Neutronic Parameters Research Articles

Evaluation of dynamical behavior of Thorium fueled SMART reactor using lumped parameter dynamic model

The point reactor dynamic model is based on the lumped parameter model and, unlike the traditional reactor kinetics model, takes into account the effects of the fuel and moderator temperature on the system reactivity. This model is a system of nonlinear and nonhomogeneous differential equations and is used in the transient analysis of an initially critical system subjected to any perturbation. In this study, transient analysis of the Th-fueled SMART reactor is performed by the point dynamics model. Initially, the reactor is kept critical using the reactivity control mechanisms such as control rods insertioın and soluble boron addition to the moderator. The required neutronic parameters such as precursors’ parameters, neutron generation time, and reactivity coefficients are calculated using the Serpent 2 Monte Carlo code. The moderator average temperature and system pressure are used as two distinct thermodynamic properties to determine thermal-hydraulic parameters such as heat capacity, conductivity coefficient, convection coefficient, and fuel thermal resistance. For different reactivity insertion scenarios, MATLAB ODE45 suite is used to solve the point dynamic equations. The validity of the provided code is also verified either through comparison with the analytically solvable form of the point dynamic model or by analyzing the asymptotic equilibrium values of the considered parameters. Finally, it is seen that when an initially critical reactor operating at a certain power level undergoes a perturbation, after a short time the system becomes again critical and operates at an almost stable power level different from the pre-perturbation power level. This, in turn, implies that the calculated neutronic and thermal-hydraulic parameters are physically meaningful.

Development of fuel depletion code for molten salt reactor with very deep burnup

A liquid-fueled molten salt reactor (MSR) can reach a deep burnup based on online reprocessing and continuously refueling, which requires significantly different burnup calculation methods for MSRs compared with those for the traditional reactors. To address the unique burnup features and consider the fidelity of isotopic evolution in an MSR, a fuel depletion code ThorMCB is developed based on the OpenMC coupled with a specific depletion code, MODEC. Furthermore, to lower the computational cost of acquiring the equilibrium state through the time evolution step by step for an MSR, an equilibrium burnup calculation code ThorMCB-eq based on the OpenMC and MODEC is developed, which can obtain the equilibrium burnup efficiently. A single fuel lattice of MSR and an a Molten Salt Fast Reactor (MSFR) benchmark are applied for verifying the correctness of the ThorMCB and ThorMCB-eq codes. Compared with a neutron transport calculation code KENO-VI coupled with MODEC, the maximum deviation of the dominant heavy nuclides (HNs) at equilibrium state by ThorMCB is less than 10%, and that of the total mass of fission products (FPs) is less than 3%. For the MSFR benchmark, the neutronic parameters including temperature reactivity coefficient, the mass evolution of main HNs and FPs and breeding ratio (BR) from ThorMCB agree with the references. The equilibrium behavior can be quickly obtained with ThorMCB-eq, and the relative mass deviations of most nuclides keep around 2% in comparison with the results of step-by-step burnup evolution with ThorMCB. Furthermore, the same fuel contents and micro one-group cross sections at equilibrium are obtained with two different types of start-up fuels and a constant power density and fuel reprocessing scheme. In conclusion, the verified results indicate that ThorMCB and ThorMCB-eq can both provide reliable simulation for depletion evolution and equilibrium burnup for MSR fuel cycle.

Generating optimal reloading patterns for CNPGS Unit-3 core using multi-objective elitist teaching-learning-based optimizer

Pressurized Water Reactors (PWRs) management presents the core reloading pattern optimization as a significant problem that contributes to the improvement of reactor productivity and optimal fuel utilization with considering safety conditions. In present study, Multi-Objective Elitist Teaching-Learning-Based Optimization (MO-ETLBO) technique is suggested to cope up the multi-objective reloading optimization problem of the Chashma Nuclear Power Generating Station (CNPGS) unit-3 core. A multivariable objective function is designed to evaluate the quality of each loading pattern while maximizing critical boron concentration (CBC), minimizing the power peaking factor (PPF), and optimally enhancing the cycle length while ensuring adequate safety margins and design limits. It has been found that the equilibrium cycle can be extended to 16.07 extended full power days (EFPDs) while maintaining the PPF and CBC within design limits. To validate the effectiveness of TLBO, the optimized loading pattern of the equilibrium core was evaluated using the deterministic computer code DONJON5, for neutronic parameter analysis. The results show that the algorithm proposed in this study is a promising approach for reloading pattern optimization in CNPGS unit-3, offering potential improvements in reactor cycle length while ensuring safety and enhancing overall performance.

Benchmark Analysis of a Prismatic Type-high Temperature Gas-cooled Reactor with the ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 Nuclear Data Libraries

In this work we performed a benchmark analysis of the High Temperature Engineering Test Reactor (HTTR) fully-loaded start-up critical core with a 30-bundle loading configuration using the Monte Carlo code Serpent 2 with the recently released nuclear data libraries ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0. The purpose of the work is to reveal the impacts of using different ENDF nuclear data libraries on neutronics calculations of a prismatic high temperature gas-cooled reactor (HTGR). The benchmark results obtained with Serpent 2 were compared against the available experimental values and those attained by different computer codes (MCNP5 and SCALE) using the nuclear data library ENDF/B-VII.0. The comparative results showed good agreement between Serpent 2, the experimental values, MCNP5 and SCALE. The results also exhibited a notable difference between using ENDF/B-VII.0, ENDF/B-VII.1 and ENDF/B-VIII.0 in predicting the neutronics parameters of the HTTR that suggests further investigation in future work.

TRACE assessment of density wave instability onset with void reactivity feedback under natural circulation

Density wave oscillation (DWO) is an important safety concern for boiling water reactors (BWR) due to their high void fraction in the core. Power extensions to existing reactors such as the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) lead to increase susceptibility of DWO-type instability following an anticipated transient without scram (ATWS). Experiments performed at the Karlstein Thermal Hydraulic Test Facility (KATHY) have reproduced the reactivity feedback mechanism in BWRs under ATWS conditions. Using a neutronics module simulator, the KATHY facility was able to provide data on the effect of different neutronic parameters on DWO onset. This paper serves to assess the capability of the thermal-hydraulics code TRACE V5P7 for simulating DWO onset and development under natural circulation with neutronic feedback. A model of the KATHY natural circulation facility is created in TRACE and a reactivity feedback mechanism is implemented using a manual control scheme to simulate the parametric effects provided by the tests. This comparison allows for an assessment of the TRACE code as well as a better understanding of the instability mechanisms and behavior under the given conditions.

A characterization method for reprocessed spent nuclear fuel through neutron and gamma-ray multiplicity counting

It is difficult to measure neutrons and gamma rays coming directly from uranium and plutonium atoms in pyroprocessed ingot, which is of great interest to nuclear safeguards. Most neutrons of the ingot are emitted from 244Cm isotope and gamma rays from its fission products. Therefore, it is necessary to identify spent fuel characteristics such as initial enrichment, burnup and cooling time from measurement and estimate plutonium mass from calibration curve or burnup calculations.In this study, we explored the possibility of characterizing pyroprocessed ingots using neutron and gamma-ray multiplicity counting. We particularly focused on finding the possible correlation between neutron and gamma-ray multiplicity properties (singles, doubles, and triples) and spent fuel properties for U/TRU/RE preprocessed ingot. We examined neutron and gamma-ray correlation for spent fuel, using the Cascading Gamma-ray Multiplicity with Fission (CGMF) and the Fission Reaction Event Yield Algorithm (FREYA) library in MCNP6.2. Due to high spontaneous fission rate, neutron and gamma-ray multiplicity parameters did not strongly correlate with multiplication factor from the point model but was proportional to 244Cm mass. The results based on the CGMF and FREYA fission library showed the similar trends for neutron and gamma-ray cases. The sum of two multiplicity data yielded best linearity with the 244Cm effective mass. Although total neutron counting (S) alone was sufficient for effectively characterizing spent fuel due to its strong linearity, this feasibility study confirmed that, in principle, it will be more effective to calculate the sum of neutron and gamma-ray multiplicity properties to directly estimate the mass of 244Cm, rather than relying on the conventional method of calculating the multiplication factor.

Neutronic analysis of new generation TVS-2M fuel assemblies in VVER-1000 reactor type

The Russian fuel assembly TVS-2M is a newly proposed fuel configuration that contains gadolinium-oxide at different concentrations mixed with enriched UO2 at different U-235 enrichment values. The purpose of this study is to investigate the neutronic behavior of the Bushehr VVER-1000 reactor core by replacing the current standard Fuel Assemblies (FAs) called TVS with the new generation of TVS-2M FAs. In such manner, the reactor core was simulated under both modes of TVS-2M and TVS FAs arrangement using the WIMS, CITATION, MCNPX, and SuperMC codes in equal terms. Neutronic parameters of the reactor core such as effective multiplication factor, excess reactivity, neutron flux distribution, power peaking factor, and control rod worth are calculated and discussed within a comparative framework of both configurations of FAs; TVS and TVS-2M.

Assessing the benefit of thorium fuel in a once through molten salt reactor

This study comprehensively evaluates thorium-based fuel utilisation in the Molten Salt Reactor (MSR) with a once-through fuel cycle. It examines the whether the use of thorium is beneficial in extending the fuel cycle length of an MSR, fuel utilisation efficiency, reactor safety, and reduction of long-lived radioactive waste. To observe those parameters, this research compares the use of thorium and uranium as fertile fuel in a single-fluid, once through MSR using Uranium-233 (U-233), Uranium-235 (U-235), and Reactor Grade Plutonium (RGPu) as well as the fissile driver. MCNP radiation transport code with ENDF/B-VII.0 continuous neutron library was used to analyse the neutronic parameters and fuel burnup in a multi-refuelling scheme for a total burnup time of 8 years. The findings indicate that the utilisation of thorium in conjunction with U-233 and U-235 significantly enhances fuel consumption and improve reactor safety. Such benefits, however, did not appear in RGPu. Moreover, thorium effectively reduces the production of long-lived radioactive waste, a crucial issue in nuclear waste management. Considering technical and environmental aspects, this study offers novel insights into using thorium as a more sustainable nuclear alternative under a once through fuel cycle, within a condition of multi-refuelling cycle which is only possible in an MSR.

Fractional neutron point kinetics model for reactivity transients of the NuScale and comparison with the classical kinetics approach

In this work, the behavior of the neutron and natural circulation parameters for Reactivity Transients in the PWR reactor called NuScale are analyzed through the novel system equations. It model includes non-integer and ordinary order differentials coupling the neutron point kinetic with the thermos-hydraulic core of the reactor. Transient processes of increase and decrease of the feed flow and the coolant inlet temperature are analyzed to identify the feedback of the thermohydraulic effects with the neutron phenomena in a non-integer order scheme. In the results, it is observed that for short times the power and reactivity have the same behavior with the classical model and the non-integer order model for all values of the anomalous diffusion coefficient simulated. However, for long times of simulation, there is more reactivity and less power, in contrast to the classical model, when reducing the value of the anomalous diffusion coefficient these counter-intuitive effects are increased. Additionally, a delay is observed between the change in reactivity and power. In conclusion, it is observed that sub-diffusive processes are relevant when there are slow changes in reactivity, which can be analyzed through the Fractional Order Neutron Density Equation.

Analysis of the atomic ratio of H and Zr effect on the neutronics parameters of ZrH moderated space nuclear reactor

Zirconium hydride (ZrH) is an ideal moderator for space nuclear reactors due to its exceptional properties, including high hydrogen content, small neutron absorption cross section, and high operating temperature. The primary goal of this study is to examine how the atomic ratio of H to Zr (H/Zr ratio) influences the neutronics parameters of a ZrH moderated space nuclear reactor, with a focus on establishing a reliable reference for ensuring the optimal safety and minimization of such reactors. The neutronics calculation based on the ZrH moderated space nuclear reactor named Topaz-II is performed using the Reactor Monte Carlo code (RMC code) with the ENDF/VII cross-section database. The effects of the H/Zr ratio were studied with a particular focus on the initial keff, the burnup, the temperature reactivity coefficient and the criticality safety. The results show that with the increase of the H/Zr ratio, the initial keff increases while the drums’ worth decreases. The Moderator Temperature Coefficient (MTC) is a positive value that rises as the H/Zr ratio increases. In the dropping accidents, the reactor full of voids with seawater is more serious, and the introduced reactivity decreases with the increase of the H/Zr ratio.

Accident tolerant fuel design optimization for SMART reactor

Fully Ceramic Microencapsulated (FCM) fuel is one of the most advancedfuel types for the next nuclear reactor generations. The use of uranium carbide fuel in FCM offers multiple benefits such as high fuel density, improved thermal conductivity and enhanced mechanical stability. Both UC and conventional fuel (UO2) have the same melting point. As a result, under reactor operating conditions, UC has better safety margins. In addition, FeCrAl is a promising cladding material. However, because of its greater thermal neutron absorption cross section, FeCrAl has a neutronic penalty when compared to Zircaloy leading to an annual reload requirement with high fuel enrichment. Owing to their low power density and excellent design, small modular reactors (SMRs) have exceptional inherent safety. In this work, we determined the optimum design of an accident tolerant fuel (ATF), with respect to neutronics, in SMART reactor using dense UC FCM (in contrast to the conventional UO2 fuel) and a combination of zircalloy and FeCrAl (FCM -Zircaloy, FCM -FeCrAl, FCM -combining of zircaloy and FeCrAl) as cladding to achieve better thermal conductivity as well as safer performance, while considering the required fuel enrichment for the proposed designs. The effective multiplication factor and burnup were calculated for neutronic assessment of the SMART core, the outcomes are showcased alongside a comprehensive assessment of the characteristics and prospective. The neutronic parameters were estimated using the stochastic MCNPX 2.6 code. We found that Fully Ceramic Microencapsulated -Uranium Carbide (FCM-UC) fuel type was more attractive than conventional UO2. The optimum cladding design as compared to the reference case and the other cases of (FCM-UC), was determined. When 50% FeCrAl cladding was employed, case 15 in our study emerged as the superior design, demonstrating maximum production of 239Pu due to higher initial concentrations of 235U and quantities of 238U. This work may offer recommendations for future optimized designs of the SMART reactor parameters.

Neutron-physical characteristics of UO2 and UN/U3Si2 fuels with Zr, SiC and APMT accident tolerant claddings

H2 production due to zirconium hydration was the primary source of explosion in the Fukushima Daiichi nuclear accident. To improve accident tolerance of the new reactor designs, advanced fuel and cladding materials are being extensively investigated. Neutronic evaluation at both the pin-cell and assembly levels are performed for the conventional UO2 and the candidate UN/U3Si2 fuels with traditional cladding (Zr) as well as the accident tolerant claddings (SiC and advanced powder metallurgic ferritic, APMT). The neutronic performance, plutonium inventory, cycle length, radial and pin power distribution, spectrum distribution, spatial self-shielding analysis, plutonium radial distribution, and reactivity coefficients are evaluated and analyzed. Combination of UN/U3Si2 and APMT produces 68% and 66 % more 239Pu and 135Xe than the UO2 and APMT systems. In overall, the reactivity change versus burnup and other main neutronic parameters of SiC cladding are quite similar to the current Zr cladding. Using SiC resulted in an average increase of 0.85 % in reactivity in comparison to Zr conventional cladding (without any change in the enrichment of the fuel). Compared with the traditionally used UO2 fuel assembly, the UN/U3Si2 fuel assembly exhibit a higher (0.2 % at BOC and 0.4 % at EOC) pin power peak value in an assembly, which should be a concern from safety aspects. The negativity of FTC and MTC for the proposed UN/U3Si2 is higher than that of UO2 fuel. These higher and enhanced FTC and MTC values of UN fuels mean an improved inherent safety feature of UN-like high U density fuels.

AP-1000 fuel assembly performance with UO2 and UO2/U3Si2 fuels and silicon carbide cladding from the viewpoint of neutronics

ABSTRACT Accident Tolerate Fuels have been highly researched since the Fukushima disaster. To improve the safety of the AP-1000 fuel assembly, an advanced fuel, UO2/U3Si2, is considered. In addition, the silicon carbide (SiC) cladding that has the capability of reducing oxidation is compared with the baseline Zircaloy for AP-1000 fuel assembly along with the two different fuel cases (UO2 and UO2-U3Si2). The present study focuses on the neutronic performance of the UO2/U3Si2 fuel and thus the main neutronic parameters such as the fuel cycle length, pin power distribution, atom density evolution of main isotopes, and the reactivity feedback coefficients are evaluated. An increase of 6.72% in the criticality period is introduced when 70%UO2-30%U3Si2 fuel with SiC cladding is considered. The pin power peaking factors, fuel temperature coefficients, and moderator temperature coefficients of the proposed fuel result in very few changes. For instance, the maximum pin power distribution difference between 70%UO2-30%U3Si2 and UO2 fuel with SiC cladding is 0.001, which is negligible. Thus, its performance is as good as the traditional UO2 fuel from the neutronic aspects and controlling power distribution.

Study of gamma, neutron, and proton interaction parameters of some immunotherapy drugs using EpiXs, NGCal, and PSTAR software

Abstract In proton therapy, the protons are used to destroy the cancer cells efficiently at the Bragg peak without much damage to normal cells. The protons can also produce neutrons, protons, and high-energy gamma rays through nuclear reactions with cancerous and healthy tissues as well as with beamline components. The effective observed dose in the therapy is enhanced due to the interaction of nuclear particles with cancerous tissues. Such nuclear particles can have several effects on drugs used in immunotherapy, such as immunotherapy in combination with proton therapy, which has been used to treat cancer. In the present investigations, the gamma, neutron, and protons interaction parameters of some immunotherapy drugs, such as dostarlimab, atezolizumab, ipilimumab, nivolumab, and pembrolizumab, are determined by using EpiXs, NGCal, and PSTAR software. It is found that the EBF and EABF for all selected immunotherapy drugs increase with increasing penetration depth, peaking at 100 keV. The peaking is more symmetric at a higher penetration depth of 40 mfp than at a lower one of 1 mfp. At lower energies of gamma photons, the EBF values increase exponentially, and at higher energies, they increase linearly with increasing penetration depth for all selected drugs. Mass attenuation factors are slightly higher for thermal neutrons than for fast neutrons for selected immunotherapeutic drugs, indicating that thermal neutrons more actively participate in these drugs than fast neutrons. The mass attenuation factor for both fast and thermal neutrons increases with increasing weight percentages of hydrogen and is found to be higher for thermal neutrons. This is the first study in the literature to investigate the radiation interaction parameters for immunotherapy drugs, and it is helpful in radiation therapy and dosimetry.

Neutronic design analysis of CEA Cadarache URANIE subcritical assembly

This paper aims to study neutronic modeling using decommissioned fuel pins of the URANIE subcritical assembly at the Cadarache center of CEA for the future Tunisian Subcritical Assembly. Indeed, the CNSTN intends to exploit the fuel to implementing its subcritical assembly. Therefore, a comprehensive computational study of the URANIE fuel was carried out to analyze the neutronic parameters for future facilities including criticality safety, the neutron flux, and neutron spectrum, for future facilities. Indeed, MCNP Monte Carlo code was used to develop a three-dimensional sketch for URANIE core. Two versions were used to estimate an effective multiplication factor (keff): MCNP5 and MCNP6. The keff was varied from 0.85842 to 0.86463 with a maximal difference of 623 pcm and depended on ENDF/B data library releases. The nominal value of keff is acquired at the optimum lattice pitch of 4.6 cm, 30 cm in reflector thickness, and 185 cm in moderator level. Moreover, the reactivity of the facility is always negative. In addition, the energy spectrum was depicted at the three planes using 283 energy groups. Furthermore, the radial neutron flux was evaluated at eight positions. The axial neutron flux was estimated at various spatial planes of the assembly core. Inside the reactor core, the findings revealed indicate that the fast flux is higher and can be reach 7 × 10−4 n/cm2.Q and decreased steadily moving away from the center to the reflector part about 8 × 10−4 n/cm2.Q.

Development of a numerical model for disassembly phase of hypothetical core disruptive accident and application to a medium-sized mixed oxide fuelled fast reactor

Safety assessment of fast reactors involves analyzing Hypothetical Core Disruptive Accidents (HCDA) that could result in reactor core disassembly. HCDA is generally analyzed in three distinct phases that include pre-disassembly, disassembly and post-disassembly phases. In the present work, a new coupled neutronics-hydrodynamics computer code called FAst Reactor DISassembly (FARDIS-1) is developed in 2-D cylindrical coordinate geometry to model the super-prompt critical power excursion during the disassembly phase of HCDA in a sodium-cooled fast reactor (SFR). The modelling approach followed is essentially similar to the one used in VENUS-II code but with a few improvements. The code is benchmarked against the predictions of VENUS-II for a sample case of HCDA in the FFTF reactor model. This is followed by an analysis of the disassembly phase in a medium-sized (500 MWe) mixed oxide fuelled fast reactor. Two cases are assumed at the beginning of the disassembly phase: a sodium voided core (conservative) and a partially voided core (realistic) and the influence of several core neutronics parameters on the transients is elucidated through parametric studies. The peak power, thermal energy release, peak pressure, maximum temperature and work-potential are estimated and their trends are discussed.

Analysis of variation of neutronic and kinetic parameters of a Heavy Water Zero Power Reactor caused by changes in the fuel lattice pitch using MCNPX code

Analysis of variation of neutronic and kinetic parameters of a Heavy Water Zero Power Reactor caused by changes in the fuel lattice pitch using MCNPX code

Neutronic assessment and optimization of ACP-100 reactor core models to achieve unit multiplication and radial power peaking factor

This study focuses on the neutronic and burnup assessment of the PWR-based Chinese ACP-100 SMR core. Four different core models with varying fuel compositions and assembly configurations were analyzed using the Monte Carlo code SERPENT 2. Three of these models shared uniformly enriched fuel assemblies of 3 wt.%, 4 wt.%, and 4.45 wt.% 235U respectively, while the fourth model (the proposed design for the ACP-100 in this study) featured a mix of differently enriched fuel assemblies along with the inclusion of burnable absorbers. The burnup analysis of the cores spanned 1570 EFPDs, evaluating several neutronic parameters (beta effective, reactivity coefficients) as well as fuel cycle characteristics (cycle burnup, discharge burnup and cycle length). Changes in the isotopic concentration of fissile, fertile, and poison materials were observed and the assembly-wise linear power distribution, neutron flux spectra, and radial power peaking factors of the models were determined. In the mixed core, a combination of borated water, IFBAs and control rods were employed to achieve exact criticality (unit keff) and uniform power distribution (unit radial power peaking factor). Insertion of control rods into 41 of the 57 assemblies, in conjunction with IFBAs placed strategically, attained a near unit radial power peaking factor (PPF) of 1.0586. Finally, the combination of IFBAs, sol-bor of 936 ppm concentration, and CRs with rod worth of 414.33 pcm per assembly brought down the keff to 1.000, but increased the PPF to 1.117. While the achievement of uniform power distribution came at the cost of fuel burnup, the proposed mixed core successfully remained critical for over 1.5 years and reached a discharge burnup of 23.37 GWd/t for a two-batch refueling scheme.

Neutronic characteristic analysis of the GAMA microreactor

The GAMA Microreactor is a low-power nuclear power plant dedicated for remote area with a power output of 300 kWe. The primary characteristics of the GAMA Microreactor are its system compactness, design simplicity, and safe operation without moving parts. The fuel is uranium hydride powder, which works simultaneously as a moderator, contained within a stainless-steel vessel and surrounded by a graphite reflector. Liquid sodium-filled heat pipes are embedded in the fuel powder to extract fission-generated heat via passive coolant flow due to capillary action. As a part of the design process, neutronic parameters of the GAMA Microreactor have been determined using SCALE6.2 and OpenMC codes. Among the analyzed parameters are the excess reactivity as a function of burn-up, shutdown margin, control rod worth, temperature coefficient of reactivity (TCR), fuel density coefficient of reactivity (DCR), and reactivity coefficient due to hydrogen dissociation. From the calculation results, the GAMA Microreactor has a maximum excess reactivity of 0.049 at the operational temperature (700 °C) and 0.0788 at ambient temperature. The TCR value is negative (−2.267 pcm/°C) at the beginning of the cycle (BOL) and −4.9618 pcm/°C at the end of the cycle (EOC). Hydrogen dissociation imposes negative reactivity on the reactor both at the BOL (−0.2631 Δk/k-%dissociation) and EOC (−0.3524 Δk/k-%dissociation). Therefore, the GAMA Microreactor maintains its inherent safety characteristics over a considerably long operation time.

Neutronic Analysis of the AP1000 Fuel Assembly with Accident Tolerant Cladding Materials

An alternative material for fuel cladding was required to prevent oxidation caused by interacting with steam, leading to improvement in core integrity. This study analyzes reactor physics parameters of various cladding material candidates for Pressurized Water Reactor (PWR) such as SS-304 austenitic stainless steel, FeCrAl alloy, APMT alloy, and silicon carbide (SiC) ceramic, as candidate Accident Tolerant Fuel (ATF). The neutronic parameters such as infinite multiplication factor (k-inf), and neutron spectrum, while temperature reactivity coefficient related to fuel temperature (DTC) and moderator temperature (MTC) is also considered, followed by a void coefficient of reactivity (VCR) of each candidate material were then compared with ZIRLO as a standard cladding material of AP1000. k-inf calculated by SRAC2006 is also compared to MCNP for various fuel assembly types. At the beginning of cycle (BOC), the 2.35% UO2 using SiC gives a higher kinf than ZIRLO at 937 pcm, while 4.45% UO2 with 88 IFBA & 9 PYREX at 796 pcm. FeCrAl, APMT, and SS-304 cladding gave a smaller k-inf compared to ZIRLO in the range of 11000-14000 pcm at 2.35% UO2 fuel assembly. The values of DTC, MTC, and VCR were still negative throughout the reactor operation which indicates that the inherent safety feature of alternative cladding was possible for this type of fuel assembly, especially for iron-based cladding material followed by an increase in fuel enrichment.