Jeremy T Busby

Structural materials for fission & fusion energy

Structural materials represent the key for containment of nuclear fuel and fission products as well as reliable and thermodynamically efficient production of electrical energy from nuclear reactors. Similarly, high-performance structural materials will be critical for the future success of proposed fusion energy reactors, which will subject the structures to unprecedented fluxes of high-energy neutrons along with intense thermomechanical stresses. Advanced materials can enable improved reactor performance via increased safety margins and design flexibility, in particular by providing increased strength, thermal creep resistance and superior corrosion and neutron radiation damage resistance. In many cases, a key strategy for designing high-performance radiation-resistant materials is based on the introduction of a high, uniform density of nanoscale particles that simultaneously provide good high temperature strength and neutron radiation damage resistance.

Materials challenges for nuclear systems

The safe and economical operation of any nuclear power system relies to a great extent, on the success of the fuel and the materials of construction. During the lifetime of a nuclear power system which currently can be as long as 60 years, the materials are subject to high temperature, a corrosive environment, and damage from high-energy particles released during fission. The fuel which provides the power for the reactor has a much shorter life but is subject to the same types of harsh environments. This article reviews the environments in which fuels and materials from current and proposed nuclear systems operate and then describes how the creation of the Advanced Test Reactor National Scientific User Facility is allowing researchers from across the United States to test their ideas for improved fuels and materials.

Corrosion Behavior of Model Zirconium Alloys in Deaerated Supercritical Water at 500°C

Abstract Several zirconium alloys with differing weight percentages of Cr, Fe, Cu, and Mo were exposed to flowing, pure supercritical water at 500°C for up to 150 days in an effort to determine their corrosion behavior for consideration in the supercritical water reactor. The weight gains of the alloys were measured, and oxides were characterized after various times. The test results showed a wide range of corrosion behavior depending on the alloy composition and process temperature. The alloys most resistant to corrosion were those containing Cr and Fe, three of which showed protective stable oxides, low corrosion rates, and no breakaway behavior. The ZrCr, ZrCu, ZrMo, and ZrCuMo alloys all exhibited high corrosion rates and nonprotective oxides. Analysis of the oxide layer showed that the oxide consisted mostly of monoclinic zirconia (ZrO2). The structure of the oxide-metal interface in the five protective alloys exhibited characteristics that were also seen in protective oxides formed at low temperatur...

Role of Scale Factor During Tensile Testing of Small Specimens

The influence of scale factor (tensile specimen geometry and dimensions) on mechanical test results was investigated for different widely used types of small specimens (SS-1, SS-2, SS-3, and SS-J3) and a set of materials. It was found that the effect of scale factor on the accurate determination of yield stress, ultimate tensile stress, and uniform elongation values was weak; however, clear systematic differences were observed and should be accounted for during interpretation of results. In contrast, total elongation values were strongly sensitive to variations in specimen geometry. Modern experimental methods like digital image correlation allow the impact of scale factor to be reduced. Using these techniques, it was shown that true stress true strain curves describing strain-hardening behavior were very close for different specimen types. The limits of miniaturization are discussed, and an ultra-miniature specimen concept was suggested and evaluated. This type of specimen, as expected, may be suitable for SEM and TEM in situ testing.

TEM Characterization of Crept and Irradiated Nano-structured Ferritic Alloys

The past ten years or so have seen the development of an exciting new class of mechanically alloyed (MA) nano-structured ferritic alloys (NFA) with outstanding mechanical properties that are mostly due to the presence of high concentrations (>10 23 m -3 ) of Ti-, Y-, and O-enriched nano-clusters (NC). Because NC may promote point defect recombination and trap transmutation-produced He in small clusters, NFA have the potential to be highly resistant to radiation damage in fission and fusion environments [1,2], and thus are being characterized following neutron and ion irradiation. Energy-filtered transmission electron microscopy (EFTEM) performed at 300 kV on a LaB6 Philips CM30 equipped with a Gatan imaging filter (GIF) has been especially beneficial for imaging NC. In particular, Fe-M jump-ratio images produced from component images recorded with 10-eV slits at energy losses of 46 and 62 eV reliably reveal NC in dark contrast. Such images are insensitive to surface oxide films or modest surface contamination and for sufficiently thin regions (<50 nm) 2-nm diameter NC are visible [3]. Additional EFTEM elemental mapping (e.g. O, Ti-L23, Cr-L23) has also been usefully applied to NFA, and focused-ion-beam (FIB) lift-out specimens have been used to good advantage [2]. Fabrication of an Fe-14.2wt.%Cr-1.95%W-0.22%Ti-0.25%Y2O3 NFA, designated 14YWT, has been described elsewhere, as have the contributions of TEM to help optimize material processing parameters [3,4]. It was also previously shown that NC in 14YWT are not detectably changed by tensile testing at 25 and 700°C with total strains of up to 39% [5] and that in MA957 (an INCO-patented Fe-14wt%Cr-1%Ti-0.3%Mo-0.27%Y2O3 NFA) neutron irradiated at 500°C to 9 displacements per atom (dpa) and with ~380 appm He, the diameter (~3 nm) and concentration (~4 x 10 23 m -3 ) of the NC differ little from those of unirradiated MA957 [1,2].

Materials Degradation in Light Water Reactors: Life After 60,???

Nuclear reactors present a very harsh environment for components service. Components within a reactor core must tolerate high temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can lead to reduced performance, and in some cases, sudden failure. A recent EPRI-led study interviewed 47 US nuclear utility executives to gauge perspectives on long-term operation of nuclear reactors. Nearly 90% indicated that extensions of reactor lifetimes to beyond 60 years were likely. When polled on the most challenging issues facing further life extension, two-thirds cited plant reliability as the key issue with materials aging and cable/piping as the top concerns for plant reliability. Materials degradation within a nuclear power plant is very complex. There are many different types of materials within the reactor itself: over 25 different metal alloys can be found with can be found within the primary and secondary systems, not to mention the concrete containment vessel, instrumentation and control, and other support facilities. When this diverse set of materials is placed in the complex and harsh environment coupled with load, degradation over an extended life is indeed quite complicated. To address this issue, the USNRC has developed a Progressive Materials Degradation Approach (NUREG/CR-6923). This approach is intended to develop a foundation for appropriate actions to keep materials degradation from adversely impacting component integrity and safety and identify materials and locations where degradation can reasonably be expected in the future. Clearly, materials degradation will impact reactor reliability, availability, and potentially, safe operation. Routine surveillance and component replacement can mitigate these factors, although failures still occur. With reactor life extensions to 60 years or beyond or power uprates, many components must tolerate the reactor environment for even longer times. This may increase susceptibility for most components and may introduce new degradation modes. While all components (except perhaps the reactor vessel) can be replaced, it may not be economically favorable. Therefore, understanding, controlling, and mitigating materials degradation processes are key priorities for reactor operation, power uprate considerations, and life extensions. This document is written to give an overview of some of the materials degradation issues that may be key for extend reactor service life. A detailed description of all the possible forms of degradation is beyond the scope of this short paper and has already been described in other documents (for example, the NUREG/CR-6923). The intent of this document is to present an overview of current materials issues in the existing reactor fleet and a brief analysis of the potential impact of extending life beyond 60 years. Discussion is presented in six distinct areas: (1) Reactor pressure vessel; (2) Reactor core and primary systems; (3) Reactor secondary systems; (4) Weldments; (5) Concrete; and (6) Modeling and simulations. Following each of these areas, some research thrust directions to help identify and mitigate lifetime extension issues are proposed. Note that while piping and cabling are important for extended service, these components are discussed in more depth in a separate paper. Further, the materials degradation issues associated with fuel cladding and fuel assemblies are not discussed in this section as these components are replaced periodically and will not influence the overall lifetime of the reactor.

Stress Corrosion Crack Initiation Susceptibility of Irradiated Austenitic Stainless Steels

The susceptibility of neutron irradiated austenitic stainless steels to the initiation of irradiation-assisted stress corrosion cracking (IASCC) was assessed. Solution annealed (SA), high purity (HP) type 304 stainless steel with and without additions of Mo and Si, and HP type 316L +Hf were strained by constant extension rate testing (CERT) in simulated 288°C BWR NWC at a rate of 3.5 × 10−7/s. CERT test data and fracture analysis showed that IASCC susceptibility increased in order of HP304, HP304+Mo, HP316L+Hf, and HP304+Si. This trend was also observed when comparing fracture surfaces of the same alloys tested by CERT in BWR NWC after proton irradiation. Differences were insignificant among reported crack growth rate (CGR) values for the same neutron irradiated alloys, and no connection between crack initiation and CGR was confirmed from the alloys tested.

Tensile and toughness assessment of the procured advanced alloys

Life extension of the existing nuclear reactors imposes irradiation of high fluences to structural materials, resulting in significant challenges to the traditional reactor materials such as type 304 and 316 stainless steels. Advanced alloys with superior radiation resistance will increase safety margins, design flexibility, and economics for not only the life extension of the existing fleet but also new builds with advanced reactor designs. The Electric Power Research Institute (EPRI) teamed up with Department of Energy (DOE) to initiate the Advanced Radiation Resistant Materials (ARRM) program, aiming to develop and test degradation resistant alloys from current commercial alloy specifications by 2021 to a new advanced alloy with superior degradation resistance by 2024 in light water reactor (LWR)-relevant environments

Computational Thermodynamics for Interpreting Oxidation of Structural Materials in Supercritical Water

The Supercritical water-cooled reactor (SCWR) is one of the advanced nuclear reactors being developed to meet the soaring energy demand. The corrosion resistance of structural materials used in the SCWR becomes one of the major concerns as the operation conditions are raised up to ~600°C and ~25 MPa as compared to pressurized water reactors (PWRs) at ~315°C and ~15.5 MPa. Oxidation has been observed as the major corrosion behavior. To mitigate the oxidation corrosion, stabilities of metals and oxides need to be understood with respect to environmental temperature and oxygen partial pressure. Computational thermodynamics provides a practical approach to assess phase stabilities of such multicomponent multi-variable systems. In this study, calculated phase stability diagrams of alloys and corresponding oxides were used to guide the interpretation of oxidation behavior of SCW-exposed structural materials. Examples include ferritic-martensitic steel, austenitic steels and Ni-base alloy, e.g., HCM12A (Fe-11Cr), D9 (Fe-15Cr-15Ni), 800H (Fe-21Cr-32Ni), and 690 (Ni-30Cr-10Fe). Calculated results are in good overall consistence with the experimental data./

Activated Corrosion Product Analysis. Analytical Approach.

The presence of activated corrosion products (ACPs) in a water cooling system is a key factor in the licensing of ITER and affects nuclear classification, which governs design and operation. The objective of this study is to develop a method to accurately estimate radionuclide concentrations during ITER operation in support of nuclear classification. A brief overview of the PACTITER numerical code, which is currently used for ACP estimation, is presented. An alternative analytical approach for calculation of ACPs, which can also be used for validation of existing numerical codes, including PACTITER, has been proposed. A continuity equation describing the kinetics of accumulation of radioactive isotopes in a water cooling system in the form of a closed ring has been formulated, taking into account the following processes: production of radioactive elements and their decay, filtration, and ACP accumulation in filter system. Additional work is needed to more accurately assess the ACP inventory in the cooling water system, including more accurate simulation of the Tokamak cooling water system (TCWS) operating cycle and consideration of material corrosion, release, and deposition rates.