Control Rod Research Articles

Delayed neutron library analysis using dynamic control rod calibration method based on results from the VR-1 reactor experiment and Serpent transient calculation

This study is focused on a comparison of experimental neutron flux data from the VR-1 reactor with results from Serpent calculation. Serpent data were calculated by the dynamic part of the code and the steady-state criticality calculation. The Serpent calculation was performed with the JEFF 3.3 and ENDF/B-VIII.0 nuclear data library. This paper expands previous work and focuses on the integral control rod worth. Comparison of the ENDF/B-VIII.0 and JEFF 3.3 nuclear data libraries with experimental data is especially valuable. The comparison shows better performance of JEFF 3.3 nuclear data instead of ENDF/B-VIII.0 for the transient calculation of the cases presented in this article. In addition, the comparison of static and dynamic reactivity provided an interesting difference, especially for small rod insertion. Moreover, the study confirms that the Serpent dynamic external source simulation mode provides an accurate description of reactor behavior, and that the shape of the integral control rod worth is close to one experimentally measured.

Startup and Multiphysics Analysis of a Compact kW-Class Thermal Spectrum Microreactor

Self-regulating compact nuclear microreactor concepts are being developed for use in space and at terrestrial remote sites. The current emphasis is on the development of high-assay low-enriched uranium–fueled reactors that rely on metal hydrides to achieve size, specific weight, and power parity with the legacy highly enriched uranium systems. In the case of terrestrial applications, metal hydride–moderated systems can also improve economic feasibility. The neutronic design of metal hydride–moderated high-temperature reactor cores is complicated by the need to optimize and synchronize delayed spatial and temporal reactivity feedback from outer reflectors. Zebra is a thermal spectrum proof-of-principle core designed for demonstrating the reactor dynamics of hydride-moderated spectrum reactors, which is studied in this work. The reactor, including the fuel, yttrium hydride moderator, heat pipes, central control rod, structural supports, and the beryllium reflector, weighs ~425 kg. The design shares several features common with nuclear criticality test geometries routinely used in National Criticality Experiments Research Center (NCERC) experiments such that a prototype of the reactor can be readily assembled and tested. Additional features, such as heat pipes and hydrogen barrier clads, were added to minimize the potential for large thermal gradients that in turn could induce hydrogen loss or structural deformation during extended periods of operation at ~1000 K. In this work, normal and off-normal state performance of the Zebra rector core was analyzed using MCNP and Abaqus-based Reactor Multiphysics (MARM) software. Up-to-date nuclear, thermal, and mechanical performance data were used to characterize the performance of the yttrium hydride moderator. Steady-state analyses established that at a postulated power between 20 and 50 kW(thermal), the reactor core is nearly isothermal irrespective of quality of conduction coupling between heat pipes and fuel plates. Additionally, heat pipe failure modes, including simultaneous failure of all heat pipes in a quadrant of a reactor, were examined to bound the maximum credible temperature spike in the reactor core for extreme off-normal operating conditions. Finally, we detail the startup design challenges for the hydride-moderated thermal core, and analyze load-following cases to achieve “self-regulation” using the Dynamic Analysis of Reactor Transients module of MARM. The reactor is self-regulating with a reactivity temperature coefficient of ~−1 pcm/K at the operating point based solely on nuclear cross-section feedback. It can be further strengthened using additional spectral shift neutron absorbers and incorporating design features that enhance core expansion. This work captures that hydride-moderated systems are feasible for various self-regulating applications once systematic checks are verified in order to achieve a well-engineered core design.

Numerical study of flow past two square cylinders with horizontal detached control rod through passive control method

A two dimensional (2-D) numerical study is conducted through the lattice Boltzmann method for flow past two square cylinders with a detached control rod placed in the horizontal position of the channel. The range of Reynolds number (Re) is fixed at Re = 150, with the length (l) of the control rod varying from l = 0.1d–21d. First, we check the accuracy of grid points and the validity of the code for present problems by comparing the results with already available data in the literature. After that, we discussed the obtained results in terms of vorticity contour, drag and lift coefficients, and force statistics. In our study of flow structure mechanisms, we examined four types of flow regimes: a small length control rod, a small and moderate length control rod, a moderate length control rod, and a larger length control rod according to flow behavior. In force statistics, we calculated the values of CDmean, CDrms, CLrms, CDamp, CLamp, and St, respectively. The mean drag coefficient contains the smallest values at the largest length, i.e., l = 19 and 21d. The maximum value of CDmean is examined at the first cylinder (C1) as compared to the second cylinder (C2) at l = 2.25d, and it is 1.2789. The values of CDmean for C2 at the length of the control rod, l = 0.1d–1.25d, and at l = 21d are negative due to the effect of thrust. The value of the Strouhal number is the same for both cylinders C1 and C2, and its highest value is 0.1036. The pressure amplitude of the drag coefficient is maximum for C1, while the pressure amplitude of the lift coefficient is maximum for C2, and its value is 2.1569.

A PWR core power control using optimized PID controller with SQP based on control rod positioning and variable coolant flow rate

To ensure safety and stability, it is optimal for nuclear power plants to operate at their nominal power level. However, to meet the needs of power systems, NPPs must also be able to adjust their output for load-following purposes. This necessitates finding an effective control algorithmto maintain stable/safe operating conditions. The current study investigates the performance of PID controllersoptimized using Sequential Quadratic Programming algorithm and applied using two control techniques: control rod positioning and variable coolant flow rate. Simulation is done using model of four-loop Westinghouse PWR-1200 MW nuclear reactor. The proposed controller’s performance is evaluated in three scenarios:sudden control rod withdrawal, sudden change in coolant inlet temperature, and linear load tracking. The influences of model parametric uncertainty are also investigated. Simulation results revealed that the performance of the control algorithm applied using variable coolant flow rate was superior to the one usingcontrol rod positioning.

Neutronic Analysis of Accident-Tolerant Cladding Materials in 3D Full Core BEAVRS PWR Benchmark Using OpenMC Code

After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident-tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron-based alloys have been suggested as potential candidates for fuel cladding due to their favourable thermomechanical properties, lower reactivity with steam, and lower hydrogen generation. This study evaluates the neutronic performance of C26M (a 2nd generation nuclear grade FeCrAl alloy), APMT™, 310SS, and 304SS cladding materials by comparing them with Zircaloy-4 cladding in a 3D PWR core at the beginning of the cycle (BOC) using OpenMC code. The results revealed that the neutronic penalty varied for different alternative cladding materials where C26M exhibited the lowest neutronic penalty value of -12551 pcm, while 310SS demonstrated the highest with a value of -17855 pcm. Additionally, important parameters in the reactor core such as neutron spectrum, reactivity coefficients, boron worth, control rod bank worth, power distribution, and radial thermal neutron flux distribution are evaluated and discussed. The analysis results showed that C26M provided a significantly higher level of neutronic performance compared to APMT™, 304SS, and 310SS. Although this study primarily focused on the neutronic performance of PWRs at BOC, future research should encompass fuel depletion analysis to delve deeper into the potential of alternative cladding materials.

Practical Design of a Breed-and-Burn Lead-Cooled Fast Reactor with Rotational Fuel Shuffling Strategy

The purpose of this study is to demonstrate a practical core design for a lead-cooled, nitride fueled, rotational fuel shuffling breed-and-burn (RFBB) fast reactor. The core design is based on the Westinghouse Lead Fast Reactor (WH-LFR) and uses natural uranium nitride fuel with a sodium bond encased in oxide dispersion-strengthened steel cladding. Simulations confirmed the potential of the reactor to maintain criticality at the equilibrium state, with a reactivity swing of less than 200 pcm at every cycle interval and an average discharge burnup of 235 MWd/kg heavy metals (HM) for a 1050 effective full-power day refueling interval. Power profiles were maintained stable at the equilibrium state, while the cladding of the discharged fuel incurred over 650 displacements per atom over its entire residency in the core. From a nonproliferation perspective, the plutonium vector for the discharge fuel aligns with reactor-grade fuel standards, with over a 70% concentration of 239Pu and over 22% 240Pu, reducing the risk of weaponization. The adopted control rod system has been shown to offer sufficient negative reactivity of over 19 $ to bring the reactor into a subcritical state. Challenges such as the susceptibility of neutron balance to material thickness and neutron leakage have been addressed, emphasizing the necessity for meticulous design improvements. A steady-state thermohydraulic analysis confirmed the heat removal capacity from the hottest channel, ensuring operational safety. This study confirmed the feasibility of the RFBB strategy for a lead-cooled nitride-fueled fast reactor and sets a precedent for future research in enhancing fuel utilization and safety in nuclear reactors.

Performance evaluation and validation of OpenMC code for criticality analysis of an MTR-type research reactor

This paper assesses the performance of OpenMC, an open-source Monte Carlo code for particle transport, through criticality analysis of an MTR-type research reactor. The evaluation involves validation against experimental data from the NUR research reactor's first fresh fuel core and a code-to-code comparison with MCNP5, using a consistent methodology ensuring identical input conditions. Key parameters, including excess reactivity, control rod worths, shutdown margin, critical configurations, kinetics parameters, neutron flux, power distributions, and peaking factors, were examined. Results indicate good agreement between OpenMC simulations, experimental data, and MCNP5 results, with marginal differences well within acceptable ranges. Discrepancies with experimental data are below 5 %, considering measurement accuracy, while code-to-code comparisons reveal relative differences below 1 % in most cases. This study supports OpenMC's reliability for MTR reactor criticality analysis, affirming the validity of the developed NUR reactor model for future applications and highlighting the code's credibility and applicability in the nuclear reactor simulations.

Core Optimization for Extending the Graphite Irradiation Lifespan in a Small Modular Thorium-Based Molten Salt Reactor

The lifespan of core graphite under neutron irradiation in a commercial molten salt reactor (MSR) has an important influence on its economy. Flattening the fast neutron flux (≥0.05 MeV) distribution in the core is the main method to extend the graphite irradiation lifespan. In this paper, the effects of the key parameters of MSRs on fast neutron flux distribution, including volume fraction (VF) of fuel salt, pitch of hexagonal fuel assembly, core zoning, and layout of control rod assemblies, were studied. The fast neutron flux distribution in a regular hexagon fuel assembly was first analyzed by varying VF and pitch. It was demonstrated that changing VF is more effective in reducing the fast neutron flux in both global and local graphite blocks. Flattening the fast neutron flux distribution of a commercial MSR core was then carried out by zoning the core into two regions under different VFs. Considering both the fast neutron flux distribution and burnup depth, an optimized core was obtained. The fast neutron flux distribution of the optimized core was further flattened by the rational arrangement of control rod channels. The calculation results show that the final optimized core could reduce the maximum fast neutron flux of the graphite blocks by about 30% and result in a more negative temperature reactivity coefficient, while slightly decreasing the burnup and maintaining a fully acceptable core temperature distribution.

Basal and prismatic slip in hafnium

Metals with hexagonal crystal structure are found in many safety-critical applications, from titanium alloy fan blades in jet engines to magnesium alloys in the automotive industry. They also exist in nuclear reactors, for example hafnium alloy control rods whose structural integrity is critical to nuclear safety. However, mechanistic understanding of the micromechanical properties and deformation behaviour of Hf remains elusive. To aid in this understanding, we performed in situ scanning electron microscopy single crystal micropillar compression tests of a commercial Hf alloy, where the samples were aligned to activate 〈a〉 prismatic and 〈a〉 basal slip systems. We then employed transmission electron microscopy to examine the dislocation structure in the deformed pillars. These two slip systems exhibited distinctly different behaviour, where prismatic slip is planar and basal slip is wavy. Prismatic slip is the easiest deformation mode, while basal slip is at least twice as difficult. Generally, the characters of prismatic and basal slip in Hf show similarities to those in some other hexagonal metals such as Ti and Zr, except for the higher relative difficulty of basal slip. Our results provide fundamental knowledge and parameters that can be used for modelling degradation and predicting performance of Hf, and shed light on the general behaviour of basal and prismatic slip systems in low-c/a-ratio hexagonal metals.

Kinetics Parameters Estimation for the Coupled Neutron-Photon Point Kinetics Model in Monte Carlo Eigenvalue Calculations

The delayed photoneutron effect should be considered in the point kinetics equation (PKE) applied to D2O and beryllium-moderated or -reflected systems because of its importance on reactor kinetics. This paper proposes neutron-photon-coupled PKE with a clear definition of adjoint-weighted kinetics parameters after providing the mathematical derivation from the neutron and photon transport and precursor equations. The Monte Carlo algorithms to estimate the kinetics parameters are induced from their definition and implemented in the eigenvalue calculation module of McCARD. They are verified through homogeneous infinite-medium problems with few-group neutron and photon cross sections by comparing with analytic solutions. The proposed PKE and kinetics parameters estimation methods are applied for the kinetics analyses of the artificial control rod drop transient problem in the RB reactor. Photonuclear physics is also implemented in a time-dependent Monte Carlo simulation to calculate the reference data of neutron amplitude in the core to be compared with the PKE results. As good agreement is shown in the neutron amplitude, it is validated that the developed PKE is capable of predicting the reactor kinetics considering photoneutrons.

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

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

Simulation of human risk scenarios for nuclear power plant operation based on finite state machine

This paper addresses risk scenarios caused by Human-Machine Interaction (HMI) during the Nuclear Power Plant (NPP) operation. A simulation method is proposed in this research based on a Finite State Machine (FSM) for analyzing human-related risk scenarios in NPP operations. The risk evolution mechanism between the physical system and operator is modeled by FSM to clarify how risk accumulation leads to accidents. It has the advantage of managing and simplifying the modeling process of complex relationships in NPP with the integration of human reliability analysis. The study focuses on the dynamic interaction between the physical system in the primary loop and the operators during the startup of NPP and evaluates human-related safety risk through simulation of the behavior of the operators handling the control rods. This research aims to present an FSM-based methodology for analyzing HMI-induced risks in NPP operations, which can be utilized to illustrate the risk evolution, especially human-related risk.

Neutronic characteristic analysis of the GAMA microreactor

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

A comprehensive framework for identifying safety-related and ageing sensitive sscs in low power nuclear research reactor

This work aims to establish a novel approach for the selection of structures, systems, and components (SSCs) important to safety and vulnerable to ageing degradation for low-power research reactors. For this purpose, the framework developed combines the Preliminary Risk Analysis (PRA) technique to categorize SSCs into four different safety classes, and Ageing Failure Mode and Effect Analysis (AFMEA) technique to find out ageing failure modes related to a specific SSC and evaluate them according to the Risk Priory Number (RPN). The intersection of the PRA and AFMEA outcomes then yields a list of SSCs that are both important to safety and strongly ageing sensitive, making it possible to optimize the selection process of SSCs to be prioritized in further preventive maintenance, periodic testing, and inspection activities. The present paper also conducts a case study to apply the proposed methodology to the Brazilian IPR-R1 TRIGA nuclear research reactor. The selected SSCs were then both AL and SS fuel rods from the reactor core, stainless steel pipe, thermocouples, pressure indicators, and water level indicators from the primary circuit, and power monitoring channels, radiation monitors, and cooling water temperature channels from the controlling system. Moreover, the approach was also able to identify SSCs that are important from a safety perspective but are not ageing sensitive (reactor tank and control rods) and components that are not important to safety but are strongly ageing sensitive (pipe filters). These findings may further serve as input for the comprehensive safety assessment conducted during the Periodic Safety Review of the facility.

Study on safety gap between fuel channels in NHR200-II

This paper presents a novel method for estimating the safety gap between deformed fuel channels. The irradiation causes deformation of the fuel channel, which narrows the control rod channel, directly affecting the safety of the reactor. Investigating the critical safety gap between fuel channels can be utilized for safety assessments in design. Specifically, we first define the safety gap and three typical dropping cases. Next, we analyze the forces on the control rod, including flow resistance and channel resistance. Then, we establish a nonlinear dynamic model and simulate three dropping cases. Finally, we explore the influence of the friction coefficient and the fuel rod bundle. We find that the critical safety gap for the NHR200-II reactor is -5.0∼-3.1 mm (standard case) and -4.2∼-2.5 mm (arbitrary case), respectively. Our study provides a novel solution to accurately evaluate the effect of deformed boxed channels on the control rod’s dropping.

A new algorithm of critical control rod position search in MC code

Delayed Material Replacement algorithm is used in control rod movement simulation. Then, critical control rod position search is studied in Monte Carlo criticality calculation. Support Vector Regression algorithm is used in fitting neutron absorption as a function of rod positions. Fission iteration batches are reorganized as blocks in order to avoid fission source perturbation caused by control rod movement. The keff result in AP1000 reactor calculation shows discrepancy less than 20 pcm, though stable fluctuations are found in rod position values during fission iteration.

Numerical Study of Control Rod's Cross-Section Effects on the Aerodynamic Performance of Savonius Vertical Axis Wind Turbine with Various Installation Positions at Suction Side

Numerical Study of Control Rod's Cross-Section Effects on the Aerodynamic Performance of Savonius Vertical Axis Wind Turbine with Various Installation Positions at Suction Side

Optimized FOPID controller for nuclear research reactor using enhanced planet optimization algorithm

Nuclear reactor control is pivotal for the safe and efficient operation of nuclear power plants, facilitating the regulation of reactor reactivity. This study introduces an optimized fractional-order proportional-integral-derivative (FOPID) controller tailored for maintaining reactivity levels in nuclear power plants, particularly during load-following operations. The controller adjusts the position of control rod to regulate power output effectively. We enhance FOPID controller's performance using a modification of Planet Optimization Algorithm (POA-M), leveraging the strengths of the Arithmetic Optimization Algorithm (AOA) to improve its exploitation capabilities. We evaluate the efficacy of POA-M-FOPID controller against traditional techniques, including POA, AOA, and Particle Swarm Optimization (PSO). We assess its performance using the Egyptian Testing Research Reactor (ETRR-2) as a case study. Our results demonstrate that the POA-M-FOPID controller outperforms alternative algorithms across various control metrics, exhibiting superior resilience and efficiency. Notably, the utilization of the POA-M-FOPID controller yields remarkable improvements in reactor power performance, achieving significantly reduced settling time (25.27 sec) and maximum overshoot (0.67 %) compared to conventional FOPID controllers incorporating POA, AOA, and PSO methods. These findings underscore the effectiveness of POA-M-FOPID in enhancing nuclear reactor control systems, offering potential benefits for broader nuclear power industry in terms of safety, stability, and operational efficiency.

Comparing the Neutronic Behavior of LR-0 reactor During the First 20 Shakes ‎from Startup When Controlled by Boron vs. Silver Control Rods

Comparing the Neutronic Behavior of LR-0 reactor During the First 20 Shakes ‎from Startup When Controlled by Boron vs. Silver Control Rods

Fault Detection and Isolation in Simulated Batch Operation of Fine Motion Control Rod Drives

Control rods and elements manage the power distribution in nuclear reactors through the motion of banks of rods distributed throughout the core. These positional changes are achieved through the actuation of fine motion control rod drive (FMCRD) mechanisms. In the BWRX-300 design by GE-Hitachi, this mechanism is electrically driven by a servomotor that allows for high-precision control of power outputs. Under operational transients, such as load-following, accurate and precise operation of these servomotors is necessary over long periods of time, so they are key maintenance targets to maintain availability and operational flexibility. Swiftly and precisely identifying faults in the drive mechanisms will support predictive maintenance and reduced costs. This paper used three different types of simulated faults to test the fault detectability of principal component analysis (PCA) when considering the simulated operations of banks of control rods and their associated servomotors. These faults were stator short-circuit faults, ball screw jam faults, and ball screw wear faults. Torque and position were monitored in the simulation. The position signal was insufficient to detect mechanical faults. Torque signals for each servomotor in the bank of rods undergoing multiple position demand changes were projected to a reduced dimensional space via PCA. Q and T2 statistics were employed for anomaly detection. Using this approach, all faults were detected, and the anomalies were isolated to the faulty FMCRD mechanism.