Research Reactor Core Research Articles

THERMAL TECHNOLOGICAL CONDITION OF IVG.1M RESEARCH REACTOR CORE UNDER VARIOUS OPERATING MODES

The relevance of the study is related to the determination of the thermal characteristics of the IVG.1M research reactor core with the low enriched uranium fuel under the nominal and design operating modes. The thermal technological condition of the IVG.1M research reactor during the start-up are determined by the readings of the temperature, pressure and coolant flow sensors of the information and measuring system. The indirect methods including the computer simulating ones are used to determine the temperature of the core structural materials and the distribution of the coolant temperature by the height of the fuel assembly. The research has been carried out using the method of the finite element analysis using the ANSYS Fluent software package. The study goal was to verify the adequacy of the calculation methodology and obtain the calculated data on the temperature distribution in the fuel assembly in the reactor power range from the nominal to design one. The article presents a description of the IVG.1M reactor, the research methodology, computer model, simulation results and the comparison of the calculated data with the experimental ones. The study scientific novelty consists in determining the temperature conditions of the fuel rods during the reactor operation at various levels of the design capacity with a conservative approach to the cooling conditions. The significance of the research results lies in the fact that a computer model can be used to determine the characteristics of the IVG.1M reactor core under the reactor various operating modes and to analyze the thermohydraulic processes in the fuel assembly.

Research of VDT scheme for porous media thermal-hydraulics analysis of plate-type fuel assembly

Thermal-hydraulics analysis of nuclear reactor core is crucial for the safe operation of nuclear power plants. Porous media models in computational fluid dynamics (CFD) code are commonly used to simulate flow and heat transfer in nuclear reactor cores because they simplify geometry and reduce computational cost. However, their low spatial resolution limits detailed analysis of internal flow and temperature fields. This research develops a Virtual Domain Technology (VDT) scheme to improve the resolution of porous media models without increasing mesh density. VDT sets localized resistance coefficients to create virtual solid and fluid domains within the porous core representation. The flow and thermal effects of these virtual domains mimic real structural effects. VDT was verified on a cylinder flow case and showed excellent agreement with CFD for velocity and pressure fields. Application to a plate-type research reactor core demonstrated VDT’s ability to capture detailed temperature profiles and flow traces consistent with CFD, using only 6% as many mesh cells. By improving spatial resolution, VDT enhances porous media-based modeling of reactor cores to provide more accurate thermal-hydraulics analysis, while maintaining computational efficiency. This will aid optimization and safety analysis for advanced reactor designs.

Ensuring radiation safety during dismantling, transportation and long-term storage of the SM-3 research reactor core

Described shortly here is a procedure of demounting, removal, transport and long-term storage of the SM-3 core, based on the previous experience of reactor refurbishment undertaken in 1991. Prior to performing refurbishment, computations and calculated data analysis were performed to prove radiation safety of this work, which included estimation of the activity level for activation products in the structural materials of the nuclear research reactor core and the radiation conditions at different stages of its handling. As evidenced by the calculated data, the activity of the main dose-forming radionuclide 60Co attains equilibrium in about 12 years of radiation exposure. Taking into account the fact that the time period between two refurbishments was longer than 12 years, the calculated values of the equivalent dose rate were normalized to the radiation monitoring data obtained during the previous refurbishment, taking into account the calculated activity of 60Co radionuclide. The normalization made it possible to confirm reliability of estimates. The obtained activity data of activation products and taking into account the time spent during the SM-3 refurbishment in 1991, the radiation impact on personnel was estimated. Calculated values of the anticipated effective radiation exposure doses to the personnel engaged in the refurbishment revealed that the main limits of the personnel radiation exposure established in accordance with NRB-99/2009 were not exceeded. Comparison of the results of calculating the equivalent dose rate with the results of radiation monitoring at various points allowed us to establish that during the calculation and analytical justification of the radiation safety of work, the assessment of reflected radiation was significantly underestimated. But the radiation monitoring data, personal radiation monitoring, as well as recorded data of automatic radiation monitoring system show that all work was performed in compliance with the requirements of regulatory documents in the field of radiation safety.

Validation of the dynamic simulation capabilities of Serpent2/Subchanflow using experimental data from the research reactor SPERT IV D-12/25

Research nuclear reactors are critical to the development of nuclear technology, but because of the complex configuration of the fuel assemblies and the different initial operating conditions, neutronic and thermal–hydraulic analyzes are performed with dedicated or adapted codes. These simulation tools might not be as accurate, since they use simple resolution methods and assumptions that do not always capture all aspects of the behavior of a nuclear research reactor. On the other hand, as time goes by, the evolution of numerical tools for the core analysis of power reactors have experienced a considerable progress. Nowadays, pin/subchannel level analysis of the core with coupled codes based on transport (SP3, MOC, SN, etc.) or Monte Carlo methods are applied in addition to the nodal diffusion codes. Hence, the research community is adapting and validating selected high-fidelity tools developed for power reactors to perform detail core analysis of e.g. Material Testing Reactors (MTR) cores at plate and subchannel level.This work deals with the validation of the high-fidelity coupled Serpent2/Subchanflow, which was modified and extended for the plate/subchannel analysis of MTR-cores, using the data of rod ejection tests performed in the SPERT IV D-12/25 reactor, especially the tests B-34 and B-35 were selected to validate the dynamic capability of Serpent2/Subchanflow. In these unique tests, experimental data e.g. thermal neutron flux, core power evolution during the rod ejection tests, and the plate cladding temperature was measured. It is noted that, due to the lack of detailed information on the initial conditions, the extraction and introduction scenarios of the transient rod for the reactivity insertion required calibrations and assumptions regarding velocity and position.The comparison of selected parameters predicted by the coupled simulations at plate/subchannel level of the SPERT IV reactor with the measured data at static and transient conditions shows excellent agreement confirming the high accuracy appropriateness of the used code for the analysis of research reactors. The calculated values of thermal neutron flux and core power evolution have a statistical error of ± 2 sigma. It was also found that the maximum temperature difference between calculated and experimental values is 7 °C and ∼ 10 °C for tests B-34 and B-35, respectively. In addition, the coupled code predicts for the first time the temperature of each plate and subchannel considering the local feedbacks between neutronics and thermal-hydraulics allowing the identification of the hottest/coldest plate in the core. The high-fidelity validation tool can provide comparison solutions for current research reactor core analysis methods, such as core analysis with point kinetics or 3D nodal diffusion codes coupled with fuel assembly-level thermal-hydraulics.

Nuclear safety operating margin using new power peaking factors for NUR nuclear research reactor

Among a series of important safety parameters necessary for a complete analysis of a nuclear reactor core, the nuclear safety operating margin of a nuclear reactor is of great importance. The methodology generally used is to combine thermalhydraulic calculations with neutronic calculations. In this work, new Power Peaking Factors (PPFs) are calculated for the NUR nuclear research reactor core X-1 configuration by two ways; a deterministic way using WIMSD-CITVAP neutronic calculation in a fine calculation model, and a second way by a probabilistic method using MCNP5 model. The obtained results give the PPFs for each nuclear reactor core channel and the hot channel is then determined. Once this channel known, an asymptotic thermalhydraulic model is applied. This model gives the coolant and the cladding temperature evolutions in the hot channel for the calculated PPFs in the two presented models. All these results are compared with those obtained using the old PPF commonly used in such calculations. Results show that PPFs obtained by deterministic method are the most conservative. Thus, this last result is retained for a nuclear safety evaluation. An equation set-up is made for calculating the critical heat flux ratio. This ratio is verified to be greater than the limit of nuclear safety operating margin for NUR nuclear reactor.

Assessment of the safety impact caused by the variations of the critical position of control rods and by the presence of an in-core flux trap on the power peaking factors in an MTR-type research reactor

The main objective of this work is to evaluate the safety impact caused by the variations of the critical positions of control rods and by the core configurations with flux trap on both the locations of the hot channels and on power peaking factors (PPFs) values in a research reactor core. Thanks to an appropriate computational methodology and to parallel computing capabilities of MCNP5, we conducted a demonstration study on two core configurations and twenty selected critical states of the NUR light water research reactor.The results obtained show that the locations of the hot channel in a given core vary continuously, in particular, around the central positions of the reactor cores as a function of the critical positions of the control rods. Moreover, it was found that the axial and radial power peaking factors depend mainly on the critical positions of the control rods whereas the local power peaking factors strongly depend on the presence of a flux trap in the core configuration. The maximum total power peaking factor can exceed, in some cases, the safety limit value (3.0) as found for the core configuration where a central flux trap was present. Indeed, it has been observed that the hot channels are mainly located at the central positions of the reactor cores of interest, that is where the fuel elements are the most consumed.Overall, due to the prohibitive values exceeding the safety limits found for the PPFs, in some critical states, it is of crucial importance to apply such methodological approach to accurately calculate them for all potential critical positions likely to be encountered during the useful life of a reactor.

Remote-controlled micro locking mechanism for plate-type nuclear fuel used in upflow research reactors

Fuel locking mechanisms (FLMs) are essential in upward-flow research reactors to prevent accidental fuel separation from the core during reactor operation. This study presents a novel design concept for a remotely controlled plate-type nuclear fuel locking mechanism. By employing electromagnetic field analysis, we optimized the design of the electromagnet for fuel unlocking, allowing the FLM to adapt to various research reactor core designs, minimizing installation space, and reducing maintenance efforts. Computational flow analysis quantified the drag acting on the fuel assembly caused by coolant upflow. Subsequently, we performed finite element analysis and evaluated the structural integrity of the FLM based on the ASME boiler and pressure vessel (B&PV) code, considering design loads such as dead weight and flow drag. Our findings confirm that the new FLM design provides sufficient margins to withstand the specified loads.We fabricated a prototype comprising the driving part, a simplified moving part, and a dummy fuel assembly. Through basic operational tests on the assembled components, we verified that the manufactured products meet the performance requirements. This remote-controlled micro locking mechanism holds promise in enhancing the safety and efficiency of plate-type nuclear fuel operation in upflow research reactors.

Computational Fluid Dynamics Analysis of an Open-Pool Nuclear Research Reactor Core for Fluid Flow Optimization Using a Channel Box

A channel box installation in the IEA-R1 research reactor core was numerically investigated to increase fluid flow in fuel assemblies (FAs) and side water channels (SWCs) between FAs by minimizing bypasses in specific regions of the reactor core, which is expected to reduce temperatures and oxidation effects in lateral fuel plates (LFPs). To achieve this objective, an isothermal three-dimensional computational fluid dynamics model was created using Ansys CFX to analyze fluid flow distribution in the Brazilian IEA-R1 research reactor core. All regions of the core and realistic boundary conditions were considered, and a detailed mesh convergence study is presented. Results comparing both scenarios are presented in the percentage of use of the primary circuit pump. It is indicated that 21.4% of fluid bypass to unnecessary regions can be avoided with the channel box installation, which leads to the total mass flow from the primary circuit for all FAs increasing from 68.9% (without a channel box) to 77.6% (with a channel box). For the SWCs, responsible for cooling LFPs, an increment from 9.7% to 22.4%, avoiding all nondesired cross three-dimensional effects, was observed, resulting in a more homogeneous fluid flow and vertical velocities. It was concluded that the installation of a channel box numerically indicates an expressive mass flow increase and homogeneous fluid flow distribution for flow dynamics in relevant regions. This gives greater confidence to believe that lower temperatures, and consequently oxidation effects in LFPs, can be expected with a channel box installation.

Uncertainty analysis of a typical research reactor performance for radioisotope production

The accurate neutron flux in the different regions of irradiation positions is very important parameter in any research reactor performance with highly effects on uncertainty of produced activities. The produced activities in a research reactor core is highly dependent on 3 main groups of activity measurement, reactor and sample. The significance of these parameters is comprehensively investigated and determined numerically for a Tehran Research Reactor (TRR) as a typical research reactor. This is done using precise MCNPX 2.6.0 simulations, several experiments during reactor operation, and measurements in laboratory. The sample’s location uncertainty is one of the main parameters in the neutron flux uncertainty. Further, shadowing of samples and control rods, samples density, self-shielding, and also power measurement mechanism could be noticeable and give rise to distinctive uncertainty in the final results. It is determined that all the considered parameters could lead up to 17.8% uncertainty in the final results at the TRR which is remarkable.

Comparative evaluation of the performance of improved genetic algorithms and differential evolution for in-core fuel management of a research reactor

This paper presents a comparative evaluation of the performance of Genetic Algorithm (GA) and Differential Evolution (DE) algorithm applied to in-core fuel management of the DNRR research reactor. Two GA variants corresponding to two selection operators, i.e., tournament (GA1) and roulette wheel (GA2) selections, respectively, with two-point crossover and scramble mutation were implemented for the ICFM problem. A comprehensive survey of the GA control parameters such as population size, crossover-type, mutation probability, and elitist archive size has been conducted to optimize the performance of the GAs. The basic DE was implemented with a standard mutation strategy DE/rand/1/bin. Numerical computations were performed based on the DNRR research reactor core loaded with 100 highly enriched uranium fuel (HEU) bundles for evaluating the performance of the GA and DE algorithms. Two main objectives were included in the fitness function to maximize the fuel cycle length and flatten the power distribution. The performance of the two GA variants and the basic DE was investigated with the same population size, fitness function, and convergence criterion. Each method was performed with 50 independent runs, and the best fitness values were collected for statistical analysis using Kruskal–Wallis and Mann–Whitney tests in comparison among the three methods. The statistical analysis shows that the performance of GA1 with tournament selection and DE are not significantly different and are better than GA2 with roulette wheel selection. DE is stable and efficient in exploring the search space to approach the global optimal solution in most runs. While, GA1 and GA2 were trapped at local optima by about 26% and 38%, respectively. However, the best solutions obtained with GA1 and GA2 after 50 independent runs are better than that obtained with DE in term of fitness values. This suggests an improvement of the basic DE is needed to maintain the potential good solutions during the evolution process.

Safety-related parameters calculation for OPAL reactor using MCNP v6.1.0 and ENDF/B-VII.1

Despite the standard cell-core approach remains as workhorse for the research reactor core’s design and analysis, several specific calculations are regularly done using Monte Carlo codes. This work presents a tailored validation of the well-known MCNP6 code for calculations within compact-core research reactors using parallel-plate with low-enriched uranium fuels, cooled and moderated by light-water and reflected by heavy-water. For such purpose, MCNP v6.1.0 and nuclear data from ENDF/B-VII.1 are here used for the calculation of safety-related parameters of the state-of-the-art OPAL research reactor, providing a detailed comparison with available experimental data. This analysis covers a wide range of parameters for fresh and burned cores, where for the latter case compositions from a deterministic (i.e. cell-core) approach are inserted into MCNP models. As a result, the aptness of the MCNP6 code to reproduce key experimental data from OPAL reactor likewise the expected level of accuracy for analogous applications are here addressed.

Heat transport and chimney design in a typical MTR reactor during natural convection cooling regime

In the present work, three dimensional CFD analysis is presented for a typical MTR reactor coolant loop during natural convection cooling regime. Design optimization studies are carried out. A model consisting of the core, the chimney, and the cold legs is simulated. The research reactor core is simulated as a porous zone with a cosine shape heat distribution. The chimney dimensions are varied to optimize its design for best thermal-hydraulic performance under natural convection conditions. The study is carried out for chimney extension ratio in the range of 1.5 ≤ Lch/Lcore ≤ 7.2 and chimney opening (expansion) ratio in the range of 1 ≤ B/b ≤ 2.5. Contours plot of temperature at different values of chimney extension and opening ratios were illustrated. The Nusslet number, coolant mass flow rate, maximum reactor core temperature, coolant velocity and core pressure drop are correlated as a function of the extension and opening ratios. Comparison with the results of CONVEC and RELAP5 codes shows a good agreement. The recommended optimum design conditions are for Lch/Lcore greater than 3.5, and B/b = 1 for best thermal-hydraulic performance.

Multiobjective Core Reloading Pattern Optimization of PARR-1 Using Modified Genetic Algorithm Coupled with Monte Carlo Methods

In order to maximize both the life cycle and efficiency of a reactor core, it is essential to find the optimum loading pattern. In the case of research reactors, a loading pattern can also be optimized for flux at an irradiation site. Therefore, the development of a general-use methodology for core loading optimization would be very valuable. In this paper, general-use multiobjective core reloading pattern optimization is performed using modified genetic algorithms (MGA). The developed strategy can be applied for the constrained optimization of research and power reactor cores. For an optimal reactor core reloading design strategy, an intelligent technique GA is coupled with the Monte Carlo (MC) code SuperMC developed by the FDS team in China for nuclear reactor physics calculations. An optimal loading pattern can be depicted as a configuration that has the maximum keff and maximum thermal fluxes in the core of the given fuel inventory keeping in view the safety constraints such as limitation on power peaking factor. The optimized loading patterns for Pakistan Research Reactor-1 (PARR-1) have been recommended using the implemented strategy by considering the constraint optimization, i.e., to maximize the keff or maximum thermal neutron flux while maintaining low power peaking factor. It has been observed that the developed intelligent strategy performs these tasks with a reasonable computational cost.

A methodology to enhance thermal neutron flux in Tehran Research Reactor core for domestic fuel test purposes

Performance of newly fabricated nuclear fuels must be tested under desired neutronic and thermal-hydraulic operating conditions prior to utilization in the reactor core. Tehran research reactor (TRR) as an MTR-type reactor is the only operating research reactor in the country that can potentially be used for irradiation tests on domestic fuels. In this study, compacting TRR core configuration, as a method to increase thermal neutron flux in the central in-core irradiation position to provide desired neutronic environment for fuel irradiation tests, is of concern. Several design criteria and limitations, including preservation of enough and suitable irradiation positions in the core for routine radioisotopes production activities of the reactor, have been considered to develop a new compact core configuration for TRR in which, desired linear heat generation rate is achieved in the test fuels and all the neutronic and thermal-hydraulic safety parameters of the proposed configuration are in accordance with the acceptance criteria stated in the reactor FSAR.

Online optical refractive index measurement in research reactor core

There is a growing interest in fiber optic measurements for applications in radiation environments. Optical fiber sensors and diagnostics can monitor many parameters of interest inside a research reactor core. For some applications, fiber optics are combined with an optical system that collects or focuses the light beam. The Radiation-Induced-Refractive-Index-Change (RIRIC) of the used glasses appears then as major phenomenon as it is a determining value for the sensor optical function. In the framework of the development of a radiation hardened confocal chromatic sensor, we implemented an on-line refractive index measuring device in order to test in a reactor core various glasses, candidates to be implemented into the sensor. The measurement relies on interferometry, which is a challenge because of the small volume available, the impossibility to make optical adjustments once installed, and the high temperature of operation. Precisely, the quantity retrieved is the optical path (product of the length L by the optical refractive index n), when L is well known, we can deduce n. But under high neutron fluence, some variation in density can be observed. The targeted online measurement of the refractive index therefore becomes an optical path measurement. We will present the device, the principle of measurement, and the first results of some index change measurement, produced by a temperature ramp from 20 °C and 350 °C. We have obtained original data for most of the candidate glasses used to design the optical system.

Confocal chromatic sensor for displacement monitoring in research reactor

Confocal chromatic microscopy is an optical technique allowing measuring displacement, thickness, and roughness with a sub-micrometric precision. Its operation principle is based on a wavelength encoding of the object position. Historically, the company STIL based in the south of France has first developed this class of sensors in the 90’s. Of course, this sensor can only operate in a sufficiently transparent medium in the used spectral domain. It presents the advantage of being contactless, which is a crucial advantage for some applications such as the fuel rod displacement measurement in a nuclear research reactor core and in particular for cladding-swelling measurements. The extreme environmental conditions encountered in such experiments i.e. high temperature, high pressure, high radiations flux, strong vibrations, surrounding turbulent flow can affect the performances of this optical system. We then need to implement mitigation techniques to optimize the sensor performance in this specific environment. Another constraint concerns the small volume available in the irradiation rig next to the rod to monitor, implying the challenge to conceive a miniaturized sensor able to operate under these constraints.

Concentration Measurement of 12 Elements in Five Herbal Plants Using Neutron Activation Analysis Approach

Introduction: Nowadays, many people use medicinal plants to manage diseases; therefore, detailed knowledge of the type and level of elements present in these plants is of prominent importance.The present study aimed to determine the weight fraction of 12 elements in the five most common medicinal plants in Iran. The names of these plants are caraway (Carum carvi), savory (Satureja hortensis), purslane (Portulaca oleracea), fenugreek (Trigonella foenum-graecum), and milk thistle (Silybum marianum) which were purchased from herbal pharmacies. Material and Methods: The neutron activation method was used to determine the elements. In the current study, neutrons from the research reactor core in Tehran, Iran were used and gamma spectra from radionuclides were recorded using a high purity germanium detector. The mass fractions of 12 elements were determined in the five abovementioned plants. Results: Caraway had the maximum amounts of elements of Fe (8,789 ppm), Cr (8 ppm), and Na (517 ppm) among the selected plants. The savory contained maximum levels of Mn (95 ppm), Cl (3,702 ppm), Ca (18,328 ppm), K (21,562 ppm), and V (2.7 ppm) and the lowest amount of Fe (195 ppm), Zn (13 ppm), Ca (2,243 ppm), Al (99ppm), Mn (26 ppm), and Mg (177ppm) were observed in fenugreek. Conclusion: The highest levels of Cr and Mg were obtained for caraway (8 ppm) and pursalne (3,915 ppm), respectively. These elements can help to decrease blood cholesterol and triglyceride levels. Furthermore, the results showed that these herbs were rich in essential nutrients for metabolic functions.

Measurement of core flow distribution in a research reactor using plate-type fuel assembly

In the present study, the core flow distribution of research reactors that used a plate-type fuel was investigated by measuring the flow rates at all fuel assembly locations to determine the flow uniformity in the reactor core. A specially instrumented dummy fuel assembly called probe fuel was developed to measure the flow rate at each fuel assembly location. The probe fuel is exactly the same as the real one, except the pressure taps and the embedded pressure conduits for measuring the pressure drop. First, hydraulic experiments were conducted to find out the pressure drop dependency on the flow rate of the probe fuel in a single fuel assembly test loop. Three correlations between the pressure drop and flow rate, which considers the water temperature effect, were derived from the measurement data of the hydraulic experiments. Using the probe fuel, the flow rates of all fuel assembly locations in the research reactor core were obtained from the measured pressure drop of the probe fuel using the correlations. The investigation of the core flow distribution in the downward forced-cooled research reactor was performed under a nominal flow condition. The flow rate measurements at each fuel assembly location were conducted for two vertical control absorber rod (CAR) positions. The measurement results showed that the primary coolant flow in the reactor core was distributed evenly with a small deviation of −2.9%~+2.3%, even though the CAR position changed. However, the ratio of the core flow and the bypass flow in the reactor was changed slightly with the position of the CARs, which seemed to be due to the changes in the flow resistance depending on the CARs’ position. In addition, the total flow rate of the primary cooling system (PCS) also increased a little when the condition of the CARs was fully withdrawn.

Active-sterile neutrino mixing constraints using reactor antineutrinos with the ISMRAN setup

In this work, we present an analysis of the sensitivity to the active-sterile neutrino mixing with the Indian Scintillator Matrix for Reactor Anti-Neutrino (ISMRAN) experimental set-up at very short baseline. In this article, we have considered the measurement of electron antineutrino induced events employing a single detector which can be placed either at a single position or moved between near and far positions from the given reactor core. Results extracted in the later case are independent of the theoretical prediction of the reactor anti-neutrino spectrum and detector related systematic uncertainties. Our analysis shows that the results obtained from the measurement carried out at a combination of the near and far detector positions are improved significantly at higher $\Delta m^{2}_{41}$ compared to the ones obtained with the measurement at a single detector position only. It is found that the best possible combination of near and far detector positions from a 100 MW$_{th}$ power DHRUVA research reactor core are 7 m and 9 m, respectively, for which ISMRAN set-up can exclude in the range 1.4 $eV^{2} \leq \Delta m^{2}_{41} \leq$ 4.0 $eV^{2}$ of reactor antineutrino anomaly region along with the present best-fit point of active-sterile neutrino oscillation parameters. At those combinations of detector positions, the ISMRAN set-up can observe the active sterile neutrino oscillation with a 95$\%$ confidence level provided that $\sin^{2}2\theta_{14}\geq 0.09$ at $\Delta m^{2}_{41}$ = 1 eV$^{2}$ for an exposure of 1 ton-yr. The active-sterile neutrino mixing sensitivity can be improved by about 22\% at the same exposure by placing the detector at near and far distances of 15 m and 17 m, respectively, from the compact proto-type fast breeder reactor (PFBR) facility which has a higher thermal power of 1250 MW$_{th}$.

Calculation of effective point reactor kinetic parameters combined with the Galerkin finite element method

Precise point reactor kinetic parameters are essential in the study of reactor dynamics. Point reactor kinetic parameters include the effective delayed neutron fraction and prompt neutron lifetime. In this work, effective point reactor kinetic parameters, which can be applied to unstructured grids, were calculated based on the Galerkin finite element method. First, two-dimensional and three-dimensional benchmarks were used to verify the calculation of steady-state neutronic parameters. Then, the Tehran research reactor core was divided into hexahedral meshes, and the forward flux and adjoint flux were calculated. Finally, the effective point reactor kinetic parameters of the Tehran research reactor were obtained by comprehensively processing the steady-state neutronic parameters.