Annular Fuel Research Articles

Study on non-uniform circumferential temperature distribution of annular fuel in the tight lattice configuration of a lead-bismuth cooled fast reactor

To guarantee the safe operation of nuclear reactors and protect fuel elements from being damaged under long-term thermal stress, analyzing the circumferential temperature distribution of the outer cladding of nuclear fuel rods in reactors with tight lattice structures is crucial. In this work, a mathematical and physical model was established through theoretical analysis to calculate the circumferential temperature distribution of the outer cladding of an annular fuel rod. The proposed model provides an analytical expression for the circumferential temperature distribution, and the key factors influencing this distribution are determined. The results of the mathematical-physical model and numerical simulation are then combined through numerical fitting. Consequently, a numerical fitting equation is obtained under the steady-state condition of the lead-bismuth reactor using annular fuel and compared with the numerical simulation data acquired via computational fluid dynamics. The results demonstrate that the proposed numerical fitting equation has high accuracy and can serve as a valuable computational tool for analyzing the circumferential temperature of fuel claddings during the operation of lead-bismuth reactors with annular fuels.

DNBR Limit Analysis for the Inner channel of Annular Fuel

The departure from nucleate boiling ratios (DNBR) limit for annular fuel might vary from that of conventional nuclear fuel rod assemblies due to the distinct features of annular fuel assemblies. Precise estimation of DNBR is crucial for ensuring the safe and optimal functioning of nuclear reactors equipped with annular fuel assemblies. This study examines the DNBR limit of inner channel for annular fuel, using the look up table-2006 (LUT-2006).A comprehensive database having data for both normal operation and accident situations in pressurized water reactors (PWRs) using annular fuel was built to analyze the DNBR limit of the inner channel of annular fuel. The critical heat flux (CHF) data in this database were categorized into two groups with the limiting quality as the boundary. The prediction accuracy of LUT-2006 and the DNBR limits were evaluated respectively. Given that the inner diameter of the inner channel of annular fuel in mainstream designs typically ranges from 8-9 mm, and considering the potential occurrence of the limiting quality phenomenon in the inner channel, it would be more suitable to utilize the LUT-2006 based on an 8 mm tube standard to establish the DNBR limit for the inner channel of annular fuel. A conservative DNBR limit of 1.308 is determined for the inner channel of an annular fuel using the LUT-2006. In future thermal-hydraulic optimization designs of annular fuel, it is advisable to utilize LUT-2006 for calculating the DNBR of the inner channel of the fuel.

Evaluating Hydrogen-based Moderators in High-Temperature Gas-cooled Reactors with 5 wt% Enriched Uranium Annular Fuel Rods

Abstract Small Modular Reactors (SMRs) based on high temperature gas-cooled reactor (HTGR) technology are being developed for providing high-temperature process heat and high-efficiency (>40%) electrical power generation. However, most of the HTGR-SMR concepts require high assay low enriched uranium (HALEU) fuel, with enrichments typically above 10 wt% 235U/U, to get sufficiently high burnup levels and fuel lifetime. The need for HALEU is primarily a consequence of the low volumetric density of fissionable material in tri-structural isotropic (TRISO) fuel particles, and also the use of graphite as a moderator/reflector. A previous study has shown that a modified prismatic HTGR fuel assembly with hydrogen-based moderator (7LiH) and cylindrical fuel elements of 5 wt% 235U/U enriched uranium can greatly reduce fuel consumption of a HTGR. However, such a design concept could lead to positive temperature reactivity coefficients (TRCs), making reactor control more challenging, with reduced passive safety. The purpose of this study is to evaluate variations of the hydrogen-based moderator in this alternative fuel assembly concept to identify configurations that achieve negative TRCs, thus improving passive safety characteristics. Calculation results demonstrate that negative TRCs can be achieved with reduced hydrogen mass such that natural uranium consumption is substantially less than that of the TRISO fuel concept, with comparable or higher core life. This study also explores the option of using 7LiOH and NaOH as hydrogen-based moderators, instead of 7LiH, thus allowing operation at higher temperatures, where hydrogen TRCs are lower.

Advanced characterization-informed machine learning framework and quantitative insight to irradiated annular U-10Zr metallic fuels

U-10Zr Metal fuel is a promising nuclear fuel candidate for next-generation sodium-cooled fast spectrum reactors. Since the Experimental Breeder Reactor-II in the late 1960s, researchers accumulated a considerable amount of experience and knowledge on fuel performance at the engineering scale. However, a mechanistic understanding of fuel microstructure evolution and property degradation during in-reactor irradiation is still missing due to a lack of appropriate tools for rapid fuel microstructure assessment and property prediction based on post irradiation examination. This paper proposed a machine learning enabled workflow, coupled with domain knowledge and large dataset collected from advanced post-irradiation examination microscopies, to provide rapid and quantified assessments of the microstructure in two reactor irradiated prototypical annular metal fuels. Specifically, this paper revealed the distribution of Zr-bearing secondary phases and constitutional redistribution across different radial locations. Additionally, the ratios of seven different microstructures at various locations along the temperature gradient were quantified. Moreover, the distributions of fission gas pores on two types of U-10Zr annular fuels were quantitatively compared.

A novel method for calculating annular fuel cell resonance effective temperature

Annular fuels can increase the power density of reactors because of their double-sided cooling design, which is of great significance to the miniaturization and long-life operation of pressurized water reactors. In this study, a thermal–hydraulic analysis code named Thermal-Hydraulic Code of Annular Fuel with Single channel (THCAFS) was developed for annular fuels, and a calculation model for the resonance effective temperature of annular fuel cells was established. An annular fuel cell of a four-loop pressurized water reactor (PWR) designed by Westinghouse company was used as the calculation object, and the numerical results of THCAFS were compared with those of existing sub-channel codes such as VIPRE-01, TAFAX, and NACAF. Meanwhile, the Monte Carlo code RMC was used to simulate the radial power distribution and burnup process of the annular fuel, and its thermodynamic behavior was simulated using the self-developed code THCAFS. The radial power distribution, nuclide density variation, and cell temperature field under different burnup levels were also obtained, and the reaction rate equivalence was used instead of the effective resonance integral to investigate the calculation method used for resonance effective temperature of the annular fuel cell. The results show that THCAFS can be preliminarily applied to annular fuel designs and thermal–hydraulic analysis, while the effective temperature calculation method can provide important reference values for studying the resonance effective temperature of annular fuels.

Investigation on axial relocation of fragmented annular fuel under loss of coolant accident conditions with discrete element method

Axial fuel relocation of conventional cylindrical fuel rods under Loss of Coolant Accident (LOCA) conditions has been intensively studied. However, characteristics of cladding ballooning, fuel fragmentation and relocation of a dual-cooled annular fuel rod could be quite different due to its innovative geometry. In this study, based on fragmentation images of annular fuels after in-pile irradiation, discrete geometry for fuel fragments was established considering the effect of burnup. With the ballooning cladding as constraint boundary and the gravity as driving force, axial relocation of an annular fuel at different burnup stages was simulated with discrete element method (DEM). The employed DEM methodology was first validated through a benchmark against a theoretical model and a reference simulation. Simulation results indicate that relocation is determined by the fuel fragmentation morphology and cladding ballooning shape. At low and medium burnup stages, it was found that although there exists obvious mass accumulation in the ballooning region, many fuel fragments above the ballooning region are stuck between the internal and external claddings, resulting in almost no missing length of the fuel stack. At high burnup stages, because of the presence of finer fuel particles at High Burnup Structure (HBS) regions, fuel jam can be mitigated or avoided, resulting in a higher peak mass fraction and an obvious missing fuel length. The critical HBS volume fraction that can totally avoid the fuel jam was determined as about 15%. In addition, the sensitivity study indicates that the higher the volume proportion of HBS is, the more obvious the axial fuel relocation is. In the absence of sufficient irradiation data, this study provides a reliable modeling methodology and valuable understandings of the axial fuel relocation in dual-cooled annular fuel rods.

Physics Evaluation of Alternative Uranium-Based Oxy-Carbide Annular Fuel Concepts for Potential Use in Compact High-Temperature Gas-Cooled Reactors

Abstract Lattice physics calculations have been carried out to evaluate the performance and safety characteristics of a modified high temperature gas-cooled reactor (HTGR) prismatic fuel block concept, based on the MHTGR-350 benchmark problem. Key changes were to replace the conventional tri-structural isotropic (TRISO)-filled fuel compacts with heterogeneous, multilayer annular fuel pellets made with UCO, ThCO, or (U,Th)CO. These fuel pellets have multiple protective cladding layers of pyrolytic carbon and silicon carbide, which will give it robust qualities. With the increased loading of U-235 in the fuel block, it was necessary to replace up to 78 fuel holes and 42 coolant holes with a hydrogen-based moderator (7LiH), in order to ensure a thermal neutron energy spectrum in the lattice. Calculation results demonstrate that the modified fuel concept has several advantages and some challenges relative to the conventional MHTGR-350 design concept. With the increased uranium loading and the reduced neutron leakage due the use of 7LiH moderator rods, higher burnup levels and lower natural uranium consumption levels can be achieved with the same level of uranium enrichment. In addition, the expected fuel residence time increased by a factor of 20 or more, making such a concept very attractive for use in small, modular, “nuclear battery” HTGRs that would only need to be fueled once. Calculation results for the current concept indicate positive graphite and hydrogen moderator temperature coefficients, and further modifications will be required to ensure a negative power coefficient of reactivity.

Investigation on radial fuel relocation and its influence on heat split phenomenon of dual-cooled annular fuel element

Radial fuel relocation due to fuel fragmentation can significantly change the heat split, further affecting the thermal performance and safety of dual-cooled annular fuels. Considering the rarity of in-pile irradiation data for annular fuels, three different fuel relocation and its recovery models were developed based on modifications of relocation models for solid fuel rods or the mechanistic relocation simulation of pre-fragmented, interacting and deformable fuel fragments. The three relocation models were implanted into the fuel performance analysis code FROBA-ANNULAR, and thermo-mechanical simulation was carried out to investigate the influence of fuel relocation on gap size variation and heat split variation in an annular fuel. The results indicate that different relocation models lead to significant difference on the prediction of gap size and heat split. Relocation models developed through mechanistic simulation of pre-fragmented fragments seems to be more reasonable, compared with the other two models developed based on unreasonable and extreme assumptions. Mechanistic simulation results indicate that radial relocation of fragmented annular fuels can be bidirectional even at the initial burnup stages, resulting in the reduction of gap sizes of both internal and external fuel-cladding gaps. In addition, it was found that the gap size variation due to radial fuel relocation significantly changed the heat split and matching-rate between heat split and flow flux distribution, exerting an influence on fuel temperature and MDNBR.

Neutronic and Thermalhydraulic Evaluation of Annular Fuel in PWR with Three Different Nanofluids

In pressurized nuclear reactors, one of the fuel design principles is the fuel centerline temperature limit. Since the thermal conductivity of UO2 is not high, the temperature increases rapidly from the fuel surface to the fuel center. To overcome this limitation, the use of annular type fuels instead of solid fuels is one of the improvement efforts. In this study, the effects of using annular fuel in a typical PWR are investigated. At the same time, the effects of adding nanoparticles (〖Al〗_2 O_3,Cu and TiO_2) to the coolant water in the case of solid and annular fuels are also investigated. For both cases, neutronic and thermalhydraulic calculations are made and compared. In neutronic point of view, it has been shown that the use of annular fuel does not cause a significant change in the k_eff value but the presence nanoparticles reduces the k_eff. This reduction is maximum for Cu+water and minimum for 〖Al〗_2 O_3+water nanofluids. In thermalhydraulic point of view, it has been shown that the annular fuel positively affects the fuel temperature limits and the addition of nanoparticles could provide an additional contribution to this, provided that it is in very small amounts (φ

Development of subchannel thermal-hydraulic analysis code for dual cooled annular fuel

Regarding the geometric structural characteristics of innovative dual cooled annular fuel and the possibility of heat split and flow distribution among the internal and external channels, the development of new computational tools is essential for estimating safety margins and accurate assessment of its thermal-hydraulic performance. The SADAF code (Subchannel Analysis Dual cooled Annular Fuel) by COBRA-EN code is developed for this purpose. In the SADAF code, using COBRA-EN code for subchannel analysis in internal and external subchannels, a program has been developed to compute new variables that need to be considered in the thermal-hydraulic assessment. Also, fuel heat transfer calculations were performed by the finite difference method. Benchmark calculations have been performed for Westinghouse 13 × 13 annular fuel rod design, the results were compared with the reference values, and a good agreement was found. Evaluation of the results shows that the total pressure drop predicted by the SADAF code differs from the benchmark results by less than 1%. This difference is about 0.4%, 1.0 K, and 1.0% for fluid temperatures, rod surface temperatures, and flow rates, respectively. It is also shown that these error levels are within acceptable margins. The results analysis exhibited the robustness of the SADAF code in sub-channel analysis of annular fuel assembly.

Design of the CAREM nuclear reactor core with dual cooled annular fuel and optimizing the thermal-hydraulic, natural circulation, and neutronics parameters

The small modular nuclear reactor (SMR) with annular fuel which is cooled internally and externally has the potential to maintain high power density with a sufficient safety margin. The benefit of the dual cooled annular fuel design is that heat can be transferred to the coolant from both the outer and inner surfaces.In this paper, design of the Argentina small modular reactor (CAREM) core with hexagonal assemblies using dual-cooled annular fuels with internal and external cooling is presented. Neutronics and thermal-hydraulic parameters are evaluated and analyzed by changing internal fuel radius. For this purpose, at the first, the dual-cooled annular fuel under clean and cold conditions is modeled and the effective multiplication factor has been calculated for different inner clad diameter. Indeed, the internal and external radius have been changed in such a way to maintain the reactor under moderated.Then, these annular fuels under full power conditions are modeled and power peaking factor has been calculated. Finally, natural circulation parameters are performed for a simulated fuel rod in the hot channel using computational fluid dynamics (CFD) simulation codes. These results are compared with the conventional CAREM reactor. One of the most prominent advantages of the annular fuels is the ability to make softer neutrons in which reduce maximum power peaking factor and improves fuel management and safety parameters in the reactor core.For data fitting, an artificial neural network is trained using the observed data. The input consists of different internal and external radiuses and output consists of fuel rods’ pitch, thermal-hydraulic and neutronic parameters. Finally, the optimal geometry of fuels is determined using the neural network by implementing the genetic algorithms.Indeed, developed Artificial Neural Network (ANN) utilizing the obtained data, predicts the thermal-hydraulic and neutronic parameters of the CAREM reactor core with dual-cooled annular fuels. Presented optimization algorithm, which has a significant ability to attain the best solutions, also determines the optimal values of natural circulation parameters (Vmax/Vavg, Vout-Vin, and pressure drops), heat transfer coefficients, MDNBR, RPPF, and excess reactivity of the reactor. Also, the designed artificial neural network and genetic algorithm has been validated using neutronic and thermal hydraulic calculations. Results indicate that by using annular fuels, better reactor safety and thermal efficiency are achieved.

Researches on Special Thermal Hydraulic Phenomena of Annular Fuel Assembly by Sub-Channel Analysis Code-NACAF

Dual cooled annular fuel is a novel fuel design, which has the potential to improve the reactor power density while maintaining or improving its safety margin. The effects of tight-lattice geometry, fuel burnup, fuel expansion, coolant channel blockage on the thermal hydraulic performance of annular fuel is studied to illustrate its special features in this paper. A sub-channel analysis code named NACAF, which includes empirical constitutive models in consideration of tight-lattice effects on prediction of pressure drop, critical heat flux and turbulent mixing, channel blockage model, heat conduction model for dual surface cooling condition, coolant flowrate distribution between inner and outer channel, is developed for annular fuel assembly or core analysis based on homogenous fluid model. Validation work is carried out with comparing NACAF results with analytical solutions, as well as numerical results of existing sub-channel code for annular fuel, such as VIPRE-01 and TAFIX. Comparison results demonstrates NACAF’s prediction error is acceptable and it has the ability to simulate thermal hydraulic performance of annular fuels or annular fuel bundles. Based on the developed and verified NACAF, the special thermal hydraulic phenomena of annular fuel are studied to clarify the features of annular fuel.

RESONANCE CALCULATION BASED ON THE GLOBAL-LOCAL METHOD FOR THE COMPLEX GEOMETRY PROBLEMS

With the development of nuclear reactor, new complicated designs introduce some challenges for the resonance method in the NECP-X code, which is called as the Global-Local method. There are two limitations of the present NECP-X. One is the geometry modeling limitation, and the other one is that the Global-Local resonance method can only deal with the cylindrical fuel rods in the current version. Therefore, some fuels like plate, annular fuels cannot be calculated in the NECP-X code because of these limitations. To overcome above issues, the constructive solid geometry (CSG) method is developed to model the complex problem, and the capability of constructing and solving the equivalent 1D plate and 1D annual fuel pin based on the framework of Global-Local method, is developed. A set of tests are calculated, including multi-annular fuel problems and plate fuel problems. The results show good accuracy of the new developed resonance self-shielding method.

压水堆核电站实心燃料堆芯向环形燃料堆芯过渡方案研究

压水堆核电站实心燃料堆芯向环形燃料堆芯过渡方案研究

Optimal design of a VVER-1000 nuclear reactor core with dual cooled annular fuel based on the reactivity temperature coefficients using Thermal hydraulic and neutronic analysis by implementing the genetic algorithms

One of the approaches which can help the enhancement of a reactor power is changing its fuel geometry. For this purpose, as well as decreasing the maximum fuel temperature in PWR reactors, the technology of annular fuels with ability of internal and external cooling shows its importance and has been considered widely. Such fuels are investigated in western PWR and VVER-1000 reactors. Hence, in this study, this fuel in VVER-1000 reactors are considered and studied thoroughly. In this paper, fuel and coolant reactivity temperature coefficients are calculated for a VVER-1000 Nuclear Reactor with Dual Cooled Annular Fuel with changes of internal radius and different power levels and effects of the fuel internal radius on the reactivity temperature coefficients are investigated. By analyzing the fuel internal radius changes in a specific range in the neutronic code, the effects of effective multiplication factor are investigated.For purpose of data fitting, an artificial neural network is trained using the observed data. The input consists of different internal and external radiuses, outputs consist of pitch, fuel and coolant reactivity temperature coefficients. Finally, the optimal geometry of fuel is determined using the neural network by implementing the genetic algorithms based on these dynamic-coefficients. In the optimization process, it has been shown that having an internal radius 2.67 mm and external radius 6.95 mm provides an optimal geometry. Also, validation of the designed artificial neural network and genetic algorithm has been done using neutronic and Thermal hydraulic calculations.

Comparison between standard solid fuel and a new annular fuel performance in the core of a PWR

Abstract In new generation PWR the annular fuel has been proposed as one of the options to achieve higher power density, larger safety margin and reduced electricity generation cost. In the current work, RELAP5 code is used to compare the thermal hydraulic parameters for both solid fuel and internally and externally cooled annular fuel in a core of a PWR. MCNP6 code is used to evaluate the neutronic design and basic safety parameters of the annular fuel. To accomplish this goal, RELAP5 input files for both solid and annular fuels are developed. In these files, a 13 · 13 array annular fuel design is used while the 17 · 17 standard array design is used for the solid fuel. A 100% core power, steady state normal operation is assumed in the current investigation. Also, MCNP6 code input files for both fuels are prepared. It is found that annular fuel has lower peak fuel temperature than the standard solid fuel, which is an important advantage of the annular fuel rather than the solid one. Also, comparisons were performed for reactivity feedback coefficients of the two fuel types at BOC. Burnup calculations were performed to study the energy conversion capability of the annular fuel as well as rim effects.

Optimal design of a Small Modular Reactor core with dual cooled annular fuel based on the neutronics and natural circulation parameters

Small Modular Reactors (SMRs) are the creative design of nuclear reactors that have received considerable attention in recent years. Since there is insufficient operational experience in SMRs, extensive research on these types of reactors may be needed to improve the current level of these systems performance. Also, one of the approaches which can help the enhancement of a reactor power is changing its fuel geometry. For this purpose, as well as decreasing the maximum fuel temperature in the reactors, the technology of annular fuels with the ability of internal and external cooling shows its importance. The nuclear SMRs with annular fuel which is cooled internally and externally has the potential to increase high power density while maintaining a sufficient safety margin. The benefit of the dual cooled annular fuel design is that heat can be transferred to the coolant at both the outer and inner channels. Hence, in this study, for the first time, dual cooled annular fuel in the Small Modular Reactors is considered and studied thoroughly. In this paper, neutronics and natural circulation parameters are calculated for an SMR with Annular Fuel which is cooled internally and externally, with changes of the internal radius. By analyzing the fuel internal radius changes in a specific range via the neutronics and computational fluid dynamics (CFD) simulation codes, the effects of power peaking factor, effective multiplication coefficient and natural circulation parameters are investigated. For the purpose of data fitting, an artificial neural network is trained using the observed data. The input consists of different internal radiuses which yields to output consisting effective multiplication factor, power peaking factor and natural circulation parameters. The Optimal geometry of annular fuel is determined using the neural network by implementing the genetic algorithms based on these neutronics and natural circulation parameters. Finally, using a simulated optimal geometry in the hot channel by neutronics and CFD simulations, thermal–hydraulic and neutronics calculations are accomplished, and these parameters are compared with the conventional SMR which uses solid (conventional) fuel. One of the main results of the evaluation is that optimal internally and externally cooled fuel in the presented SMR has a satisfactory margin for the departure from nucleate boiling (DNB) and maximum fuel rod temperature relative to the solid fuel which is used in the conventional SMRs. Indeed, fuel center temperature of conventional SMR is much more than that of presented optimal SMR. Also, presented optimal geometry increases the core effective multiplication coefficient which leads to increase in the excess reactivity. Also, reduces total power peaking factor and improves natural circulation parameters compared to the conventional SMR.

Design of a Small Modular Nuclear Reactor with dual cooled annular fuel and investigation of the fuel inner radius effect on the power peaking factor and natural circulation parameters

Recently, a kind of fuel called “dual-cooled annular fuel” is used in the nuclear reactors, which allows the coolant to flow from the inside and outside of the fuel rod and finally causes a substantial increase in power density. The dual-cooled annular fuel rod has both inner and outer coolant channels. Also, Small Modular Reactors (SMRs) are the innovative design of nuclear reactors making remarkable interest during recent years. Since there is not enough available operating experience on SMRs, it might be possible to initiate extensive investigations on these types of reactors for the purpose of improving the current performance level of these systems, significantly.In this paper, for the first time, the core of a Small Modular Reactor (SMR) is designed based on the use of internally and externally cooled annular fuels. And then, neutronics and natural circulation parameters of this type of reactor are analyzed. For this purpose, at the first, the dual-cooled annular fuel under clean and cold conditions is modeled for the different inner clad diameter with same distance gap to gap of fuel rod and effective multiplication factor has been calculated. Then, these annular fuels under full power conditions are modeled for each inner clad diameter and power peaking factor has been calculated. Finally, natural circulation parameters calculations are performed for a simulated fuel rod in the hot channel using computational fluid dynamics simulation codes. These calculations are compared with a conventional NuScale reactor that does not use this kind of fuel. One of the most prominent of its advantages is the ability to moderate more of the neutrons in the reactor core with this type of fuel, which can increase the core effective multiplication, reduce total power peaking factor and improve natural circulation parameters. Increasing of the effective multiplication coefficient at cold and clean condition leads to increase the excess reactivity of the core which increases the operation cycle without refueling and it is useful from fuel management view-point. Also, it was found that by increasing the inner radius by more than 0.5 (cm), the power peaking factor increases, which is due to the more increasing of the pitch lattice relative to the mean free-path and excessive reduction of the mean power-density in the core and, also, increasing the flux gradient.

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.

Prediction of temperature distribution in annular fuels using orthogonal collocation method

In this paper, radial temperature distribution in an annular fuel rod is calculated using the orthogonal collocation method (OCM) and the finite volume method (FVM). Using the non-dimensional transient heat transfer equation and defining the relative radius of the nodal points inside each solid region, the annular fuel rod is transformed into the interval [0,1]. By doing this, one can use the Legendre polynomial roots in the interval [0,1] as the collocation points. In order to solve the system of equations, the boundary conditions appropriate to different discretization schemes were applied. To validate the proposed scheme, results were compared with those reported in the literatures, finite different method (FDM) and analytical solution. It was found that although FVM and FDM are simpler to implement, the OCM with fewer nodes is much faster than the FVM for comparable accuracy.