Advanced oxidation-resistant iron-based alloys for LWR fuel cladding☆
Introduction
In 1975 Hyman Rickover summarized the key considerations that led to his decision almost three decades earlier to use zirconium alloys for the fuel cladding in U.S. Navy’s pressurized water nuclear reactor [1].
Almost four decades later, zirconium alloys enjoy a monopoly for uranium oxide fuel cladding material in light water reactors (LWRs) and are used in other fuel bundle structures such as portions of the grid spacers, as the channel box material for boiling water reactor (BWR) fuel assemblies, and elsewhere. Zirconium triumphed over other fielded clad options such as stainless steel, beryllium, and aluminum due to a combination of its small neutron capture cross section, reasonable corrosion resistance, and structural integrity under envisioned operating conditions. Today, zirconium alloy technology benefits from decades of active research and development, enabling tailored alloy chemistries and production techniques for optimized performance under pressurized or boiling water reactor conditions. The ongoing development activities have led to the introduction of new generations of zirconium-based alloys that exhibit enhanced corrosion resistance while limiting the detrimental irradiation effects (i.e., growth and creep) to those that are frequently inconsequential to reactor operation [2]. As summarized by Edsinger [3], these metallurgical enhancements, combined with incremental improvements in management for optimized reliability over the decades, have led to an impressive record of fuel reliability for the current zirconium-alloy cladding. To achieve this milestone, zirconium-alloy-cladded fuel production, fuel bundle handling, core operation, and water chemistry control have all been simultaneously optimized and improved over this time period.
This manuscript reexamines iron-based alloys for their potential application as nuclear fuel cladding to replace zirconium alloys. The motivation behind this effort is twofold. First, specific limitations with zirconium alloys under both design-basis and beyond-design-basis nuclear reactor accident scenarios provide incentive to explore alternative cladding options that may enable improved accident tolerance while maintaining good performance under normal operating conditions. These limitations are perceived to be serious, and therefore a renewed discussion is needed in the technical and policy communities.
Secondly, several new generations of increasingly higher performance steels are now commercially available that may offer significant performance improvements over the relatively simple austenitic steels that were utilized as cladding in some early commercial fission reactors [4]. Both alloy systems were successful in their early application, although zirconium alloys ultimately prevailed largely owing to their reduced thermal neutron cross section—a clear advantage over iron alloys for compact submarine reactor application. Driven by the nuclear navy, the zirconium alloy system became the focus of significant R&D, resulting in alloys with superior combined nuclear and corrosion performance and a ready-made commercial infrastructure. Finally, superior stress corrosion cracking resistance in BWRs drove the complete phase out of stainless steel claddings. However, given the significant progress over the past five decades in advanced steel making, including corrosion-resistant steels (combined with our present understanding of chemistry control and environmental effects), high-strength steels, and the ability to form thin-walled tubing, it appears clear that the commercial steels utilized in the early commercial reactors can be significantly improved upon, much as the zirconium alloys were improved upon since the days of Rickover’s program.
A review of the limitations with the current zirconium alloy system under design-basis and beyond-design-basis accident scenarios is offered. This review also provides simple metrics to guide selection of new clad materials. Based on the findings of a recent experimental survey of the high-temperature steam oxidation kinetics of historic, present-day commercial, and advanced steels [5], the potential to achieve higher margins of safety through replacement of zirconium alloys with advanced iron-based alloys inside LWR cores is showcased with a simplified set of analyses. A review of earlier generation stainless steels as nuclear fuel cladding in LWRs is provided and the anticipated performance of advanced iron-based alloys is briefly evaluated from a materials and environmental performance perspective, focusing on irradiation behavior, corrosion, and mechanical integrity under LWR normal operating conditions. Finally, the reactor physics characteristics and the impact on economics of nuclear electricity production upon adoption of advanced steel nuclear fuel cladding is examined with a specific set of case studies. In this manner, impactful areas for future targeted R&D in this area are identified and a broad basis for consideration of these advanced cladding concepts is provided.
Section snippets
Limitations with zirconium alloy cladding
Given decades of active research and development, today’s zirconium alloy cladding technology exhibits optimized behavior under normal operating conditions where the fuel rod failure rate (typically involving a localized breach of the cladding) is on the order of a few ppm per year. Industry-led fuel reliability programs with a long-term focus have resulted in dramatic reductions in the number of fuel failures across the LWR fleet [6]. These improvements in fuel performance are concomitant with
Gains in safety margins using advanced oxidation-resistant iron-based alloys
In order to reduce the slope of the rapid temperature excursions in the cladding and thereby gain valuable additional coping time during severe accidents, the magnitude of LHRox needs to be suppressed. One way to achieve this is to consider materials that exhibit slower steam oxidation kinetics than zirconium alloys.
A recent set of experiments was undertaken to examine high-temperature 1–20 bar steam oxidation behavior of a wide variety of iron-based alloys and SiC-based materials with the
Historical application of steels as LWR cladding
In the initial days of LWR deployment, austenitic stainless steels were a popular choice for cladding material of UO2 fuel pins for both experimental and commercial PWRs and BWRs [4], [53], [54], [55]. The alloys utilized during the 1960–1975 time frame included Types 316, 304, 304L, and 347 stainless steels (both cold-worked and annealed), and nickel alloys such as Inconel 800 and Inconel 600 [4].
The earliest reported cladding failures occurred in the fuel pins of numerous BWRs. For example,
Prospects for the utilization of advanced iron alloys in LWRs
Significant advances in both austenitic [71], [72], [73] and ferritic/martensitic [71], [72], [74], [75], [76], [77] steels have been achieved over the past four decades since relatively simplistic austenitic steels were last used as a fuel cladding material in LWRs. The typical cycle time for the development of a new generation of steels is about 15 years [72], [74], which translates into approximately three to four generations of improved austenitic steels compared to the Type 304, 316, and
Reactor physics and economic aspects of iron-based alloy cladding
Following high temperature steam oxidation behavior and expected materials performance under normal operating conditions, two essential areas remain that need to be addressed for iron based alloy cladding systems for LWR fuel elements: reactor physics and economics. Though a neutronic penalty upon the transition to iron based alloy cladding materials is expected, it is important to quantify its magnitude and discuss alternate fuel design configurations that reduce it. The magnitude of this
Conclusions
The following conclusions are drawn given the discussion provided in this manuscript:
- 1.
Zirconium alloys are fundamentally susceptible to severe degradation under beyond-design-basis-accident conditions. The chemical and physical degradation processes in the core during an accident sequence are considerably exacerbated by rapid oxidation of zirconium alloys at T > 1200 °C.
- 2.
Advanced iron alloys offer potential for improved oxidation resistance (reduced hydrogen generation) and improved strength
Acknowledgments
The authors would like to extend their gratitude to Larry Ott, Kevin Robb, and Graydon Yoder in the Reactor and Nuclear Systems Division as well as Bruce Pint and Sebastian Dryepondt in the Materials Science and Technology Division at ORNL for their technical insight. Thoughtful discussions and guidance received from Robert Montgomery at PNNL and Dion Sunderland at Anatech Corp are also gratefully acknowledged. The work presented in this paper was supported partially by the Advanced Fuels
References (157)
- et al.
Journal of Nuclear Materials
(2013) - et al.
Journal of Nuclear Materials
(2012) - et al.
Journal of Nuclear Materials
(1979) - et al.
Journal of Nuclear Materials
(2011) - et al.
Journal of Nuclear Materials
(1991) Nuclear Engineering and Design
(1994)- et al.
Nuclear Engineering and Design
(2004) - et al.
Journal of Nuclear Materials
(2010) Journal of Nuclear Materials
(1999)- et al.
Nuclear Engineering and Design
(2010)
Nuclear Engineering and Design
Journal of Nuclear Materials
Journal of Nuclear Materials
Nuclear Engineering and Design
Journal of Nuclear Materials
Nuclear Engineering and Design
Journal of Nuclear Materials
Journal of Nuclear Materials
Journal of Nuclear Materials
Journal of Nuclear Materials
Materials Today
Journal of Nuclear Materials
Journal of Nuclear Materials
Journal of Nuclear Materials
Materials Science and Engineering R: Reports
Nuclear Engineering and Design
Journal of Nuclear Materials
Journal of Nuclear Materials
Comprehensive Nuclear Materials
Journal of Nuclear Materials
Comprehensive Nuclear Materials
Nuclear News
Solid State Phenomena
Journal of ASTM International
Nuclear Engineering and Technology
Journal of ASTM International
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