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To address concerns about the degradation of unprotected internal surfaces of nuclear fuel claddings, inner-side coatings have been proposed as a complementary approach to the accident tolerant fuel protective coating concept. In addition to the increased coping time during severe events, these coatings are expected to provide enhanced protection in normal operation conditions. This study analyzes the neutronic performance of chromium (Cr) inner-side coatings in VVER-type fuel assemblies. Different aspects, such as reactivity and cycle length penalties, enrichment requirements, neutron flux, and the associated isotopic concentration changes are discussed, considering both the coating thickness and position. The results show that for the same thickness, reactivity penalties due to the use of inner-side Cr coatings will be (~30% on average) less compared to external coatings. The fuel assembly operating cycle showed reductions by ~ 5.5 effective full-power days when a 10-µm-thick internal Cr coating is introduced, while a 10-µm two-sided coated assembly possessed an ~13.6-day shorter operating cycle compared to an uncoated fuel assembly of the same specifications. The neutron flux showed slight shifts and hardening in the thermal energy region. The analysis of nuclide inventories showed relative increases in these inventories, which were proportional to the thickness. For the fissile plutonium isotope 239Pu, this relative increase reached a peak of 0.25% and 0.42% (for the 10-µm and 20-µm internal Cr coatings) at a fuel burnup of 18 MWd/kg heavy metal (HM). While for 241Pu, the observed highest relative increases for the 10-µm- and 20-µm-thick internal Cr coatings were 0.74% and 1.03%, respectively. The 135Xe isotopic concentration showed a relative increase that reached 0.2% and 0.4% for the 10-µm and 20-µm internal Cr coatings at a burnup of 34 MWd/kgHM, while the 149Sm concentration increased by 0.2% and 0.5% for the 10-µm and 20-µm internal Cr coatings, respectively. While these observed isotopic concentration changes were generally small for the studied inner-side coatings, the results showed that the changes remain subject to further increases as the amount of coating material gets higher. Therefore, it is important for the coating thickness to be optimized, taking into account the impact of such nuclide inventory changes. Possible fuel-clad gap reductions and the associated effects on heat transfer, as well as gap tolerance to fission products and fuel relocations, will require further studies, especially so that additional enrichments may be applied. [ABSTRACT FROM AUTHOR]

Accident tolerant fuel (ATF) is expected to delay or prevent core damage by providing additional coping time under accidents involving loss of core cooling. The effect of extended coping time may vary depending on the plant response to accidents. Age-related component degradation that deteriorates plant performance over time could have an impact on the actual advantages of ATF. The potential safety benefits of two near-term ATF candidates, including Cr-coated Zr cladding and FeCrAl cladding, are assessed for a 2-in. loss-of-coolant accident with failed high-pressure safety injection using the dynamic event tree (DET) approach considering possible stress corrosion cracking of steam generator (SG) tubing under aging. The DET approach allows likelihood quantification of accident sequences leading to core damage, including stochastic variation of system response and human actions during accident mitigation. The safety benefits of the selected ATF claddings in terms of additional coping time and the core damage frequency reduction rate under specified accident situations were quantitatively estimated. The results show that the deployment of the two selected ATF claddings is expected to lead to longer coping times and lower core damage frequency due to the wider safety margin to peak cladding temperature they provide. The safety advantages would be greater as SG tube degradation proceeds. Thus, the two ATF candidates would lead to less severe consequences in terms of likelihood of core damage and susceptibility to the SG tube degradation than UO2-Zr fuel. [ABSTRACT FROM AUTHOR]

Accident tolerant fuels (ATFs) are new nuclear fuels developed in response to the accident at the Fukushima power station in March 2011. The goal of ATFs is to withstand accident scenarios through better performance compared to currently employed fuels (e.g., small-scale hydrogen generation). This paper targets a method for evaluating and comparing ATF performance from a probabilistic risk assessment (PRA) perspective by employing a newly developed combination of event trees and dynamic PRA methods. Compared to classical PRA methods based on event trees and fault trees, dynamic PRA can evaluate with higher resolution the safety impacts of physics dynamics and the timing/sequencing of events on the accident progression without the need to introduce overly conservative modeling assumptions and success criteria. In this paper, we analyze the impact on the accident progression of three different cladding configurations for two initiating events [a large break loss-of-coolant accident (LB-LOCA) and a station blackout (SBO)] by employing dynamic PRA methods. The goal is to compare the safety performance of ATFs (FeCrAl and Cr-coated cladding) and the currently employed Zr-based clad fuel. We employ two different strategies. The first focuses on the identification of success criteria discrepancies between the accident sequences generated by the classical PRA model and the set of simulation runs generated by dynamic PRA using ATF. The second one, on the other hand, directly uses dynamic PRA to evaluate the impact of timing of events (e.g., recovery actions) on accident progression. By applying these methods to the LB-LOCA and SBO initiating events, we show how dynamic PRA methods can provide analysts with detailed and quantitative information on the safety impact of ATFs. [ABSTRACT FROM AUTHOR]

Since the accident at Fukushima, one major goal of reactor safety research has been the development of accident tolerant technologies that can mitigate or delay fuel degradation during a beyond-design-basis accident. One major effort has been focused on the development of coatings for light water reactor fuel cladding. Chromium-coated zirconium-alloy clad is one of the leading options. In this work, the MELCOR systems code (version 1.8.6 User-Defined Generalized Coating) is used to evaluate the performance of Cr-coated Zr-alloy clad as compared to Zr-alloy clad and APMT FeCrAl-coated Zr-alloy clad for a pressurized water reactor (i.e., Surry) for a station blackout (SBO) accident scenario. Our focus is primarily on the accident progression behavior depending on oxidation kinetics and the assumed failure criterion for the coated cladding material. Our simulation and comparison indicate that the presence of the coating material can significantly reduce the initial rate of hydrogen generation and delay the time when hydrogen generation becomes significant. This decrease in the rate of oxidation and delay in timing can provide additional coping time for compensatory operator actions. We also note that the effect of extended auxiliary feedwater system operation (long-term SBO) can increase this additional coping time in combination with Cr-coated Zr-alloy, but it is limited by other primary system failures (e.g., hot-leg creep rupture) that will occur driven by core decay heat and independent of coated cladding effects. Finally, we observe that while the initial suppression of hydrogen generation for Cr-coated Zr-alloy clad compared to Zr-alloy is notable, the overall amount of hydrogen produced is similar since hydrogen can also be produced through competing oxidation of stainless steel components during the accident progression. Our future work is focused on the uncertainty analysis of the oxidation rate data, coating failure criteria, and severe accident modeling limitations in order to better quantify accident tolerant fuel clad benefits. [ABSTRACT FROM AUTHOR]

Steady-state internal flow boiling experiments were conducted on various materials, including accident tolerant fuel cladding material Fe–12Cr–6Al (C26M2) alloy, Zircaloy, and metal-based materials, at atmospheric pressure (84 kPa), 10°C inlet subcooling, and 200 kg/m2‧s mass flow entering the test tubes until critical heat flux (CHF) was reached. The clad thickness effects on flow boiling CHF were evaluated showing a negative relation between CHF and clad thickness up to 0.711 mm. An approach was established to mechanistically understand the measured CHF differences among the tested materials using thermal effusivity, activity, diffusivity, and surface thermal economy. No clear relations were observed within the range of thermal properties of the tested materials. Compared to past CHF data for a mass flux of 300 kg/m2‧s, the CHF data for 200 kg/m2‧s showed increased relative differences among materials. This result implies that higher mass flux may further decrease apparent material sensitivity to CHF. [ABSTRACT FROM AUTHOR]