FeCrAl Cladding Research Articles

Radial distributions of power and fuel temperature in annular U3Si2 fuel with FeCrAl cladding

The present work performs systematic study on neutronic properties of annular fuel due to its high efficiency in heat transfer and low peak fuel temperature. Recent physical experiment on annular fuel indicates the present neutronic study is urgent and necessary. The present work concentrates on the investigations on the reactivity and radial distributions of power and fuel temperature for various annular fuels with different moderator to fuel ratios and different inner to outer coolant ratios. An analytical formula fx,s including fuel exposure (s) and relative radial (x) is proposed to describe the radial power distributions. fx,s has a second order polynomial dependence on s with coefficients following a double exponential form versus x. Monte Carlo simulations show that the suggested function fx,s gives a nice description on simulation data with rather small deviations and can immediately provide radial distribution of power and burnup at any fuel exposure and relative radius. The present analytical formula can be directly used for determining radial fuel temperature distribution. Numerical results based on radial power distributions show that about 150 K peak fuel temperature is reduced by dual-coolant annular design. In addition, the radial fuel temperature distributions in annular fuels are quite flatter than that in cylindric fuel rod.

Full-core analysis for FeCrAl enhanced accident tolerant fuel in boiling water reactors

The impact of replacing Zircaloy with FeCrAl, a candidate enhanced accident-tolerant fuel cladding material, was evaluated for 10 × 10 boiling water reactor fuel bundles. Results from a series of full-core parametric studies estimated that replacing UO2/Zircaloy with UO2/FeCrAl would require an average enrichment increase of 0.6% 235U throughout the fuel lattice with the cladding and channel box thicknesses halved and fuel pellet diameter increased. Full-core results indicated that UO2/FeCrAl models with these geometric/enrichment specifications matched the base UO2/Zircaloy cycle length of 527 effective full power days. Optimization studies of the full-core design established loading and control blade patterns for both Zircaloy and FeCrAl models. A side study was conducted modeling a hybrid fuel bundle consisting of FeCrAl cladding and a SiC/Ni/Cr channel box. By halving the cladding thickness, the enrichment level required was less than that of the Zircaloy base case design after performing loading pattern optimization of the hybrid bundle core. Lastly, the thermomechanical performance of a Zircaloy-cladded fuel rod was compared to that of a FeCrAl system. Results from this analysis show that, if starting from the same fuel-cladding gap thickness, a FeCrAl-clad fuel rod operates with a greater average fuel centerline temperature, comparable axial elongation and radial displacement, and longer time to gap closure compared to a Zircaloy-clad fuel rod. This fuel performance analysis was primarily based on the commercial Kanthal APMT FeCrAl alloy but also used available data for the C35M FeCrAl alloy developed at Oak Ridge National Laboratory.

Research on thermal behaviors of several accident tolerant fuels based on 5 × 5 bundle subchannel model

Accident tolerant fuel (ATF) materials need to be studied by subchannel analysis under a variety of transient conditions before being applied to Light-Water-Reactors (LWRs). The present study assesses the thermal behaviors of various ATF claddings and fuels under four different transient conditions which will give rise to the occurrence of Departure from Nucleate Boiling (DNB) phenomenon. A 5 × 5 rod bundle model is built to conduct the subchannel analysis, and maximum cladding temperature (MCT) and maximum fuel centerline temperature (MFCT) of different ATF cladding and fuel groups are obtained. The simulation results show that both MCT and MFCT are reduced with the utilization of FeCrAl cladding, superior to HNLS/ML-A and SA3/PyC150-A claddings that only reduce MCT but increase MFCT. Significant MFCT reduction but slight MCT increase are achieved in most transient cases with the application of UO2 + BeO/SiC fuel pellets. Due to the different safety limit of various ATF materials, safety margin analysis is performed as well. The results show FeCrAl possibly is the best alternative cladding material among claddings studied, because both cladding safety margin (CSM) and fuel pellet safety margin (FPSM) are improved. Compared to single UO2 fuel, a larger fuel pellet safety margin is still provided by UO2 + BeO/SiC fuel pellets.

Radial distributions of power and isotopic concentrations in candidate accident tolerant fuel U3Si2 and UO2/U3Si2 fuel pins with FeCrAl cladding

Monte Carlo simulations show similarity on radial distributions of power and isotopic concentrations at any effective full power depletion time among five kinds of fuel-cladding combinations with the same cycle length, including the normal UO2-zircaloy combination, the candidate Accident Tolerant Fuel (ATF) UO2/U3Si2-FeCrAl combination, and three kinds of candidate ATF U3Si2-FeCrAl combinations. An analytical formula f(x,s) including the fuel exposure (s) and the relative radial (x) is proposed to describe the radial properties for all five kinds of fuel-cladding combinations. f(x,s) has the form of the second order polynomial term of s with the exponential type of coefficients depending on x. It is shown that the suggested function f(x,s) gives a nice description on the simulation data with rather small deviations and can immediately provide radial distribution of power, burnup, and isotopic concentrations of 235U, 238U, 239Pu, and 241Pu at any fuel exposure and relative radius. It is useful to discuss the fuel temperature through the present analytical formula. The realistic radial power distribution gives flatter radial temperature distribution compared with the uniform power distribution. Because of the different thermal conductivities of fuels and claddings and the different thicknesses of claddings, the present discussed five kinds of fuel-cladding combinations have different radial temperature distributions, although their radial power distributions are quite similar. The present work provides an analytical formula to describe the radial properties of the ATF which is expected to be helpful for further neutronic and multi-physics coupling studies.

Iron–Chromium–Aluminum (FeCrAl) Cladding Oxidation Kinetics and Auxiliary Feedwater Sensitivity Analysis—Short-Term Station Blackout Simulation of Surry Nuclear Power Plant

Accident tolerant fuels (ATF) and steam generator (SG) auxiliary feedwater (AFW) extended operation are two important methods to increase the coping time for nuclear power plant safety response. In light of recent efforts to investigate such methods, we investigate both FeCrAl cladding oxidation kinetics and SG AFW sensitivity analyses, for the Surry nuclear power plant Short-Term Station Blackout simulation using the MELCOR YR 1.8.6 systems code. The first part describes the effects of FeCrAl cladding oxidation kinetics. Zircaloy cladding and two different oxidation models of FeCrAl cladding are compared. The initial hydrogen generation time (>0.5 kg) is used as the evaluation criterion for fuel degradation in a severe accident. Results showed that the more recent oxidation correlation by ORNL predicts much less hydrogen generation than Zircaloy cladding. The second part investigates the effects of three different methods of AFW injection into the SG secondary side. We considered three different methods of water injection; i.e., constant water injection into the secondary side (case 1); water injection based on secondary side water level in boiler region (case 2); water injection based on secondary side water level in the downcomer region (case 3). The case of constant water injection is the most straightforward, but it would have the tendency to overfill the SG with excess water. Water injection with downcomer level control is more reasonable but requires DC power to monitor level and to control AFW injection rate.

Multiphysics modeling of accident tolerant fuel-cladding U3Si2-FeCrAl performance in a light water reactor

The performance of the proposed advanced fuel-cladding system U3Si2-FeCrAl was studied based on the suggestion that U3Si2-FeCrAl system is a potential accident tolerant fuel candidate from a neutronics assessment. The U3Si2 fuel was reported with a number of advanced thermophysical properties and FeCrAl alloy was reported with a better oxidation resistance and a larger thermal neutron absorption cross section. This study presents the development of a model for U3Si2-FeCrAl combination system by the modification of our CAMPUS code. Then the fuel performance was compared with UO2-Zircaloy, UO2-FeCrAl and U3Si2-Zircaloy systems. And finally sensitivity analysis has been performed on the fuel input parameters of UO2 fuel and cladding thickness of FeCrAl cladding. The U3Si2-Zircaloy system was found to decrease the fuel centerline temperature dramatically but with a very early gap closure time due to the large thermal expansion coefficient of U3Si2 fuel, while the U3Si2-FeCrAl system was found to not only decrease the fuel centerline temperature a lot, but also further delay the gap closure time, which would delay the pellet-cladding mechanical interaction time. Sequentially, the fission gas release and plenum pressure were also found to be decreased a lot. Then the nuclear reactor safety is expected to be improved by applying the U3Si2-FeCrAl system.

Thermophysical and Mechanical Analyses of UO2-36.4vol % BeO Fuel Pellets with Zircaloy, SiC, and FeCrAl Claddings

The thermophysical performance and solid mechanics behavior of UO2-36.4vol % BeO fuel pellets cladded with Zircaloy, SiC, and FeCrAl, and Zircaloy cladding materials coated with SiC and FeCrAl, are investigated based on simulation results obtained by the CAMPUS code. In addition, the effect of coating thickness (0.5, 1 and 1.5 mm) on fuel performance and mechanical interaction is discussed. The modeling results show that Zircaloy claddings are more effective in decreasing fuel centerline temperature and fission gas release than other kinds of cladding material because of the smaller gap between cladding and fuel at the same burnup. SiC claddings and SiC-coated Zircaloy claddings possess smaller plenum pressure than other kinds of cladding. SiC claddings contribute more to fuel radial displacement but less to fuel axial displacement. FeCrAl claddings exhibit very different radial and axial displacements in different axial positions. FeCrAl-coated Zircaloy claddings have a lower fuel centerline temperature than Zircaloy claddings at burnup below 850 MWh/kg U, but a higher fuel centerline temperature at higher burnup. The gap between FeCrAl-coated Zircaloy claddings and fuel pellets closes earlier than that of Zircaloy claddings. SiC-coated claddings increase fuel radial and axial displacements, and cladding axial displacements of inner and outer cladding surfaces.

Prediction of peak cladding temperature in a three-loop pressurised water reactor with accident-tolerant fuel during loss-of-coolant accident

Safety analysis of a PWR fuelled with ATF (Accident-Tolerant Fuel) has been performed at LB-LOCA condition. The ATF being used is uranium silicide (U3Si2) and FCMF (Fully Ceramic Microencapsulated Fuel) with silicon carbide (SiC) and FeCrAl alloy as a cladding material. The objective of this research is to obtain dynamic characteristics of ATF-fuelled PWR at LB-LOCA condition. RELAP5-3D system code was used to model the reactor and simulate the transient. A safe shutdown of the reactor was assumed after a depressurisation following a double-ended guillotine breach in the main pipe. The results of simulations show that during LB-LOCA with partially functioning ECCS, the transient PCTs were far below the maximum allowable limit. The use of ATF could decrease the maximum transient PCT. It is shown that U3Si2 fuel with FeCrAl cladding has the minimum PCT transient and the shortest quench time to steady state condition after transient initiation.

Fuel performance simulation of iron-chrome-aluminum (FeCrAl) cladding during steady-state LWR operation

Alternative cladding materials have been proposed to replace the currently used zirconium (Zr)-based alloys, in order to improve the accident tolerance of light water reactor (LWR) fuel. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys that exhibit much slower oxidation kinetics in high-temperature steam than Zr-alloys. This behavior should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. Within the development of these alloys, suitability for normal operation must also be demonstrated. This article is focused on modeling the integral thermo-mechanical performance of FeCrAl clad UO2 fuel during normal reactor operation. Finite element analysis has been performed to assess commercially available FeCrAl alloys (namely Alkrothal 720 and APMT) as a candidate fuel cladding replacement for Zr-alloys, using the MOOSE-based fuel performance code BISON. These simulations identify the effects of the mechanical-stress and irradiation responses of FeCrAl and provide a comparison with Zr-alloys. In comparing these cladding materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (∼4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. These power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. The fuel rod designs and operating conditions used here are based on the Peach Bottom BWR with representative GE-12/14 fuel geometries, and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. Individual sensitivity analyses of the fuel and cladding creep responses were also performed, which indicated the influence of compliance for each material, separately, on the stress state of the fuel cladding. These parametric analyses are performed using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history.

Neutron cross section sensitivity and uncertainty analysis of candidate accident tolerant fuel concepts

The aftermath of the Tōhoku earthquake and the Fukushima accident has led to a global push to improve the safety of existing light water reactors. A key component of this initiative is the development of nuclear fuel and cladding materials with potentially enhanced accident tolerance, also known as accident-tolerant fuels (ATF). These materials are intended to improve core fuel and cladding integrity under beyond design basis accident conditions while maintaining or enhancing reactor performance and safety characteristics during normal operation. To complement research that has already been carried out to characterize ATF neutronics, the present study provides an initial investigation of the sensitivity and uncertainty of ATF systems responses to nuclear cross section data. ATF concepts incorporate novel materials, including SiC and FeCrAl cladding and high density uranium silicide composite fuels, in turn introducing new cross section sensitivities and uncertainties which may behave differently from traditional fuel and cladding materials.In this paper, we conducted sensitivity and uncertainty analysis using the TSUNAMI-2D sequence of SCALE with infinite lattice models of ATF assemblies. Of all the ATF materials considered, it is found that radiative capture in 56Fe in FeCrAl cladding is the most significant contributor to eigenvalue uncertainty. 56Fe yields significant potential eigenvalue uncertainty associated with its radiative capture cross section; this is by far the largest ATF-specific uncertainty found in these cases, exceeding even those of uranium. We found that while significant new sensitivities indeed arise, the general sensitivity behavior of ATF assemblies does not markedly differ from traditional UO2/zirconium-based fuel/cladding systems, especially with regard to uncertainties associated with uranium.We assessed the similarity of the IPEN/MB-01 reactor benchmark model to application models with FeCrAl cladding. We used TSUNAMI-IP to calculate similarity indices of the application model and IPEN/MB-01 reactor benchmark model. This benchmark was selected for its use of SS304 as a cladding and structural material, with significant 56Fe content. The similarity indices suggest that while many differences in reactor physics arise from differences in design, sensitivity to and behavior of 56Fe absorption is comparable between systems, thus indicating the potential for this benchmark to reduce uncertainties in 56Fe radiative capture cross sections.

Potential impact of accident tolerant fuel cladding critical heat flux characteristics on the high temperature phase of reactivity initiated accidents

The present paper details scoping RELAP5-3D simulations to assess the potential impact of boiling heat transfer coefficients and critical heat flux (CHF) on peak cladding temperature (PCT) of key accident tolerant fuel (ATF) cladding types in a design basis accident (DBA). The accident event studied is a Hot Zero Power (HZP) Reactivity-Initiated Accident (RIA) in a Pressurized Water Reactor (PWR), so the present study is focused on bubble crowding CHF. Results are presented for the sensitivity of PCT and the duration time of film boiling to the key parameters in the boiling curve: nucleate, transition, and film boiling heat transfer coefficient, and CHF. This included a generic study of perturbations in boiling heat transfer characteristics for Zircaloy-4 cladding as well as a study which focused on the key candidate ATF cladding types. For Zircaloy-4 cladding, the simulation results show that both PCT and duration of film boiling are particularly sensitive to the value of CHF. We also compared the performance of FeCrAl claddings and SiC/SiC claddings with Zircaloy-4 cladding under the same RIA scenario. A 10% CHF enhancement for FeCrAl cladding can significantly reduce the PCT of FeCrAl claddings. Lower irradiated thermal conductivity of SiC/SiC claddings does cause a high temperature gradient in the radial direction.

On the Minimum Thickness of FeCrAl Cladding for Accident-Tolerant Fuel

The thickness of iron-based accident-tolerant fuel (ATF) cladding is discussed in this technical note because its thickness tends to be reduced from conventional zirconium alloy cladding. Structural stability may be lost if the thickness cannot withstand external pressure. Thus, the minimum allowable thickness of the ATF cladding is studied here from the viewpoint of preventing a cladding collapse. The elastic buckling theory is used to obtain the minimum thickness. The uncertainties of the mechanical properties and dimension tolerances are taken into consideration. The ovality of the cladding is also incorporated. An example calculation is carried out for APMT cladding. It is evaluated that the minimum thickness is 0.45 mm when the safety factor against the buckling is set as 2.0 and 1% of the cladding radius is accommodated for the ovality. A reference guideline of the minimum thickness depending on the mechanical property variation is suggested.

Solid-liquid phase equilibria of Fe-Cr-Al alloys and spinels

Ferritic FeCrAl alloys are candidate accident tolerant cladding materials. There is a paucity of data concerning the melting behavior for FeCrAl and its oxides. Analysis tools have therefore had to utilize assumptions for simulations using FeCrAl cladding. The focus of this study is to examine in some detail the solid-liquid phase equilibria of FeCrAl alloys and spinels with the aim of improving the accuracy of severe accident scenario computational studies.

An investigation of FeCrAl cladding behavior under normal operating and loss of coolant conditions

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Further experiments are required to confirm these observations.

Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO2–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications made to a version of the MELCOR code by the Idaho National Laboratory (INL) that allows its application to alternate fuel-cladding systems. Preliminary analysis results are presented in this paper for candidate silicon carbide and FeCrAl cladding concepts.

Neutronic Analysis on Potential Accident Tolerant Fuel-Cladding Combination U3Si2-FeCrAl

Neutronic performance is investigated for a potential accident tolerant fuel (ATF), which consists of U3Si2fuel and FeCrAl cladding. In comparison with current UO2-Zr system, FeCrAl has a better oxidation resistance but a larger thermal neutron absorption cross section. U3Si2has a higher thermal conductivity and a higher uranium density, which can compensate the reactivity suppressed by FeCrAl. Based on neutronic investigations, a possible U3Si2-FeCrAl fuel-cladding system is taken into consideration. Fundamental properties of the suggested fuel-cladding combination are investigated in a fuel assembly. These properties include moderator and fuel temperature coefficients, control rods worth, radial power distribution (in a fuel rod), and different void reactivity coefficients. The present work proves that the new combination has less reactivity variation during its service lifetime. Although, compared with the current system, it has a little larger deviation on power distribution and a little less negative temperature coefficient and void reactivity coefficient and its control rods worth is less important, variations of these parameters are less important during the service lifetime of fuel. Hence, U3Si2-FeCrAl system is a potential ATF candidate from a neutronic view.

Iron-chrome-aluminum alloy cladding for increasing safety in nuclear power plants

After a tsunami caused plant black out at Fukushima, followed by hydrogen explosions, the US Department of Energy partnered with fuel vendors to study safer alternatives to the current UO2 -zirconium alloy system. This accident tolerant fuel alternative should better tolerate loss of cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. General electric, Oak ridge national laboratory, and their partners are proposing to replace zirconium alloy cladding in current commercial light water power reactors with an iron-chromium-aluminum (FeCrAl) cladding such as APMT or C26M. Extensive testing and evaluation is being conducted to determine the suitability of FeCrAl under normal operation conditions and under severe accident conditions. Results show that FeCrAl has excellent corrosion resistance under normal operation conditions and FeCrAl is several orders of magnitude more resistant than zirconium alloys to degradation by superheated steam under accident conditions, generating less heat of oxidation and lower amount of combustible hydrogen gas. Higher neutron absorption and tritium release effects can be minimized by design changes. The implementation of FeCrAl cladding is a near term solution to enhance the safety of the current fleet of commercial light water power reactors.

The potential impact of enhanced accident tolerant cladding materials on reactivity initiated accidents in light water reactors

Advanced cladding materials with potentially enhanced accident tolerance will yield different light-water reactor performance and safety characteristics than the present zirconium-based cladding alloys. These differences are due to cladding material properties, reactor physics, and thermal hydraulics characteristics. Differences in reactor physics are driven by the fundamental properties (e.g., neutron absorption cross section in iron for an iron-based cladding) and also by design modifications necessitated by the candidate cladding materials (e.g., a larger fuel pellet to compensate for parasitic absorption).This paper describes three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative pressurized water reactor with both iron-chromium-aluminum (FeCrAl) and silicon-carbide fiber silicon carbide ceramic matrix composite (SiC/SiC) materials. This study shows similar RIA neutronic behavior for SiC/SiC cladding configurations versus reference Zircaloy cladding. However, the FeCrAl cladding response indicates similar energy deposition but with shorter pulses of higher magnitude. This is due to the shorter neutron generation time of the core models based on FeCrAl cladding. The FeCrAl-based cases exhibit a more rapid fuel thermal expansion rate than other cases, and the resultant pellet-cladding interaction may occur more rapidly. The conclusions in this paper are based on a limited set of simulated super prompt RIA transients.

Neutronics and fuel performance evaluation of accident tolerant FeCrAl cladding under normal operation conditions

Neutronics and fuel performance analysis is done for enhanced accident tolerance fuel (ATF), with the Monte Carlo reactor physics code Serpent and INL’s fuel performance code BISON. The purpose is to evaluate the most promising ATF candidate material FeCrAl, which has excellent oxidation resistance, as fuel cladding under normal operational conditions.Due to several major disadvantages of FeCrAl coating, such as difficulty in fabrication, diametrical compression from reactor pressurization, coating spallation and inter diffusion with zirconium, a monolithic FeCrAl cladding design is suggested. To overcome the neutron penalty expected when using FeCrAl as cladding for current oxide fuel, an optimized FeCrAl cladding design from a detailed parametric study in literature is adopted, which suggests reducing the cladding thickness and slightly increasing the fuel enrichment. A neutronics analysis is done that implementing this FeCrAl cladding design in a Pressurized Water Reactor (PWR) single assembly. The results show that the PWR cycle length requirements will be matched, with a slight increase in total plutonium production.Fuel performance analysis with BISON code is carried out to investigate the effects with this FeCrAl cladding design. The results demonstrate that the application of FeCrAl cladding could improve performance. For example, the axial temperature profile is flattened. The gap closure is significantly delayed, which means the pellet cladding mechanical interaction is greatly delayed. The disadvantages for monolithic FeCrAl cladding are that: (1) fission gas release is increased; and (2) fuel temperature is increased, but the increase is less than 50K even at high burnup.The better strength, corrosion, and embrittlement properties of FeCrAl enable the fabrication of FeCrAl cladding with thinner walls. FeCrAl cladding proves to be a good alternate for zircaloy cladding, given the advantages and insignificant disadvantages shown by fuel performance analysis.