Control Rod Position Research Articles

Mechanism model for effect of eccentricity on capacitive control rod position indicators

The continuous identification technology for position signals based on the capacitive method has a good application potential in the field of the control rods position measurement in nuclear reactors. The main source for error of the Capacitive Control Rod Position Indicators (CCRPIs) comes from the effect of eccentricity. The conformal mapping technique was used in this paper, and the geometry of the CCRPIs could be transformed. A model for revealing the mechanism of the effect of eccentricity was established, and it was tested and verified by the method of numerical calculation. The characteristics of the distribution of the sensitive field was clarified by the model. The results indicate that the maximum error of the model is smaller than 13 %, and the eccentric effect of the indicator can be described by the model.

Development and experimental validation of conceptual models for solenoid-based in-vessel two-channel control rod position indicator

In this study, we proposed in-vessel two-channel control rod positioner models for small nuclear reactors and carried out performance tests to evaluate applicability of the models. Embedding control rod positioners in reactors reduces the volume of small reactors and minimizes the penetration parts of reactor. In this study, two models having two physically separated solenoids were developed for measuring the position of the control rod by using the inductance change of a solenoid. First, the principle of measuring the position of a ferromagnetic rod by using a solenoid was formulated, and the change in solenoid inductance with the rod position was calculated through finite element analysis. Based on the proposed models, test products were manufactured and tested on three items to confirm the feasibility of the proposed solenoid-based two-channel control rod position indicator.

Numerical comprehensive evaluation of the flow control effect on a circular cylinder with a control rod

The drag reduction of the single circular cylinder is achieved by changing the position of an additional control rod in the present work. In order to look for where to place the control rod will help the drag force exerted on the main cylinder surface as little as possible, the additional control rod is arranged at various positions in the downstream region, the upstream region, and the middle region of the single circular cylinder, respectively. The numerical results show that the maximum drag reduction rate of the single circular cylinder can be up to 21.68%. While the additional control rod is located at some specific positions in the flow field, the drag of the single circular cylinder will increase. However, if two bluff bodies (the main circular cylinder and the control rod) are considered as a whole system, due to the extra drag contribution of the control rod, the maximum drag reduction effect of the system is 8.65%. Additionally, the lift exerted on the main cylinder and the control rod has changed a lot due to the interaction between two bluff bodies. Furthermore, the Dynamic Mode Decomposition analysis method is employed to probe the mechanism of drag variation employing the dominant flow modes.

Reinforcement Learning-Based Control Sequence Optimization for Advanced Reactors

The last decade has seen the development and application of data-driven methods taking off in nuclear engineering research, aiming to improve the safety and reliability of nuclear power. This work focuses on developing a reinforcement learning-based control sequence optimization framework for advanced nuclear systems, which not only aims to enhance flexible operations, promoting the economics of advanced nuclear technology, but also prioritizing safety during normal operation. At its core, the framework allows the sequence of operational actions to be learned and optimized by an agent to facilitate smooth transitions between the modes of operations (i.e., load-following), while ensuring that all safety significant system parameters remain within their respective limits. To generate dynamic system responses, facilitate control strategy development, and demonstrate the effectiveness of the framework, a simulation environment of a pebble-bed high-temperature gas-cooled reactor was utilized. The soft actor-critic algorithm was adopted to train a reinforcement learning agent, which can generate control sequences to maneuver plant power output in the range between 100% and 50% of the nameplate power through sufficient training. It was shown in the performance validation that the agent successfully generated control actions that maintained electrical output within a tight tolerance of 0.5% from the demand while satisfying all safety constraints. During the mode transition, the agent can maintain the reactor outlet temperature within ±1.5 °C and steam pressure within 0.1 MPa of their setpoints, respectively, by dynamically adjusting control rod positions, control valve openings, and pump speeds. The results demonstrate the effectiveness of the optimization framework and the feasibility of reinforcement learning in designing control strategies for advanced reactor systems.

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.

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.

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.

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.

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%.

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.

A new monitoring method for control rods position of small modular reactors (SMRs)-based coaxial capacitor

The position of control rods is one of the core parameters that can affect the power of nuclear reactors, and it is needed to be monitored by the position indicators. Based on the central shielding effect of the driving rod, the design of the coaxial capacitive position indicators is proposed in this paper. The theoretical rod position recognition model of the indicator is constructed by the conformal mapping, and it is verified by the numerical method. The results manifest that the performance of the indicator can be accurately analyzed by the theoretical model. The performance of the indicator is studied, and it is compared with that of the existing capacitive position indicator. The results show that compared with the existing design, the precision and the sensitivity of the new indicator can be enhanced by two orders of magnitude at the same time.

Research on Neutronics Safety Parameters of the AP1000 Nuclear Reactor under Different Conditions

Changes in temperature during reactor operation may cause changes in physical parameters, leading to core overheating and accidents. It is essential to analyze and assess the safety parameters of the core under different operating conditions. This paper investigates the effects of fuel temperature, moderator density, boron concentration, and control rod state on AP1000 safety parameters. The study uses RMC and NJOY to calculate the changes in reactivity factor, effective delayed neutron fraction, and neutron generation time of the AP1000 reactor under different operating conditions. The changes in reactivity coefficients, neutron fluxes, and relative power densities of AP1000 reactors are analyzed for normal and accidental operating conditions. The results indicated that the reactivity coefficient remained negative under accident conditions, which ensured the safe operation of the reactor. The delayed neutron fraction, neutron flux, and power density distributions are affected by fuel temperature, moderator density, and control rod position. The control rod worth was sufficient for the emergency shutdown of the reactor under accidental conditions. It is demonstrated that the operation of the AP1000 reactor under study conditions is safe and controllable.

Multivariate Image Analysis for Core Monitoring in PWRs

In pressurized water reactor (PWR) it is crucial for the operator to monitor the reactor parameters at the same time, such as temperature, pressure, boron concentration, control rod position, coolant density, etc., in order to make proper decision. However, the huge size of data reading from the different instrumentations, in addition to the limited human ability to visually detect, interpret, assess add a lot of uncertainty to the operator qualitative and quantitative analysis of the reactor performance. Therefore, this paper proposes the utilization of radial thermal flux maps (Neutron Images) technique, positioned on the reactor core as a sensitive monitoring technique for all changes of the reactor parameters as a result of the position of the control rods changing. Hence, the features contained in these neutron images are extracted (Multivariate image analysis and regression) via Principal Component Analysis (PCA), and Cluster Analysis (Dendrogram). To determine the effectiveness of the suggested technique in determining the location of the control-rods, several simulations are run. The 3D TRITON FORTRAN-code was utilized to simulate the radial thermal neutron flux of the Westinghouse 2775-MWth PWR benchmark at 100% thermal power generation. The SIMCA software programme is used to develop, test, and generalise the PCA model. Additionally, clustering analysis (CA) is carried out using the statistics software programme Minitab in order to demonstrate the effectiveness of the suggested method.

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.

Coupled 3D neutronics/thermal-hydraulics analysis of Superphénix start-up tests with DYN3D/ATHLET code system

The capabilities of the DYN3D/ATHLET coupled code system were recently extended to allow for 3D neutron kinetics/thermal hydraulics analyses of Sodium cooled Fast Reactors at reactor system level. In this paper, the DYN3D/ATHLET coupled code system was applied to analyze the Superphenix reactor transient benchmark. The benchmark comprises six operational transients initiated during the SPX reactor start-up tests, covering a wide range of reactor operating states. The peculiarity of these transients is the necessity to consider thermal expansions of major structural elements of the primary system, which significantly influence the transient position of control rods in the core. The paper includes a brief summary of the benchmark specifications, description of the neutron kinetics and thermal hydraulics models, an analysis of the simulation results, and a comparison to the experimental data.

Neutronic Parameter Analysis of Plate-Type Fueled TRIGA 2000 Reactor by MCNPX

A novel simulation to calculate the neutronic parameters of the TRIGA 2000 reactor using plate-type fuel has been performed. The plate fuel used was produced by the Indonesian Nuclear Industry (PT INUKI) with U3Si2-Al material. Neutronic parameters based on INUKI’s plate-type fuel dimension and the current TRIGA’s configuration were simulated using MCNPX. The simulation was performed by modeling the complete reactor’s configuration on a fresh fuel core state. We obtained the kinetic parameter values from the simulation, i.e., delayed neutron fraction of 8.11×10‑3, a prompt neutron lifetime of 2.0551×10‑4 s, and an average neutron generation time of 1.87×10‑4 s. The excess reactivity of the reactor was 9.02 %Δk/k, while reactivity in the one-stuck-rod state was below ‑0.5 $ with an average value of ‑3.40 %Δk/k (‑4.19 $). The average thermal neutron flux peak occurred at the central irradiation position with the value of 3.0×1013 to 3.1×1013 n/(cm2 s). The reactor has a power peaking factor of 1.379 in the control rod position of 0 % on D3 fuel. The reactor had a negative feedback reactivity coefficient, except for the moderator coefficient. These results suggest that the current configuration of plate-type fuel met the nuclear reactor neutronic safety standards.

A safety enhanced sodium-cooled MOX fueled fast reactor core concept

ABSTRACT We have developed a 750MWe sodium-cooled MOX fueled fast reactor core with enhanced safety against typical ATWS. Design goals were set as follows: 1) negative void reactivity for the loss of flow (ULOF); and 2) burnup reactivity less than 1 $ for the transient over power (UTOP). To achieve the goals the following were considered: 1) adoption of the axially heterogeneous core with a sodium plenum and a gas expansion module to reduce void reactivity; and 2) addition of minor actinides (MAs) to the internal blanket to reduce burnup reactivity. Configurations of the core were optimized as follows. First, void reactivity was reduced by a parameter survey for the inner core height, B-10 content of the upper shield, and standby position of the backup control rods. Second, burnup reactivity was reduced by a parameter survey for the MA content in the internal blanket, etc. We designed the safety enhanced fast reactor core that did not undergo sodium boiling if the ULOF occurred and we maintained fuel integrity for the UTOP by the transient analysis. By installing a self-actuated shutdown system in the backup control rods, the improved operational reliability of this shutdown system and the improved safety can be expected.

Control rods’ position history effect on the Indonesian multipurpose research reactor RSG-GAS operation and safety parameters

The present work extends the analysis of the first working core of RSG-GAS to the burnup calculation. During the fuel management calculations of RSG-GAS, the control rods are usually assumed to be withdrawn from the core. Due to the exclusion of the control rod, it is hypothesized that the integral control rod worth shows significant differences compared to the experimental data. Therefore, a burnup calculation was performed in this study for the first core of RSG-GAS to evaluate the impact of operating control rods. The analyses were performed using continuous-energy Monte Carlo code, Serpent 2, and ENDF-B/VIII.0 nuclear data library for both cases with and without the inclusion of control rods. As a result, it is found that the calculated control rod worth, as well as the excess reactivity at the beginning of the cycle of the RSG-GAS second core from the burnup calculation considering the control rod position, agreed with the experimental data.

Control rod driven long-term load-follow operation in small modular reactor

The need for extensive load following with Small Modular Reactor (SMR) is increasing as its potential coupling with intermittent renewable energy sources increases. SMR is generally considered suitable for load following, though, its load maneuvering capability depends on the system design features and reactor control algorithm. In this study, it is shown that a control rod driving algorithm based on the core outlet temperature of a typical IPWR type of SMR, SMART100 facilitates the extensive load following operations (IAEA, 2018, Song et al., 2010, Lee et al., 2012). Addition of control parameter such as Axial Offset (AO) and Boron concentration change into the control algorithm makes a reactor control possible when the AO is close to the limit or control rod position is close to Power Dependent Insertion Limit(PDIL). Simulations for various scenarios were performed to verify the above algorithm Short-term simulations less than 2 days and Long-term simulations longer than 7 days were performed. For each simulation, random demand or scheduled demand considering demand changes incurred by renewable energy was used. In all simulations, it is confirmed that the reactor facilitates control rod driven load-following operation within the operation limit. Further, it is shown that more efficient load-following operation is achieved by additional adjustment of control band of outlet temperature.

Investigation of the Effect of Rod Worth Uncertainty on the Reactivity Measurements of CEFR Start-Up Tests via McCARD Monte Carlo Calculations

In the reactivity measurements of the China Experimental Fast Reactor (CEFR) start-up tests, five independent control rods were moved to compensate for the reactivity insertion. Since the recorded control rod positions were converted to reactivity via S-curves (rod worth curves), any uncertainty in the S-curves can propagate to all reactivity measurements. In this work, we rigorously derived the uncertainty of the reactivity in terms of the statistical uncertainty of the S-curves with Monte Carlo (MC) simulations. Additionally, the average error of the control rod worth from the MC calculation was estimated from experiments and embedded into the uncertainty formulation. The formulation shows that the error of the reactivity is highly correlated to the changes in the control rod position during the reactivity measurements. McCARD MC simulations were then conducted for the CEFR start-up tests, and the calculated reactivity and uncertainty were compared with the measurements. The main error factor of each reactivity calculation was figured out by quantifying the uncertainty components. With the uncertainty formulation, the calculation results showed a better agreement compared with the measurements, as the relative errors were observed mostly within 2σ of the uncertainty.