Control Rod Research Articles

Preliminary analysis of in-orbit operation accidents for an ultra-small Lithium-cooled space reactor power system

To study the safety characteristics of an ultra-small (5.2 MWt) lithium-cooled space reactor power system (ULCR) under postulated accidents, a transient analysis code for the ultra-small space reactor coupled with a closed Brayton cycle dynamic conversion and heat pipe radiator was established. A preliminary analysis of four typical postulated accidents occurring at 100 percent power in-orbit operation of the ULCR system was conducted to define postulated problem areas and establish safety requirements. The postulated accidents mainly include (1) reactivity insertions, (2) coolant partial flow loss, (3) turbomachinery failure, and (4) radiator partial failure. The results showed that: a) since the ULCR system was designed to operate at high-temperature conditions (coolant temperature greater than 1500 K), the safety margin of the ULCR system was small. Emergency shutdown control or reduced power operation action should be taken proactively to ensure ULCR's safety in different accidents. b) When the control rod was misoperated and large reactivity was inserted, the maximum temperature of the lithium coolant was very close to the saturation temperature (the difference was 24 K), and the system was operated at extremely high temperatures (the average temperature of the core coolant was 1750 K); c) In coolant flow loss accident, when the coolant flow of the system dropped below 40% of the rated mass flow, the lithium coolant boiling would occur in the primary loop system, and the emergency shutdown control action was required to ensure the safety of the system.

Application of new axial power distribution synthesis method using in-core detector signal in digital core protection system

ABSTRACT A new method for the generation of axial power distributions using artificial neural network (ANN) technique and real-time measured planar radial peaking factor, Fxy (hereinafter ‘Live Fxy’) method for digital core protection system of pressurized water reactors is presented. ANN with Simulated Annealing (SA) technique to find global optimum solution was used to calculate core average axial power distribution. In Live Fxy method, axial node-wise Fxys were calculated by multiplying pin/box factors to the measured assembly powers inferred from in-core detector signals without calculating detailed 3D power distributions. The hot pin power distribution for DNBR and LPD was then obtained by multiplying the core average axial power distribution and Fxys at axial nodes. The validation of the method was performed for various core conditions (e.g. core power levels, control rod positions, etc.) of Korean OPR1000 power plant. The result showed a decrease in RMS errors by 2.2% ~ 7.0% (3.6% on average), and the minimum thermal margins for DNBR and LPD were increased by 6.4% and 15.6%, respectively. The application of the method would improve the accuracy of axial power distribution and thermal margin, contributing to the operability and safety of digital core protection system.

A Multisource Uncertainty Fusion Reliability Evaluation Method for the Control Rod Drive Mechanism of Nuclear Power Plants

The reliability of a pressurized water reactor power plant’s control rod drive mechanism (CRDM) is affected by many factors, such as operation states, unit performance, and dynamic environments. Multiple sources of uncertainties, including random, interval, and fuzzy, exist when analyzing the reliability of CRDMs. Modeling reliability without considering the fusion of multisource uncertainties may result in model distortion and may not provide realistic assessment results. This paper proposes a multisource uncertainty fusion method powered by the margin-based variable transformation and the Bayesian network propagation. The interval and fuzzy variables are transformed into random variables to obtain the corresponding generalized probability density function of the CRDM’s feature parameters. After that, the CRDM’s reliability can be evaluated via probability quantization of the Bayesian network inference result through sampling algorithms. A magnetic-jack type CRDM case is presented to verify the proposed method, and the results show that this method can fuse three types of uncertainty variable in a unified way and effectively obtain reliability evaluation results.

Determination of 10B/11B in boric acid and B4C using LA-ICPMS

Abstract Boron isotopic ratio measurement is very important in the nuclear industry because it is used as neutron poison (in the form of boric acid or a B4C pellet) and control rod material (B4C pellet) in nuclear power plants. Since 10B has a higher neutron absorption cross-section, enriched 10B is used in the nuclear industry and the extent of enrichment varies from 50–90 %. Therefore, it is essential to determine the 10B/11B ratio in boric acid and B4C. The prime purpose of the present study is to utilize LA-ICPMS for evaluating 10B/11B ratio both in solution (boric acid) and solid (B4C pellet) without any pre-treatment so that the difficulty in dissolution or powdering process of B4C (one of the hardest materials) can be avoided and the B4C pellet can be reused. The results obtained from LA-ICPMS for the B4C pellet are discussed and compared with the solution ICPMS, TIMS, and LIMS. The solutions of boric acid (natural abundance) and dissolved B4C (natural & ∼67 % enriched isotopic composition) were analyzed by LA-ICPMS and validated by liquid sampling ICPMS data.

Control of vortex shedding and acoustic resonance of a circular cylinder in cross-flow

This study experimentally investigates the effectiveness of a control rod in suppressing self-excited acoustic resonance within a range of Reynolds numbers (Re) spanning from 2.1 × 104 to 1.6 × 105. The investigation focuses on specific parameters, including diameter ratio (d/D) values of 0.1, 0.2, and 0.3; gap ratio (G/D) values of 0.05, 0.1, and 0.2; and angular positions (θ) ranging from 0 to 180 degrees. Comparative analyses are conducted between cases featuring the control rod and a reference case (base case) without it. The near-wake flow field is characterized using Particle Image Velocimetry (PIV), and aeroacoustic response measurements are employed to quantify the aeroacoustic noise emission, particularly during self-excited acoustic resonance. Simultaneous measurements of fluctuating lift force and aeroacoustic response measurements, facilitate the quantification of energy transfer from the flow field to the acoustic field during self-excited acoustic resonance. The results reveal that the control rod’s placement significantly impacts the Strouhal periodicity, with outcomes heavily dependent on the rod’s angular orientation. At certain angular positions, the control rod reduces the sound pressure level (SPL) generated during acoustic resonance excitation. However, at different angular positions, the rod exacerbates resonance excitation. This variability is attributed to the control rod’s profound influence on the vortex core formation and the energy transfer mechanism during acoustic resonance.

Multi-physics code coupling methodology for the simulation of accidental transients in sodium cooled fast reactors

The French Alternative Energies and Atomic Energy Commission (CEA) has played a substantial role in advancing fourth-generation reactor technologies, including Sodium-cooled Fast Reactors (SFR), which are anticipated to offer enhanced efficiency, adaptability, and sustainability compared to existing nuclear technologies. In the context of safety assessment, it is crucial to precisely simulate accidental scenarios like Unprotected Loss Of Flow (ULOF) transients. These transients entail intricate multi-physics interactions within the reactor, particularly within the core, encompassing thermal-hydraulics, reactor physics, and fuel performance (e.g. changes in densities or temperatures will lead to neutronics effects, thus power, then causing feedbacks onto said densities and temperatures).In order to capture those complex phenomena, a multi-physics coupling approach for SFRs, which combines the CATHARE3 code for system scale thermal-hydraulics, the APOLLO3 code for reactor physics, and the GERMINAL V2 code for nuclear fuel performance has been developed at CEA in the recent years.This paper presents a comprehensive methodology of such coupling and its integration into a Verification, Validation and Uncertainty Quantification (VVUQ) process. The first part of the paper provides an introduction to the simulation tools used for the coupling. The following section explains the multi-physics coupling approach, facilitated by CEA’s Collaborative Code Coupling Platform (C3PO), which enables code integration through a Python interface to create a comprehensive simulation model. A detailed examination of the coupling algorithm and data transfer mechanisms is also provided.The generality and reliability of this approach are demonstrated by simulating two accidental transients: the Unprotected Loss Of Flow (ULOF) in the ASTRID reactor and the Loss Of Flow WithOut Scram (LOFWOS) in the Fast Flux Test Facility (FFTF), both corresponding to a primary pump trip without control rods falling. The results show that the proposed multi-physics coupling scheme is effective and robust for SFR simulations. The coupled model enables a more accurate and realistic representation of the complex phenomena occurring in the reactor during an accident.This research underscores the significance of multi-physics coupling in the analysis of SFRs and serves as a valuable point of reference for future SFR studies. The proposed approach offers an efficient way to investigate SFR responses during accidents or incidents, thereby advancing the development of dependable and precise simulation tools for SFR design and evaluation.

Seismic responses of a fuel assembly in nuclear heating reactor

The fuel assembly (FA) stands as a pivotal component within nuclear reactors, impacting safety, reliability, and economic viability of the entire facility. In the 200 MW nuclear heating reactor (NHR200-II), the FA embodies a novel structural configuration, characterized by a 9 × 9 fuel bundle arrangement enclosed within a singular channel. During seismic events, challenges like potential FA damage and control rod insertion failures can pose safety risks. Therefore, a comprehensive structural integrity assessment is essential. The seismic responses of the NHR200-II FA unidirectionally and bidirectionally under different boundary conditions, using a time-compressed El-Centro N-S earthquake (1.5 times amplification, ∼0.36 g peak acceleration), are analyzed in detail. We examine stress distribution, channel-spacer grid interactions, dynamic clamping, acceleration responses, and response spectra. Results confirm the robust seismic performance of NHR200-II FA, with potential for future seismic fluid–structure interaction analysis based on the model used in this work.

Reactor core design with practical gadolinia burnable absorbers for soluble boron-free operation in the innovative SMR

The development of soluble boron-free (SBF) operation in the innovative Small Modular Reactor (i-SMR) requires effective strategies for managing excess reactivity over extended operational cycles. This paper introduces a practical approach to reactor core design for SBF operation in i-SMR, emphasizing the use of gadolinia burnable absorbers (BA). The study investigates the feasibility of Highly Intensive and Discrete Gadolinia/Alumina Burnable Absorber (HIGA) rods for controlling excess reactivity sustainably. Through comprehensive analysis and simulations, the reactivity behavior with varying quantities of HIGA rods is examined, leading to the development of optimized fuel assembly designs. Furthermore, the integration of HIGA rods with integral gadolinia BA rods is discussed to enhance reactivity control and operational flexibility further. This approach utilizes the spatial self-shielding effect of gadolinia for extended reactivity management, crucial for stable and efficient reactor performance. The paper thoroughly addresses core design considerations, including fuel assembly configurations and control rod patterns, to ensure safety and performance in initial and reload cycles. This research advances the development of SBF operation in i-SMR by offering practical reactivity management solutions.

Adoption of a fast dynamic model to simulate transient operations of MNSR

After developing a thermal–hydraulic model, it should be coupled to a point kinetic model to completely simulate transient behavior of a miniature neutron research reactor (MNSR). In this study introduction and validation of such a dynamic model which could be used as computing source of a micro-simulator, is presented. A time dependent point kinetic model in association to a simple control algorithm and reactivity feedbacks calculation (poisons and temperature) is adopted. In order to validate the model, it is compared against MNSR operational data in different scenarios such as normal fixed power operations, and reactivity insertion operations, and post shutdown scenarios. Result indicates that the coupled model could estimate the critical rod position for different power levels with an error less than 5%. Although the control algorithm is simple, it could predict the actual trend of control rod position accordance to the actual data. The coupled model simulates different reactivity insertion accident similar to the reactor operations in such a way that the error between estimated power peak, peak time, and long-term power is less than 5%.

Materials Science Analysis of RCCA Structural Materials and Calculations for Lifetime Extension

The paper considers the possibility of lifetime extending of the rod control cluster assemblies in operation by determining their residual lifetime using an approach that takes into account the actual position of the rod control cluster assembly in the core. Excessive conservatism was introduced by assuming that the immersion of the 10th working group in the core during all fuel campaigns is 70 %, while the actual position is 85-90 %. Taking into account the actual position of the rod control cluster assemblies indicates much less irradiation of the lower, most loaded part of control rod cladding and much less burnup of the 10B compared to 70 % immersion of the rod control cluster assemblies, since they are in the area of lower neutron flux density.
 The results of a comprehensive analysis of available studies and test results to determine the phenomena that can potentially approach the achievement of the boundary state of the structural components or limit the lifetime of control rods and rod control cluster assemblies in general are presented. Such phenomena include general corrosion of structural materials (42CrNiMo, Cr18Ni10Ti) (decrease in wall thickness, increased transfer of radioactive corrosion products the coolant); radiation resistance (degradation of mechanical characteristics) of control rod cladding materials (42CrNiMo) and other structural components of the rod control cluster assemblies (42CrNiMo, CrKh18Ni10Ti); radiation resistance of neutron-absorbing materials (B4C, Dy2O3∙TiO2); compatibility of absorbing materials with the control rod cladding.
 Based on the results of post-irradiation examinations of control rods tested in research reactors as dummy rods or rod control cluster assembly rods operated in the VVER-1000 core, the following boundary values were established that can be used as criteria for maintaining the integrity of the control rod cladding: maximum fast neutron fluence in the lower, most loaded part of the control rod cladding (in the weld area of the lower end plug and the cladding) should not exceed 34×1021 n/cm2; maximum burnup of the 10B isotope should not exceed 45 %. If these two criteria are met, the possibility of other potentially damaging phenomena is too low.
 According to the calculations that use the schedule of changes in the position of the rod control cluster assemblies of the control group and the power of one of the power units during one fuel cycle, one can preliminary conclude that the proposed approach will extend the lifetime of almost all rod control cluster assemblies from 25500 to 38000 hours in the automatic control group and from 75600 to 113500 hours in the shutdown group. The residual lifetime should be assessed for a particular power unit and each rod control cluster assembly separately, since each rod control cluster assembly has its own operating history.

Integrating core physics and machine learning for improved parameter prediction in boiling water reactor operations

This study introduces a novel method for enhancing Boiling Water Reactor (BWR) operation simulations by integrating machine learning (ML) models with conventional simulation techniques. The ML model is trained to identify and correct errors in low-fidelity simulation outputs, traditionally derived from core physics computations. These corrections aim to align the low-fidelity results closely with high-fidelity data. Precise predictions of nuclear reactor parameters like core eigenvalue and power distribution are crucial for efficient fuel management and adherence to technical specifications. Current high-fidelity transport calculations, while accurate, are impractical for real-time predictions due to extensive computational demands. Our approach, therefore, utilizes the standard two-step simulation process-assembly-level lattice physics calculations followed by whole-core nodal diffusion computations-to generate initial results, which are then refined using the ML-based error correction model. The methodology focuses on improving simulation accuracy in regular BWR operations rather than developing a universal ML predictor for reactor physics. By training an advanced neural network model on the difference in high-fidelity and low-fidelity simulations, the model can reduce the nodal power error from low-fidelity simulations to around 1% on average and the core eigenvalue down to under 100 pcm. This result is under the condition of the normal variations of control rod pattern and core flow rate changes in standard BWR operations used in the training and evaluation of the machine learning model. This work suggests a promising approach for achieving more accurate, computationally feasible simulation solutions in nuclear reactor operation and management.

Thermodynamic equilibrium state calculations for oxidation and corrosion reactions of B4C and oxide-based neutron absorber compounds in reactor control rods

The challenges associated with conventional boron carbide (B4C) control rod materials, including helium gas accumulation and susceptibility to oxidation and corrosion in various environments, have been thoroughly explored. To address these issues, a comprehensive investigation into the potential of oxide-based neutron absorber compounds for control rods has been undertaken. Thermodynamic equilibrium state calculations have been conducted to assess the oxidation and corrosion reactions of various neutron absorber materials (Gd, Hf, Sm, Er, Eu, Dy-based oxide), including B4C, in various environments such as oxygen, steam, air, and air–steam mixtures. The results have unveiled vulnerabilities of B4C in these environments, notably the generation of gases such as H2, CO, and boron compounds. Furthermore, neutron absorber oxide materials have exhibited potential susceptibility to oxidation and corrosion in steam environments. Consequently, the potential of oxide-based neutron absorber compounds, formed by combining neutron absorber oxide materials with highly oxidation-resistant substances (ZrO2, TiO2), has been explored. Thermodynamic equilibrium state calculations indicate that these compounds maintain robust resistance to oxidation and corrosion across various environments. This paper demonstrates the superiority of oxide-based neutron absorber compounds as alternatives to existing boron carbide neutron absorber materials. Additionally, the oxide-based neutron absorber compounds are expected to extend the lifetime of the LWR control rod as well as improve stability and resistance to oxidation and corrosion.

New RMC Energy Deposition Treatment and its Application in Multi-Physics Simulation

Accurate estimation of energy deposition is important in core physics and severe accident analyses for design optimizations. In this study, a new energy deposition treatment is implemented in the Reactor Monte Carlo (RMC) code, offering multiple modes with varying levels of fidelity and computational requirements. The most precise mode is utilized in coupling simulations between RMC and the subchannel thermal-hydraulic analysis code SUBCHAN, incorporating an explicit moderator heating fraction in the coupling interface. The new treatment is verified against references from MCNP, Serpent, and OpenMC for three light water reactor (LWR) assembly cases, and great agreement is achieved. Energy deposition in different materials and components is emphasized in Kilowatt Reactor Using Stirling TechnologY (KRUSTY) modeling, and the results obtained using different modes are compared. The RMC-SUBCHAN coupling calculations for the three LWR assembly cases, employing the most accurate model, reveal a maximum increase of 94.6 K in the control rod centerline temperature, with a normalized energy deposition of 35.9% in the control rod regions. In the assembly case with gadolinium (Gd) burnable poison, a temperature increase of 7.3 K is observed in the Gd rod centerline, while the coolant outlet temperature decreases by 1.6 K due to the reduced explicit moderator heating fraction of 2.1%, compared to the constant 2.6% in the previous coupling scheme.

Deep Learning Approach to Within-Bank Fault Detection and Diagnostics of Fine Motion Control Rod Drives

Control rod motion is one of the primary means of regulating the rate of fission in a nuclear reactor core to ensure safe and stable operation. Reactor power distribution and thermal power output can be fine-tuned by adjusting the control rod position. For high-precision control of rod movements, Fine Motion Control Rod Drives (FMCRDs) are often used. The operation of FMCRDs provides a unique opportunity to implement condition monitoring related to the intermittency of motion and the use of control rod banks. This research sets out to detect three types of faults in an electrically driven FMCRD. In addition to detecting faults, this work will attempt to determine both the type of fault and the source of each fault, completing the fault detection and diagnostics (FDD) pipeline on a scarcely researched system. The three types of faults to be investigated are short-circuit faults, ball screw wear faults, and ball screw jam faults. This is a potential advancement to the within-bank FDD of this specific drive system intended for deployment in an advanced nuclear reactor plant. Using encoder-decoder structured convolutional neural networks and autoencoders, the three tested faults were confidently detected and isolated as well as reasonably diagnosed by monitoring the FMCRD servomotor torque.

Pitfalls in Using Isotopic Distributions for Structural Interpretation by Mass Spectrometry.

This brief cautionary note reports a failure in a common and useful assumption, namely, that the isotopes of the elements occur in their natural abundance ratios in commercially sourced organic compounds. Some commercial sources of tris(pentafluorophenyl)borane, B(C6F5)3, show severely depleted 10B, while materials from other suppliers display natural isotopic abundances. The depletion varies from lot-to-lot, and it was confirmed by inductively coupled plasma (ICP) mass spectrometry. The isotope 10B is used in the nuclear power industry, as a neutron absorber in the power control rods. It is speculated that the residual 11B generated when preparing 10B-enriched boron carbide for control rod use, provides 11B-rich raw material that is then used for commercial B(C6F5)3 synthesis.

Optimal controller design for reactor core power stabilization in a pressurized water reactor: Applications of gold rush algorithm.

Nuclear energy (NE) is seen as a reliable choice for ensuring the security of the world's energy supply, and it has only lately begun to be advocated as a strategy for reducing climate change in order to meet low-carbon energy transition goals. To achieve flexible operation across a wide operating range when it participates in peak regulation in the power systems, the pressurised water reactor (PWR) NE systems must overcome the nonlinearity problem induced by the substantial variation. In light of this viewpoint, the objective of this work is to evaluate the reactor core (main component) of the NE system via different recent optimization techniques. The PWR, which is the most common form, is the reactor under investigation. For controlling the movement of control rods that correspond with reactivity for power regulation the PWR, PID controller is employed. This study presents a dynamic model of the PWR, which includes the reactor core, the upper and lower plenums, and the piping that connects the reactor core to the steam alternator is analyzed and investigated. The PWR dynamic model is controlled by a PID controller optimized by the gold rush optimizer (GRO) built on the integration of the time-weighted square error performance indicator. Additionally, to exhibit the efficacy of the presented GRO, the dragonfly approach, Arithmetic algorithm, and planet optimization algorithm are used to adjust the PID controller parameters. Furthermore, a comparison among the optimized PID gains with the applied algorithms shows great accuracy, efficacy, and effectiveness of the proposed GRO. MATLAB\ Simulink program is used to model and simulate the system components and the applied algorithms. The simulation findings demonstrate that the suggested optimized PID control strategy has superior efficiency and resilience in terms of less overshoot and settling time.

The dynamics of cantilevered structures subject to axial flow

The dynamics of slender cantilevered cylinders subjected to internal or external axial flow has been studied extensively from the 1960s onwards. In the early studies, the flow was directed from the clamped end towards the free end of the cantilever. The same is true for cantilevered plates (or “flags”) in axial flow. More recently, however, the dynamics in reverse axial flow, i.e., flow directed from the free towards the clamped end, has received attention, partly as curiosity-driven research, but also because of engineering applications. For example, cantilevered pipes aspirating fluid are used in ocean mining, cantilevered cylinders in reverse axial flow may be found as control rods in nuclear reactors, and cantilevered plates in reverse axial flow are a candidate system for energy harvesting.The present paper provides a summary of the dynamics of these systems in conventional and reverse axial flow and compares their dynamical behaviour. For example, cantilevered cylinders in conventional axial flow are subject to a weak static divergence (buckling) at sufficiently high flow velocities, and to vigorous flutter at higher flow velocities. Cylinders in reverse axial flow, on the other hand, are subject to weak flutter at low flow velocities and to a large amplitude static divergence at higher flows. In the first case the dynamics is sensitive to the shape of the free end, and in the second hardly at all.The differences in the dynamical behaviour in reverse flow vis-à-vis conventional flow for pipes and plates, not so neatly reversed as for cylinders, are also discussed, and some general conclusions drawn for all three systems, regarding similarities and differences in the dynamics and sensitivity to free-end shape arising from the direction of fluid flow.

Whole core analysis of BEAVRS benchmark for hot zero power condition using MVP3 with JENDL-5

ABSTRACT With the release of the latest Japanese evaluated nuclear data library JENDL-5, the prediction accuracy of JENDL-5 for neutronics parameters of the BEAVRS benchmark for the hot zero power condition was evaluated in this study. The criticality, control rod bank worth (CRW), isothermal temperature coefficient (ITC), and in-core detector signals were calculated and compared with the measured data for evaluation. For the criticality, the calculation-to-measurement (C/E) values varied between 1.0001 and 1.0045. Sensitivity analysis by replacing cross-section data from the JENDL-4.0u1 with JENDL-5 revealed that 1H, 235U, 238U, and 16O significantly affected the criticality. The individual CRW agreed within 50 pcm, and total CRW also agreed within 100 pcm from the measured values. The ITC results calculated with a temperature deviation of 5.56 K case were negatively overestimated compared with the measured values, whereas those of with 2.78 K were improved and agreed with the measured values within a standard deviation. The axial detector signals indicated a maximum relative error of 4.46% and the root mean squared error (RMSE) of 2.13%. The differences between the previous version of JENDL-4.0u1 and JENDL-5 were also investigated.

Time-dependent neutron transport method based on three-dimensional method of characteristics

The 3D Method of Characteristics (MOC) has high numerical accuracy, geometrical flexibility, making it suitable for the simulation of advanced reactors with complex geometry, while the development of transient solvers based on it is still relatively limited. In this work, transient method was developed based on 3D MOC solvers of OpenMOC and predictor–corrector quasi-static method and the C5G7-TD benchmark problems were used to verify the implementation. In 2D cases and 3D moderator density change cases, the results of OpenMOC were in good agreement with the results of other transient codes. In 3D control rods movement cases, there are oscillation errors between the existing results of different transient codes due to the control rod cusping effect. This work is based on the refined axial modeling of 3D MOC, which avoids the cusping effect and provides new results of C5G7-TD benchmark.

OpenMC-based high-fidelity model for the Moroccan TRIGA Mark II research reactor

The Moroccan TRIGA Mark II serves as a pivotal cornerstone in the advancement of nuclear science and technology within the region. Precise modeling is crucial for ensuring safe and efficient reactor operation. OpenMC, known for its versatility as a stochastic code with a powerful Python API, offers unique capabilities for reactor analysis, optimization, and automation. This makes OpenMC the ideal choice for neutronic analysis in the development of multi-physics models aimed at enhancing the accuracy and reliability of reactor analysis. Leveraging its Python interface, we have developed a high-fidelity model for the Moroccan TRIGA Mark II, enhancing modeling precision. A central element of this study involved making crucial modifications to OpenMC’s source code, enabling the prompt method and calculating the βeff. These modifications are essential, representing a significant advancement in the capacity to conduct a comprehensive analysis of a wide range of critical neutronic parameters. These parameters include the neutron flux spectrum, keff, control rods worth and core excess reactivity, and power peaking factors. The results exhibit strong agreement with the performed experimental measurements and MCNP6.2 calculations.The findings underscore the model’s precision and reliability, making it a valuable asset for neutronic analysis and, particularly, for developing a comprehensive multi-physics model of the Moroccan TRIGA Mark II.