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The austenitic Ni-base Alloy 600 has been extensively used as structural material in primary water reactors (PWR). Despite good resistance against general corrosion in water-cooled nuclear power reactors, the material has been susceptible to stress corrosion cracking (SCC). These observations have led to ongoing discussions of the underlying embrittlement mechanism(s). Internal oxidation of the grain boundary (GB) at typical operating temperatures is one such mechanism, although debate continues on the exact mechanisms at play.
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Zirconium alloys are used in civil nuclear reactors as fuel cladding, due to their relatively low neutron absorption cross section and being ability to maintain integrity during operation in a challenging environment – under elevated temperature, in corrosive water, under stress, and while being bombarded with different types of irradiation. In anisotropic materials, such as the hexagonal close packed Zr crystal structure, irradiation-induced defects diffuse at different rates in different crystallographic directions.
The Neutral Beam Test Facility (NBTF) at Consorzio RFX (Padua, Italy) hosts two experiments: MITICA, the 1 MeV full-scale prototype of the ITER Neutral Beam Injector (NBI), and SPIDER, the low energy 100 keV ITER HNB full-size Ion Source [1]. The exploitation of SPIDER and MITICA is necessary to make the future operation of the ITER NBI efficient and reliable; a fundamental aspect to achieving thermonuclear-relevant plasma parameters in ITER.
All tests in this program have been performed under simulated PWR primary cooling water conditions. The oxide layer development and morphology is addressed in the literature and more intensively being investigated during the last ten years. The oxide layer that is typically observed under these conditions has a double layer structure. The outer layer is composed of large particles of Fe3O4 and the inner layer mainly consists of small particles of FeCr2O4, e.g.
Alloy 690 has been utilized since the late 1980s as a replacement for Alloy 600 in pressurized water reactors (PWR) pressure boundary components due to laboratory data indicating higher resistance to stress corrosion cracking (SCC). Although to date no SCC incidents have been reported on Alloy 690 components in service, the growing interest of extending the operation life of PWRs beyond 60 or even 80 years has raised concerns for the potential occurrence of long-range ordering (LRO) in Alloy 690 and its compatible weld metals.
Most of countries store spent nuclear fuels in pools (SFP) which are built in nuclear power plants. As number of nuclear power plants and corresponding number of spent fuels increased, density in SFP storage rack also increased. In this regard, maintain subcriticality of spent nuclear fuels was raised as an issue and BSS was selected as structural material and neutron absorber for high density storage rack. BSS has better mechanical properties than other neutron absorbers which were fabricated based on aluminum alloys.
Nickel-based Alloy 690 and the associated weld Alloys 52 and 152 are typically used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels, because of their increased resistance to stress corrosion cracking (SCC) relative to Alloys 600, 82, and 182. Many of these reactors are expected to operate for 40-80 years. Likewise, advanced water-cooled small modular reactors (SMRs) will use Ni-Cr alloys in their primary systems and are expected to receive initial operating licenses for 60 years.
A cornerstone of aging management programs for commercial nuclear reactors is the condition monitoring techniques used to determine insulation degradation of cables. Improved condition monitoring methods has been the focus of research especially for low voltage cables. There are many effective methods available such as elongation at break, indenter modulus, oxidation induction, etc.
Based on US Energy Information Administration (EIA) Annual Energy Outlook (AOE) predictions, by 2050, the US nuclear capacity for electricity generation may decrease to ~ 80% of 2021 levels, Moreover, by 2050, nearly 50% of the existing US LWR fleet will be within 10 years of 80 years of operation suggesting that without a Life Beyond 80 (LBE) plan and limited new builds and advanced reactors, the US could lose up to 50% of its nuclear capacity resulting in a ~30-gigawatt (GW) capacity shortage by 2060. These numbers could change dramatically depending upon oil and gas supplies, and the growth of renewables.
Microreactor technology has the potential to provide efficient, modular, and inherently safe baseload power that can be used in regions that are too remote to support the larger, light water reactor (LWR) technology that dominates today’s nuclear energy landscape. To generate enough power and thermal efficiency to be attractive, the microreactors must be operated at higher temperatures (approximately 1112-1652°F or 600-900°C) than traditional LWR’s, and therefore are cooled using technologies such as heat pipes with gas, sodium, or molten salt coolants.
The corrosion of Zircaloy-4 under autoclave conditions without the presence of radiation is relatively well understood, with the development of cyclic corrosion kinetics that are well simulated by correlative predictive models (1) (2). Under irradiation in a PWR environment, however, the corrosion kinetics of Sn-containing Zr alloys are severely accelerated and although early corrosion behaviour is unchanged, after an oxide thickness of ~5 μm, accelerations of up to 40 x out-of-pile behaviour are observed (3) (4). Among the likely contributors to this accelerated corrosion are neutron irradiation damage to both the substrate and oxide, gamma irradiation, radiolysis, and hydrogen effects.