Reactivity Initiated Accident Conditions Research Articles

A study on the fracture pattern change of high-burnup fuel cladding failed by pellet-cladding mechanical interaction failure under reactivity-initiated accident conditions

ABSTRACT In a part of the historical reactivity-initiated accident (RIA) simulated tests on high-burnup fuels performed at the Nuclear Safety Research Reactor, the fuel failure caused by pellet-cladding mechanical interaction (PCMI) led to splitting into upper- and lower-part pieces or even fragmentation of the cladding tube. A massive release of fuel fragments accompanied this fracture pattern change from previously known axial cracks and thus identified as a potential concern in safety evaluation regarding core coolability. Dedicated out-of-pile mechanical tests were performed with unirradiated Zircaloy-4 cladding specimens to clarify the condition of such fracture pattern change. The specimens were pre-hydrided and subjected to loading with axial-to-hoop strain ratios of ℇz/ℇθ = 0.5–1.25, simulating the effects of hydrogen embrittlement and pellet-cladding mechanical bonding of high-burnup fuels, respectively. The results indicate that higher biaxiality of the loading and lower ductility (failure strain level) assist the fracture pattern change. This study proposes a conservative criterion that a PCMI failure splits the cladding tube into more than two pieces when strain ratio ℇz/ℇθ >0.75 and a concurrent hoop strain < 1.7% at the failure instant.

Thermal-Mechanical performance analysis of a petal-shaped fuel rod under RIA condition

This paper establishes the performance analysis model of a petal-shaped fuel rod (PSFR) under Reactivity Initiated Accident (RIA) condition based on thermal elastic–plastic theory and finite element method. The results show that the temperature changes of fuel and cladding are basically synchronized, and the temperature extremes appear in the central region, with the distribution characterized by a large middle and two small sides. The stress is concentrated in the center region of the cladding. At the beginning of operation, the stress at the lobe root of cladding is the largest. With the increase of operation time, the stress concentration at the lobe tip is gradually obvious. At 150th days of operation, a sudden RIA occurred and the power factor of the PSFR rapidly increased over 5.38 times and the cladding stress exceeded the yield strength. The axial and circumferential center cladding lobe roots and nearby may be at greater risk of plastic deformation. These results will provide an important theoretical basis for the determination of the deformation and accident operation risk of PSFRs.

High-temperature rupture failure of high-burnup LWR-MOX fuel under a reactivity-initiated accident condition

A reactivity-initiated accident (RIA)-simulated test CN-1 on a high-burnup 64 GWd/t mixed-oxide fuel rod sheathed with M5™ cladding was conducted at the Nuclear Safety Research Reactor, resulting in fuel failure. A small opening with slight ballooning deformation characterized the post-test visual appearance of the test fuel rod. Simulation using fuel performance codes FEMAXI-8/RANNS predicted rod survival under early phase loading induced by pellet-cladding mechanical interaction and subsequent boiling transition, and the cladding-surface temperature measured online confirmed the occurrence of boiling transition. The experimental observation and simulation indicate that the failure was caused by a high-temperature rupture following increased rod-internal pressure. The RANNS sensitivity analysis revealed that a mechanical state parameter dedicated to predicting plastic instability might be an effective index for evaluating the risk of rupture failure during RIAs.

Numerical simulation of mechanical interaction between pellet and zircaloy-4 cladding under reactivity-initiated accident conditions

Reactivity-initiated accident (RIA) is one of the accidents considered by fuel safety criteria. The power pulse caused abrupt expansion of the pellet, leading to pellet and cladding mechanical interaction (PCMI). Cladding ductility is gradually decreased in the process of oxidation and hydrogen uptake during irradiation. As a result, the extrusion of the pellet to the cladding causes the cladding to produce greater hoop stress and hoop strain, which leads to damage. A simulation of PCMI using ABAQUS was designed to analyze cladding mechanic properties. By comparing the calculation results of the simulation scheme with the data obtained in the reactor experiment, it is proved that the simulation calculation results are in good agreement with the experimental data in the reactor, and the rationality of the simulation scheme is verified. Based on the out-pile mandrel equipment simulated PCMI, the corresponding finite element model was established to study related test parameters.By analyzing the results of the simulation calculation outside the reactor, it is found that the pellet material, cladding temperature, compression rate, tensile load, friction coefficient and the cladding hoop stress and hoop strain have a certain relationship: (1) The choice of pellet material affects the reactor. The overall law of hoop stress-strain in the external simulation results basically has no effect. (2) The temperature of the cladding has effect on the hoop stress peak. (3) The compression rate has effect on the time for the cladding to reach the hoop peak stress. (4) Tensile load will affect the starting position of the hoop stress-strain of the cladding. (5) The friction coefficient has a significant influence on the stress and strain trend of the cladding. The outside experiment can be carried out more effectively, and it can provide guidance and a basis for the outside experiment.

CAMPUS-ANNULAR: Preliminary multiphysics modeling of annular fuel performance under normal operating and accident conditions

The internal and external cooled annular fuel was proposed firstly by MIT, which is found to be able to improve fuel performance dramatically. Based on the CAMPUS code, an annular fuel performance code called CAMPUS-ANNULAR is further developed in this work. And the multiphysics models of annular UO2 fuel equipped with Zircaloy cladding are developed under reactor normal operating, Loss Of Coolant Accident (LOCA) and Reactivity Initiated Accident (RIA) conditions, respectively. Then the BISON cladding failure criterion utilized under LOCA condition is introduced, and a loss of coolant and reflood model is implemented under LOCA condition for the first time. Additionally, the radial power distribution of annular fuel is introduced and analyzed in this work. Finally, the models implemented in this work are verified and the annular fuel performance under normal operating, LOCA and RIA conditions are investigated and compared with those of solid fuel. The fuel maximum temperature is found to be reduced dramatically by using annular fuel under normal operating, LOCA and RIA conditions, which may further lead to the cladding failure time to be delayed significantly under LOCA condition. Besides, the risk of fuel melting is found to be reduced by using annular fuel under RIA condition.

Behavior of high-burnup BWR UO2 fuel with additives under reactivity-initiated accident conditions

ABSTRACT Fuels with additives are expected to provide enhanced fuel performance in fission gas retention owing to their large grain size, which elongates fission gas migration path. To investigate behavior of the fuels during a reactivity-initiated accident (RIA), RIA-simulated experiments OS-1 and LS-4 were performed on ADOPT (chromia- and alumina-doped UO2) fuel of 64 GWd/t and chromia-doped UO2 fuel of 48 GWd/t, respectively. The OS-1 rod failed at a fuel enthalpy increase of 160 J/g due to pellet–cladding mechanical interaction failure, which was the lowest failure limit among the test results ever obtained at the NSRR on high-burnup fuels from 40 to 65 GWd/tU. Comparison of the hydride morphologies in the cladding metallic layer between the rods subjected to the past NSRR tests suggests the contribution of radially oriented hydrides during base irradiation to the low failure limit. The LS-4 rod survived for a peak fuel enthalpy increase of 549 J/g, which resulted in cladding deformation of ~2.4% in the residual hoop strain and FGR of 1.4–6.1%. Whereas the low fission gas release exhibits the effect of additives, the cladding deformation is within the range explained by the deformation mechanism essentially identical to those recognized for high-burnup undoped fuels.

An Improved Steady-State and Transient Analysis of the RSG-GAS Reactor Core under RIA Conditions Using MTR-DYN and EUREKA-2/RR Codes

Steady-state and transient analysis of reactor core under Reactivity-Initiated Accident (RIA) conditions are important for reactor operation safety. The reactor dynamics are influenced by neutronic and thermal-hydraulic aspects of the core. In this study, steady-state and transient analysis under RIA conditions of the RSG-GAS multipurpose reactor was carried out using MTR-DYN and EUREKA-2/RR programs. Neutronic calculations were performed using a few group cross-sections generated by Serpent 2 with the latest cross-section data ENDF/B-VIII.0. Steady-state conditions were carried out with a nominal power of 30 MW, while transient under RIA conditions occurred because the control rod was pulled too quickly while the reactor operated. These transient RIA conditions were performed for two cases, during start-up with an initial power of 1 W, and within power range with an initial power of 1 MW. Thermal-hydraulic parameters considered in this study are reactor power, the temperature of the fuel, cladding, and coolant. The calculated maximum fuel temperature at a steady state is 126.02°C. Meanwhile, the calculated maximum fuel temperature during RIA conditions at the initial power of 1 W and 1 MW are 64.38°C and 137.14°C, respectively. There are no significant differences in thermal-hydraulic parameters between each used program. The thermal-hydraulic parameters such as the maximum temperature of the coolant, cladding, and fuel under this postulated RIA condition are within the acceptable reactor operation safety limits.

Evaluation of anisotropic elastic and plastic parameters of zircaloy-4 fuel cladding from biaxial stress test data and their application to a fracture mechanics analysis

ABSTRACT The mechanical properties of fuel cladding near the elastic limit are essential in considering its failure limit during a pellet-cladding mechanical interaction phase under reactivity-initiated accident (RIA) conditions. The mechanical properties of a Zircaloy-4 cladding tube, such as orthotropic elasticity and anisotropic constants for Hill’s plasticity law, were evaluated based on the biaxial stress test data, focusing on the equivalent plastic strain up to ~2.5%. Samples with various fabrication conditions, such as cold-worked, recrystallized, and stress relieved after cold-work with Q-factors of 2, 3, and 4 were investigated. The evaluation revealed noticeable difference in the elastic and plastic properties between cold-worked, recrystallized and stress relieved samples, including their anisotropic feature, while the impact of Q-factors was minor. The derived mechanical properties were applied to calculation of the fracture mechanics parameter, J-integral, which was found sensitive to the mechanical property differences. Critical J-integral values were then estimated by comparing the calculation and failure limit data from biaxial-expansion-due-to-compression tests on precracked tubes. This produced systematically lower J-integral values of the stress relieved tube than those previously evaluated based on the failure limit data from in-pile RIA-simulated tests.

BISON validation of FeCrAl cladding mechanical failure during simulated reactivity-initiated accident conditions

A reactivity-initiated accident (RIA) is a postulated design basis accident in light water reactors (LWRs) in which a rapid reactivity insertion induces a fission rate increase and a fuel pellet temperature rise. During RIA, the fuel pellet thermally expands and may cause pellet-cladding mechanical interaction (PCMI). Separate effects PCMI tests were performed on C26M FeCrAl cladding tube samples, introducing biaxial stress via a well understood modified burst test (MBT) system. A high-speed camera in the MBT system captured the projections, covering a 360° view of the cladding deformation and enabling a digital image correlation (DIC) method to quantify the surface strains with high fidelity. Representative hot zero-power RIA mechanical loading conditions were applied to the sample, and the test duration ranged from 20 to 500 ms at an average temperature of 573 K. BISON finite element–based fuel performance code modeling was performed against the high-fidelity DIC data produced from the MBTs. Validation calculations were conducted with 2D models and systematically compared with test data of the burst time, burst pressure, burst hoop strains, and hoop strain rates. Based on the behaviors from the separate effects test, BISON calculations satisfactorily predicted the cladding deformation behaviors. Sensitivity analysis was conducted to identify highly influential mechanical properties responsible for the cladding failure behavior during the MBT experiments. The results highlight the significance of the cladding's mechanical strength in governing cladding strain, followed by the significance of the mechanical interactions between the pellet and the cladding.

FEMU based identification of the creep behavior of Zircaloy-4 claddings under simulated RIA thermo-mechanical conditions

This study aims to identify the creep behavior of Zircaloy-4 fuel claddings under simulated reactivity initiated accident (RIA) thermo-mechanical conditions. In a previous work ballooning creep tests performed in simulated RIA conditions were described. A challenge to overcome when analyzing these experiments is that ballooning tests imply structural effects during the deformation of the specimen that have to be taken into account to identify the creep behavior of the material. In this paper a FEMU (finite element model updating) based identification is proposed to identify the creep law of the cladding, weakly coupled to the phase transformation of the material at high temperatures. Since the loading conditions are solely known in the part of the optical field of the cameras used for stereocorrelation, only this region is modeled using a Love-Kirchhoff assumption to impose the boundary conditions through the sample thickness. A Norton creep law, whose parameters are expressed as a function of the β phase fraction in the material, is identified and reproduces the first 10 s of the experiments with mean errors on the radial displacement rates of about 10%. Finally, an extension to higher time scale is proposed in Norton’s law to model the non linearity in the material response by taking into account the grain growth contribution.

Post-transient examination of performance of uranium silicide fuel and silicon-carbide composite cladding under reactivity-initiated accident conditions

• First transient testing of accident-tolerant fuel designs under transient nuclear heating reactivity-initiated accident conditions. • Claddings experienced dimensional changes but maintained primary rod-like geometries. • Elemental species diffusion from the U 3 Si 2 fuel to the zircaloy cladding was observed. • No elemental species diffusion from the U 3 Si 2 fuel to the silicon-carbide cladding was observed. • Through wall thickness cracking observed in the silicon-carbide cladding resulting from pellet cladding interaction from thermal expansion of the pellet. New fuels are being developed for light water nuclear reactors with the initial goal of improving accident tolerance. More recently, these new fuel systems are also being recognized for their potential to support higher fuel burnups and thus improved economics for reactor plants through improved fuel utilization. Fuel safety testing of these new fuels systems is being conducted under transient irradiation conditions in the Transient Reactor Test facility at Idaho National Laboratory. A test campaign including two rodlets of U 3 Si 2 fuel in Zircaloy-4 cladding and two rodlets of U 3 Si 2 fuel in Silicon-Carbide Composite cladding has been performed to begin the development of fuel-safety criteria in design-basis reactivity-initiated accident conditions. Post-Transient-Irradiation Examinations have been performed and found that the accident tolerant materials maintained their primary rod-like geometry. Elemental species diffusion was found to occur in the Zircaloy rodlets indicating rapid diffusion kinetics at the temperatures achieved. No elemental species diffusion was found in the Silicon-Carbide rodlets. However, through-wall thickness cracks were observed due to displacement loading of the cladding from thermal expansion of the fuel pellets. This work presents the first experimental evidence of the performance of these fuel systems under transient nuclear heating and representative reactivity-initiated accident energy depositions.

Multiphysics Modeling of Thorium-Based Fuel Performance With Cr-Coated SiC/SiC Composite Under Normal and Accident Conditions

Using the finite element multiphysics modeling method, the performance of the thorium-based fuel with Cr-coated SiC/SiC composite cladding under both normal operating and accident conditions was investigated in this work. First, the material properties of SiC/SiC composite and chromium were reviewed. Then, the implemented model was simulated, and the results were compared with those of the FRAPTRAN code to verify the correctness of the model used in this work. Finally, the fuel performance of the Th0.923U0.077O2 fuel, Th0.923Pu0.077O2 fuel, and UO2 fuel combined with the Cr-coated SiC/SiC composite cladding and Zircaloy cladding, respectively, was investigated and compared under both normal operating and accident conditions. Compared with the UO2 fuel, the Th0.923U0.077O2 and Th0.923Pu0.077O2 fuels were found to increase the fuel centerline temperature under both normal operating and reactivity-initiated accident (RIA) conditions, but decrease the fuel centerline temperature under loss-of-coolant accident (LOCA) condition. Moreover, compared to the UO2 fuel with the Zircaloy cladding, thorium-based fuels with Cr-coated SiC/SiC composite cladding were found to show better mechanical performance such as delaying the failure time by about 3 s of the Cr-coated SiC/SiC composite cladding under LOCA condition, and reducing the plenum pressure by about 0.4 MPa at the peak value in the fuel rod and the hoop strain of the cladding by about 16% under RIA condition.

Computational fluid dynamics assessment of the Helium-3 Enhanced Negative Reactivity Insertion (HENRI) system for the Transient REActor Test (TREAT) facility

The Transient REActor Test facility (TREAT) is an air-cooled, graphite moderated nuclear reactor that is used for irradiation testing of nuclear fuels and materials. The shutdown of the Halden reactor in Norway lead to an increased interest of an indigenous nuclear facility that can perform Reactivity Initiated Accident (RIA) tests for Light Water Reactors (LWRs). Idaho National Laboratory (INL) has been leading the effort of making minor modifications to the TREAT facility to enhance its testing capabilities so it can mimic the neutronic boundary conditions representative of a RIA for a LWR. The power burst of the TREAT reactor needs to be shortened from 89 to 40 msec. to properly irradiate the fuel sample under RIA conditions. INL has proposed to use helium-3, a strong neutron absorber, as a form of negative reactivity to shorten the pulse of the reactor. In support of the development of a transient control rod for the TREAT, a Computational Fluid Dynamics (CFD) model is needed to confidently simulate the density evolution of helium-3 in this system.

Transient testing of uranium silicide fuel in zircaloy and silicon carbide cladding

Using the newly restarted Transient Reactor Test (TREAT) facility at Idaho National Laboratories, fresh (unirradiated) U3Si2 fuel pellets were tested under reactivity-initiated accident conditions in dry helium capsules. These tests were performed with total energy depositions sufficient to induce melting of the U3Si2 fuel pellets based on available thermophysical property data. The instrumented experiments utilized in-pile thermocouples and infrared pyrometers to measure the cladding temperature during the transient, which are in reasonable agreement with predicted peak cladding temperatures using a simple BISON fuel performance model. This same model is then used to predict expected cladding stresses during the transients in areas where pellet cladding interaction is expected. It was shown that for transient energy depositions up to 528 J/g rod like geometry of the test articles was maintained. While it is acknowledged that more confirmatory testing is required in more prototypic environments, these first of their kind experiments have demonstrated that unirradiated U3Si2 fuel and unirradiated SiC-SiC claddings can likely retain a “rod like” geometry for transient energy depositions up to 528 J/g.

Fission gas release from irradiated mixed-oxide fuel pellet during simulated reactivity-initiated accident conditions: Results of BZ-3 and BZ-4 tests

Fission gas release from high-burnup mixed-oxide (MOX) fuel pellet with Pu fissile contents of 4.1% during reactivity-initiated accident (RIA) was estimated based on radial Xe concentration profiles in fuel pellet using scanning electron microscopy/electron probe microanalysis (SEM/EPMA) before and after pulse irradiation tests called BZ-3 and BZ-4. The tests were conducted at the Nuclear Safety Research Reactor (NSRR) in Japan Atomic Energy Agency (JAEA). The depression of Xe concentration in the Pu-spot and the crack formation in the peripheral region of fuel pellet were observed after BZ-3 and BZ-4. This kind of cracks was considered to promote the release of the fission gas retained in the Pu-spot. Based on the EPMA results and the effect of the crack formation, the amounts of the release of the gas retained in Pu-spot was estimated to be 0.13 and 0.04 in BZ-3 and BZ-4 tests, respectively. Moreover, taking account of the release of the gas accumulated in the grain boundary in Pu-spot-excluded region and the release of the gas dissolved in the matrix, the total amount of fission gas release was evaluated as 0.33–0.43 and 0.15–0.27 in BZ-3 and BZ-4 tests, respectively. Comparing the results of the fission gas release with the results of rod-puncture test, it was suggested that fission gas release from not only the Pu-spot but also the Pu-spot-excluded region.

Mechanical failure of high-burnup fuel rods with stress-relieved annealed and recrystallized M-MDA cladding under reactivity-initiated accident conditions

ABSTRACT To evaluate the effects of the hydride morphology and initial temperature of fuel cladding on the pellet-cladding mechanical interaction failure under reactivity-initiated accident (RIA) conditions, RIA-simulated experiments were performed on high-burnup fuels with stress-relieved annealed (SR) and recrystallized (RX) M-MDATM cladding at room and high (~280°C) temperatures. The results demonstrated that the failure-limit trend of RX-cladded fuels being lower than that of SR-cladded fuels for a similar hydrogen content holds up to at least about 700 wtppm. The observation of the fracture surfaces of failed RX cladding suggests a contribution of radially oriented hydrides to the crack formation and/or penetration, which coincides with the aforementioned failure-limit trend. The temperature effect, namely the failure-limit rise at a high temperature, is evident irrespective of the hydride morphology, while the degree of the temperature effect decreases as the hydrogen content increases.

Mechanical failure of fresh nuclear grade iron–chromium–aluminum (FeCrAl) cladding under simulated hot zero power reactivity initiated accident conditions

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing iron-chromium-aluminum (FeCrAl) cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens’ outer surfaces. Experiments were conducted at approximately 275 °C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8–3.4%, corresponding to quasi-static energy depositions of approximately 110–260 calories per gram UO2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the FeCrAl cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.

Evaluation of Radiation Effects on Residents Living Around the NSRR Under External Hazards

Abstract The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics—annular core pulse reactor (TRIGA-ACPR) type, located in the Nuclear Science Research Institute (NSRI). The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident (RIA) conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. The graded approach is introduced in the regulation. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. There is fuel in the reactor core, fresh fuel storage, and spent fuel storage. As the effects from reactor core, we evaluate the external exposure to radiation and exposure from the release of fission products assuming that loss of function to shut down the reactor, break of cladding tubes, loss of reactor pool water, and collapse of the reactor building. As the effects from fresh fuel storage, we evaluate the internal exposure assuming that the fresh fuel particles released into the air because of breaking into pieces. In addition, we evaluate the critical safety assuming that the critical safety shapes of the fresh fuel storage are lost. As the effects from spent fuel storage, we evaluate the critical safety assuming that the critical safety shapes of the spent fuel storage are lost. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions caused by the tornado, earthquake, and the accompanying tsunami.

Simulation of reactivity initiated accident thermal transients on nuclear fuels with laser remote heating

The investigation on the behaviour of nuclear fuels under thermal loads representative of RIA (Reactivity Initiated Accident) conditions in nuclear reactors is a significant challenge. In this work we introduce a new concept of a laboratory experiment relying on high power lasers to reproduce the temperature gradients experienced by nuclear fuel pellets during such events. The concept is experimentally demonstrated on UO2 samples, supported with numerical simulations to estimate the temperature gradients within the samples. This is a first step in the perspective of applying the technique on irradiated fuels.

Fuel performance optimization of U3Si2-SiC design during normal, power ramp and RIA conditions

Benefited from its higher mechanical strength and much-improved oxidation resistance at very high temperatures (>1200 °C), SiC cladding is one of the promising accident tolerant fuel (ATF) concepts. Considering the thermal conductivity degradation of SiC cladding after irradiation and its negligible creep rate, adopting conventional UO2 with SiC cladding can lead to aggravated fission gas release and elevated fuel centerline temperatures. In this work, U3Si2 fuel with its high thermal conductivity is considered in place of UO2, when adopting SiC cladding. First, the preliminary full core simulation for U3Si2-SiC was conducted to show its holistic fuel performance compared to the current UO2-Zircaloy fuel. Then the most limiting fuel rod was selected based on the calculated chemically vapor deposited (CVD) SiC failure risk from the full core simulation. For the limiting rod, fuel-clad gap optimization was conducted to minimize CVD SiC failure risk. Subsequently, power ramp history was appended to the normal power history and full core simulation analysis for updated U3Si2-SiC design was re-performed to ensure the effectiveness of the chosen gap width. For the limiting rod, reactivity-initiated accident (RIA) analysis at hot zero power (HZP) was analyzed to evaluate its performance and infer its potential failure modes. The results from normal operation and power ramp simulations implied that CVD SiC failure risk was mainly induced by the strong pellet-cladding mechanical interaction (PCMI) and the failure probability of CVD SiC can be effectively minimized to 2.0×10-7 through enlarging the fuel-clad as-fabricated gap width. The leading failure mode for U3Si2-SiC fuel system during RIA changed from fuel melting to Ceramic Matrix Composite (CMC) failure with increasing burnup.