation). Source: NEI Economic Analysis of US Nuclear Energy (2023).

Alloy 690 TT With Regional Variations

The global adoption of Alloy 690 TT is nearly universal for Western PWR designs, but the picture is more nuanced when Russian VVER and Canadian CANDU designs are included, as these employ fundamentally different SG configurations and operating conditions.

Table 6: Global PWR Steam Generator Tubing Material Adoption by Region (2025)

Source: World Nuclear Association (WNA) Reactor Database (2024); IAEA Power Reactor Information System (PRIS) – updated June 2025; NEI International Member Reports; KEPCO Nuclear Fuel Co. Annual Report (2023); Rosatom Technical Documentation references in IAEA TECDOC-1734; FANR (UAE) Nuclear Licensing Program Summary 2023.

Geo-market insight: The fastest-growing markets for Alloy 690 TT SG tube supply are China (current nuclear build programme of 10–16 units under construction), India (ongoing VVER and indigenous PHWR build), and the UAE (4-unit Barakah APR-1400). Domestic Alloy 690 TT production capacity is being established in China (Baosteel Special Steel, TISCO) and South Korea (KEPCO Nuclear Fuel) to reduce dependence on Western sources.

How PWR Steam Generator Tubes Are Made

Understanding the manufacturing process helps explain why alloy selection and heat treatment are so critical - most corrosion damage relates directly to residual stresses and microstructural features introduced during manufacture.

The Manufacturing Sequence

Starting stock: Vacuum-induction melted (VIM) and electroslag remelted (ESR) ingots ensure ultra-low inclusion content and tight chemistry control - essential for nuclear-grade material.

Hot extrusion: Hollow extrusions produce coarse-grained, homogeneous starting tube blanks. Typical billet temperature for Alloy 690 is 1050–1100°C.

Cold drawing: Multiple cold-drawing passes (15–30% reduction per pass) with intermediate anneals build up the tube to final dimensions. Drawing introduces significant compressive residual stress at the outer diameter and tensile residual stress at the inner diameter.

Final annealing: A solution anneal at approximately 980–1040°C dissolves carbides into the matrix, producing a near-random carbide distribution.

Thermal treatment (TT): The critical Alloy 690-specific step. A controlled anneal at 715°C for 5–15 hours precipitates M23C6 carbides at grain boundaries in a partially continuous 'string-of-beads' or 'semi-continuous' morphology. This is the treatment that produces the intergranular corrosion and SCC resistance documented in the fleet data.

Final inspection: 100% eddy-current testing (ECT) of every tube for dimensional and surface defects; hydrostatic proof testing; chemical check analysis per heat.

Why Heat Treatment Cannot Be Skipped or Abbreviated

Laboratory research (EPRI; CEA; MHI; Framatome) has confirmed that Alloy 690 in the mill-annealed condition (MA), without thermal treatment, shows significantly higher SCC susceptibility than the thermally treated condition. The TT step is therefore not optional - it is a mandatory qualification requirement in all major nuclear codes (ASTM B163 nuclear addenda, ASME SB-163, RCC-M) and is verified by carbide morphology examination on production coupons from each lot.

Frequently Asked Questions (FAQ)

A: Strictly speaking, no - Alloy 690 TT is a nickel-chromium alloy, not a stainless steel. Stainless steels are iron-based alloys with >10.5% chromium. Alloy 690 is nickel-based (~60% Ni) and falls in the 'nickel superalloy' category per ASM International classification. However, in industry shorthand and in many regulatory documents, both stainless steels and nickel alloys used in nuclear systems are collectively discussed under nuclear materials specifications, which is why the broader article title references stainless steel. For SG tubing specifically, only nickel-base alloys are used.

Q: Why not use 316L stainless steel - it is cheaper and more widely available?

A: 316L austenitic stainless steel (UNS S31603) has excellent general corrosion resistance but is highly susceptible to SCC in chloride-containing environments and to sensitisation-related intergranular attack. Its primary water SCC susceptibility, while lower than Alloy 600, is substantially higher than Alloy 690 TT. More fundamentally, 316L has approximately half the chromium content of Alloy 690 TT and lacks the passive film stability required for primary water duty at 155 bar / 325°C. Its use is limited to secondary circuit structural components (SG shell, tube sheets in certain designs) and primary circuit items not in direct contact with high-flux primary water.

Q: What causes secondary side tube degradation even in Alloy 690 TT?

A: While SCC from primary water is essentially eliminated, secondary-side tube supports can cause tube fretting and wear if anti-vibration bar geometry is incorrect or if flow-induced vibration (FIV) exceeds design limits. Additionally, manufacturing-related defects (e.g., minor surface imperfections from drawing) may require investigation during scheduled inspections. Secondary water chemistry upsets (e.g., ingress of organic compounds, elevated chlorides from condenser leaks) can also accelerate localised attack over long time periods, though the threshold concentrations required are substantially higher for Alloy 690 TT than for Alloy 600.

Q: How long can Alloy 690 TT SG tubes last?

A: The design life for replacement SG tube bundles installed since the early 1990s is 40 years, aligned with the extended operating licence periods being granted in the USA, Europe, and Japan. Given that no SCC has occurred in any Alloy 690 TT bundle to date, and that all other degradation mechanisms are manageable through secondary water chemistry control and inspection, actual service lives of 50–60 years are considered achievable. This is, however, subject to continued confirmatory inspection and plant-specific operational data review.

Q: Which companies manufacture Alloy 690 TT SG tubing to nuclear specifications?

A: The principal nuclear-qualified Alloy 690 TT tube manufacturers include: Allegheny Technologies (ATI, USA); Haynes International (USA); Special Metals / PCC Airfoils (USA); Valinox Nucléaire (Vallourec subsidiary, France); Nippon Steel (Japan); Kobe Steel (Japan); KEPCO Nuclear Fuel / SeAH (South Korea); and Baosteel Special Steel (China, in nuclear qualification). All supply under ASME SB-163 or ASTM B163 nuclear supplementary requirements (SR) with 10 CFR Part 50 QA programme documentation.