Rod Position Research Articles

Structural Design and Numerical Simulation of FDM Molding Chamber Based on Thermal-Fluid-Structure Coupling

Four different design of FDM molding chamber with hot air circulation were proposed to meet the requirements of temperature uniformity of FDM rapid prototyping equipment. Fluid and solid domain model of FDM molding chamber was established by finite numerical simulation method in order to analyze the influence of the heating rod, blower, air inlet and outlet position on the temperature uniformity and air flow. The results show that the temperature uniformity and air velocity evenness are highest when the heating rod is located on both sides of the molding chamber, the blower is lied at the bottom, the air inlet is located above the left side, and the air outlet is above the right side. And the temperature uniformity is up to 99%. FDM molding chamber prototype was developed according to the simulation results. Air flows evenly, and temperature uniformity is as high as 97.66%, which verified the accuracy and effectiveness of the simulation results deeply.

BANDUNG TRIGA 2000 REACTOR POWER ANALYSIS AS A FUNCTION OF THE NUMBER OF FUEL ELEMENTS AND THE POWER PEAKING FACTOR

The reactivity value of the Bandung TRIGA 2000 reactor core has decreased over time, so the power generated by the reactor is also getting smaller, despite the control rod position is fully withdrawn. Therefore, it is necessary to reshuffle and refuel the fuel element to increase the excess reactivity by considering the safety parameters, such as axial and radial power peaking factors, DNBR, dTsat, and temperature on the cladding and in the center of the fuel element. The analyzed reactor safety parameters are the number of fuel elements, which varied at 105, 110, and 115 elements, as well as power peaking factor, which varied at 1.55, 1.65, 1.75, 1.85, and 1.95. The calculations were done using MCNP and COOLOD-N2 programs. If DNBR ≈ 1.3 is determined as the safety limit for the operation of the Bandung TRIGA 2000 reactor, at PPF 1.95 (105, 110, and 115 fuel elements), it can be considered to operate the reactor at the power of 600-700 kW. However, at PPF of 1.75 (105, 110, and 115 fuel elements), the reactor can be operated at the power of 700-800 kW, and at PPF of 1.55 (105, 110, and 115 fuel elements), the reactor can be considered for operation at the power of 800-900 kW. The results of these calculations can be used for consideration in determining the operating limits of the Bandung TRIGA 2000 reactor.Keywords: TRIGA 2000, fuel element, power peaking factor, DNBR, boiling

Review on sensors to measure control rod position for nuclear reactor

Control rod plays a crucial role in the safety and operation of the nuclear reactor. Meanwhile, the measurement of the control rod position is conducted by the control rod position measurement sensor (CRMS), which concerns the start-up, power regulation and shut-down of the nuclear reactor. However, there still lacks research on comprehensive reviewing the state of art progress on the CRMS of the nuclear reactor. In this review, we first introduce the classifications of the CRMS, which includes the inductance type, non-inductance type, and so on. Then, we explain principles of various CRMSs, in particular, physical models of the inductance type in details. Furthermore, the experimental and numerical validation of the CRMS, which include the calibration, validation of static measurement, validation of scram performance test, numerical validation of the CRMS are illustrated. Lastly, further concerned issues of the CRMS are discussed.

An improved control rod selection algorithm for core power control at TRIGA PUSPATI Reactor

The 1 MWth TRIGA PUSPATI Reactor known as RTP undergoes more than 37 years of operation in Malaysia. The current core power control utilized Feedback Control Algorithm (FCA) and a conventional Control Rod Selection Algorithm (CRSA). However, the current power tracking performance suffers and increase the workload on Control Rod Drive Mechanism (CRDM) if the range between minimum and maximum rod worth value for each control rod has a significant difference. Thus, it is requiring much time to keep the core power stable at the power demand value within the acceptable error bands for the safety requirement of the RTP. In conventional CRSA, regardless of the rod worth value, the lowest position of the control rod is selected for up-movement to regulate the reactor power with 2% chattering error. To improve this method, a new CRSA is introduced named Single Control Absorbing Rod (SCAR). In SCAR, only one rod with highest reactivity worth value will be selected for coast tuning during transient and the lowest reactivity worth value will be selected for fine-tuning rod movement during steady-state. The simulation model of the reactor core is represented based on point kinetics model, thermal-hydraulic models and reactivity model. The conventional CRSA model included with control rod position dynamic model and actual reactivity worth curve data from RTP. The FCA controller is designed based on Proportional-Integral (PI) controller using MATLAB Simulink simulation. The core power control system is represented by the integration of a reactor core model, CRSA model and FCA controller. To manifest the effectiveness of the proposed SCAR algorithm, the results are compared to the conventional CRSA in both simulation and experimentation. Overall, the results shows that the SCAR algorithm offers generally better results than the conventional CRSA with the reduction in rising time up to 44%, workload up to 35%, settling time up to 26% and chattering error up to 18% of the nominal value.

Effects of geometry on resistance in elliptical pipe flows

This paper considers the significant role of cross-sectional geometry on resistance in co-axial pipe flows. We consider an axially flowing viscous fluid in between two long and thin elliptical coaxial cylinders, one inside the other. The outer cylinder is stationary, while the inner cylinder (rod) is free to move. The rod poses a resistance to the axial flow, while the viscous fluid poses a resistance to any motion of the rod. We show that the equations for flow in the axial direction – driven by a prescribed flux – and for flow within the cross-section of the domain – driven by the motion of the rod – decouple in the asymptotic limit of small cylinder aspect ratio into axial Poiseuille flow and transverse Stokes flow, respectively. The objective of this paper is to calculate numerically the axial and cross-sectional resistances and to determine their dependence on cross-sectional geometry – i.e. rod position and the ellipticities of the rod and bounding cylinder. We characterise axial resistance, first for three reduced parameter spaces that have not been fully analysed in the literature: (i) a circle in an ellipse, (ii) an ellipse in a circle and (iii) an ellipse in an ellipse of equal eccentricity and orientation, before extending our geometric parameter space to determine the overall optimal geometry to minimise axial flow resistance for fixed cross-sectional area. Cross-sectional resistance is characterised via coefficients in a Stokes resistance matrix and we highlight the interdependent effects of cross-sectional ellipticity and boundary interactions.

Dynamometer-Card Classification Uses Machine Learning

This article, written by JPT Technology Editor Chris Carpenter, contains highlights of paper SPE 194949, “Beam-Pump Dynamometer Card Classification Using Machine Learning,” by Sayed Ali Sharaf, Tatweer Petroleum; Patrick Bangert, SPE, Algorithmica Technologies; and Mohamed Fardan, Tatweer Petroleum, et al., prepared for the 2019 SPE Middle East Oil and Gas Show and Conference, Manama, Bahrain, 18–21 March. The paper has not been peer reviewed. In reciprocating rod pumping systems, analysis of dynamometer-card data can deliver valuable insight into the status of the pump, and can indicate if future action is required. The complete paper explains the steps taken to improve surveillance of beam pumps using dynamometer-card data and machine-learning techniques, and reviews lessons learned from executing the operator’s first artificial intelligence (AI) project. Background The oldest and most widely used method of artificial lift is called beam pumping, or the sucker-rod lift method. A dynamometer is a device used on beam pumps that measures load on the polished rod (top) and plots the load in relation to the rod position as the pumping unit moves through a stroke cycle. This plot is known as the surface card. A pump card is a plot of load vs. position on the pump’s plunger. The pump card is more useful for surveillance purposes because it filters effects of anything above the plunger and provides standard pump card shapes for interpreting pump operating conditions. Identification and diagnosis of beam pumps using the pump card is an expensive human visual-interpretation process, not only requiring significant labor time but also deep expertise in the domain. Use of machine-learning techniques for pattern recognition can help automate the visual interpretation process, increasing efficiency and reducing maintenance activities resulting from missed early diagnosis. Data Collection Sucker-rod pumps are widely deployed in the Bahrain Field. There are two different types of communication layers used across the field, radio and optical fiber. Almost 300 optical-fiber- connected beam-pump units (BPUs) have been selected for surveillance and advanced analysis. Fiber is used to avoid any communication issues during data collection for the AI project. The sampling rate of collection for each pump is one reading per 20 seconds, thus covering three pump cards per minute. All BPUs are connected to a central server that converts hardware-communication protocol used by a programmable logic controller into the open-platform-communication (OPC) protocol. Fig. 1 shows the details of the 209 values representing a single pump card being read from a BPU and stored in the operator database. During the data-collection period, a total of 5,380,163 pump cards from 297 BPUs were collected and stored in the database. Data Labeling and Preparation In machine learning, an important step after data collection is data labeling, which is usually performed manually by experts. The labeled data set is used to feed machine-learning algorithms that detect patterns specific to each class. A total of 35,292 pump cards has been labeled (more than 1,000 cards per day on average). A useful feature of the labeling software is that it allows labeling an entire time period of cards in one shot, although such labeling should be performed with care.

Validation of coupled fission matrix – TRACE methods for thermal-hydraulic and control feedback on the Penn State Breazeale Reactor

A new method for calculating power shape and critical control rod position is described and validated in view of increasing the fidelity of the Penn State Breazeale Reactor's (PSBR) fuel management code. Currently, the PSBR fuel management code is biased against experimental results and requires the use of empirical correction factors. This is likely due to a combination of assumptions, such as the control rods being fixed in the all-out position and the use of a uniform fuel temperature. The proposed method supplies this data using fission-matrix-based neutronics coupled with TRACE thermal-hydraulics. The fission-matrix-based neutronics relies on Serpent for precalculation of fission matrices. Steady-state fuel temperature distributions, power shapes, and control rod positions are calculated through iteration between TRACE and fission matrix calculations. Validation is performed on the coupled TRACE-fission matrix model for powers ranging from 50 kW to 1 MW. For these cases, the RMS average error with experimental data is 12.7 K for instrumented element fuel temperature and $0.07 for reactivity change. Agreement is generally within estimated experimental uncertainty. The effect of various modeling parameters, such as the fuel-cladding gap heat transfer coefficient and the exact radial location of the instrumented element thermocouple, are shown. Finally, the impact of this increased modeling fidelity on the depletion calculation is estimated. In the future, this validation should be repeated on later core loadings of the PSBR where more detailed experimental data is available. Comparison between the current PSBR fuel management code and the proposed model should also be made.

Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation

Abstract Since estimating the minimum departure from nucleate boiling ratio (MDNBR) requires complex calculations, an alternative method has always been considered. One of these methods is neural network. In this study, the Back Propagation Neural network (BPN) and Radial Basis Function Neural network (RBFN) are introduced and compared in order to estimate MDNBR of the VVER-1000 light water reactor. In these networks, the MDNBR were predicted with the inputs including core mass flux, core inlet temperature, pressure, reactor power level and position of the control rods. To obtain the data required to design these neural networks, an externally coupledcode was developed and its ability to estimate the thermo-hydraulic parameters of the VVER-1000 reactor was compared with other numerical solutions of this benchmark and the Final Safety Analysis Report (FSAR). After ensuring the accuracy of this coupled-code, MDNBR was calculated for 272 different conditions of reactor operating, and it was used to design BPN and RBFN. Comparison of these two neural networks revealed that when the output SMEs of the two systems were approximately the same, the training process in RBFN was much faster than in BPN and the maximum network error in RBFN was less than in BPN.

Dynamic characteristics of the capacitance control rod position measurement sensor system in NHR-200

In this paper, the dynamic mathematical model of the capacitance control rod position measurement sensor (CCRMS) were established. The experimental research, numerical calculation, and model validation were carried out for the case of the velocity dynamic response calculation, scramming dynamic response calculation, scramming time test and scramming performance measurement test of the CCRMS. The results show that the validity of the proposed model is verified and the CCRMS is basically realized the dynamic process measurement function with one step display measurement requirement. Furthermore, the dynamic characteristics of the double-helix structure CCRMS system were obtained based on the proposed model and experimental research. Moreover, the dynamic measuring error of the CCRMS is less than ±1.20 mm, the delay time of the CCRMS is t = 3τ which is less than 0.84 ms, and the dynamic response time of the capacitance measuring instrument is about 8 ms.

Experimental study on strengthening of RC beams with Side Near Surface Mounted technique-CFRP bars

This paper presents an experimental program that was carried out for the purpose of studying the global flexural performance of reinforced concrete beams strengthened internally with Carbon Fiber Reinforced Polymer (CFRP) rods using the Side Near Surface Mounted (SNSM) technique. The CFRP rods were placed laterally, adjacent to the longitudinal steel bars inside precut grooves. The strengthening length and position of CFRP rods as well as the type of filling material were the main variables investigated in this study. Moreover, a detailed comparison between the Side Near Surface Mounted (SNSM) and the Near Surface Mounted (NSM) techniques for the strengthening of RC beams using CFRP rods was conducted as well; in order to validate and assess the effectiveness of the SNSM technique.The test results showed that using the SNSM-CFRP rods technique allowed improving in a significant way the load carrying capacity of RC beams but decreased their ductility along the deflection at maximum load. The results obtained indicated that the failure mode was influenced by the length of CFRP rods and the filling material characteristics, while the strengthening position did not have significant impact. The SNSM strengthening technique can be used as an alternative method to the NSM method, and in some cases, it may be used to prevent the non-conventional failure modes, due to degradation of the NSM strengthening system, such as the pull out of CFRP-rods or premature debonding failure. The conventional analytical model accurately predicted the strength capacity and mid-span deflection of the SNSM strengthened beams.

NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR

NEUTRONIC AND THERMAL HYDRAULICS ANALYSIS OF CONTROL ROD EFFECT ON THE OPERATION SAFETY OF TRIGA 2000 REACTOR. Analysis of neutronic and thermal-hydraulics parameters of whole operation cycle is very important for the safety of reactor operation. During the reactor operation cycle, the position of the control rods will change due to reactivity changes. The purpose of this study is to determine the effect of control rods position on neutronic and thermal-hydraulics parameters in relation to the safety of reactor operation of the TRIGA 2000 reactor using silicide fuel of MTR plate type. Those parameters are power peaking factor, reactivity coefficients, and steady-state thermohydraulic parameters. Neutronic calculations are performed using a combination of WIMSD/5 and Batan-3DIFF codes and for thermal-hydraulics the calculations are done using WIMSD/5 and MTRDYN codes. The calculation results show that the reactivity coefficient values are negative for all control rod positions both at CZP and HFP conditions. The MTC value decreases when the control rod is inserted into the active core while the FTC value increases. The total ppf results and temperature in steady-state rise when the control rods are inserted of into the active core whereby the maximum value occurs at the position of the control rods of 20 cm from the bottom of the active core. The calculation results of ppf, reactivity coefficient, and thermal-hydraulics parameters lay below safety limits, indicating that the TRIGA 2000 reactor can safely use U3Si2-Al silicide fuel as a substitute fuel for cylindrical type fuel.Keywords: neutronic, thermal-hydraulic parameter, control rod effect, TRIGA 2000, silicide fuel.

EFFECT OF THE CONTROL ROD POSITION ON CRITICALITY (S CURVE) SIMULATIONS USING THE SCALE COMPUTER CODE IN THE KARTINI REACTOR

Abstract The Kartini Reactor has been operated up to 100 kilo watts. There are two types of nuclear fuel used in the Kartini Reactor, type 104 and type 204. Currently the nuclear fuel utilization license granted by BAPETEN allows 77 pieces, consisting of 73 type 104 and four type 204. Based on Kartini Reactor fuel utilization report, there are 71 pieces of nuclear fuel in the reactor core, consist of 69 type 104 and two pieces of type 204. There are four pieces of type 104 in the bulk shielding. The control rod categorizations are divided into regulating rod, shim rod and safety rod. Regarding the operation license renewal of Kartini Reactor, we conducted an assessment of various aspects related to the safety of the reactor. One of the aspects we conducted was regards to the neutronic. On the initial phase of the neutronic aspect study, a simulation was performed using the SCALE computer code. This study will only evaluate the relation between the bank of the control rod position to the effect of criticality in the Kartini Reactor. The results showed that fully up criticality has a value of 1.02173 ± 0.00018, and fully down criticality has value of 0.96026 ±0.00021. The reactor reaches criticality when the position of the control rod is at an altitude of 42.85135 cm.

Research on the capacitance control rod position measurement sensor in NHR-200

The accurate and reliable measurement of the control rod position concerns the operation and safety of the 200 MW nuclear heating reactor (NHR-200). In this paper, the static mathematical model of the capacitance control rod position measurement sensor (CCRMS) were established. The experimental research, finite element method (FEM) simulation, and model validation were carried out for the case of the ungrounded metal medium, grounded metal medium, dielectric medium control rod measured rod and eccentricity. The results show that the validity of the proposed model is verified and the CCRMS is basically realized the one-step display measurement requirement. Furthermore, the eccentricity inconsistency of the measured rod is solved. In addition, the end effect of the CCRMS can be reduced reasonably but the eccentricity of the measured rod still have some influences to a certain extent. Moreover, the static characteristics of the double-helix structure CCRMS were obtained based on the proposed model.

Determination Of First Criticality In TRIGA 2000 Reactor Modificated Using Plate Type Fuel With MCNP6

This research offers a fuel conversion from cylinder type to plate type for solving the operation problems as a solution when the TRIGA reactor fuel production run out. The research has been implemented utilizing MCNP6 software to model and simulate the core of the TRIGA 2000 reactor plate type fuel. The geometry of the simulated reactor consists of a reflector, a moderator and a reactor tank. The fuel was U3Si2Al with 19.75% enrichment and the density was 2.96 gU/cm3, the reflector was graphite and the control rod was AgInCd. The first criticality value with 15 fuel assemblies and the 4 control rods position at 11.97cm from the top surface of the reactor core was (1.00000 ± 0.00016) in agreement with the reference value for initial configuration of TRIGA reactor core was (1.00 ± 0.01).

Heat transfer in a pressurized fluidized bed with continuous addition of fines

The goal of this work was to investigate the impact of pressure and the presence of fine particles (i.e., a surrogate for pulverized fuel) on the overall surface-to-bed heat transfer coefficient in relation to an oxygen-fired pressurized fluidized bed combustor. Experiments were conducted in a pilot-scale fluidized bed with an inner diameter of 0.15 m under cold flow conditions. A tube bundle consisting of five horizontal staggered rows was completely submerged in the bed. One of the tubes was replaced by a heating cartridge housed in a hollowed copper rod. Five thermocouples distributed at 45° intervals along the copper rod circumference measured the surface temperature and ensured that local effects were included. The bed material was glass beads of 1.0 mm in diameter while the fines were glass beads of 60 μm in diameter and thus susceptible to entrainment. The fine particles were continuously fed to the fluidized bed and then captured downstream by a filter system. Fluidization was conducted at 101, 600 and 1200 kPa (abs) with excess gas velocities (Ug-- Umf) of 0.21, 0.29 and 0.51 m/s. Fine particle feed rates were 0, 9.5 and 14.4 kg/h. Two heating rod positions (2nd row and 4th row) were studied. Overall, the heat transfer coefficient approximately doubled when pressure was increased from 101 to 1200 kPa. At atmospheric conditions, where the slug flow regime occurred, the maximum heat transfer coefficient was at the bottom of the rod, while it moved to the side of the rod at high pressures where the bubbling regime occurred. When the heating rod was moved from the 2nd row to the 4th row, the averaged heat transfer coefficient increased by 18%, 9% and 6% at 101, 600 and 1200 kPa. The addition of fine particles decreased the average heat transfer coefficient by 10 to 20 W/m2 K, where the averaged coefficients were approximately 220 and 450 W/m2 at 101 and 1200 kPa respectively, but there was no effect on the angular heat transfer profile around the tube surface. The results showed that average heat transfer coefficients matched the correlation developed by Molerus et al. (1995) within a 5% difference across all conditions when fines were not present.

An iterative fission matrix scheme for calculating steady-state power and critical control rod position in a TRIGA reactor

Although Monte Carlo simulations are the most accurate representation of nuclear reactor cores, they can be computationally expensive. This computational expense can prohibit the use of Monte Carlo in situations that require sequences of neutronics calculations, such as accounting for temperature feedback. In the present work, we demonstrate a fission matrix methodology to calculate the fission source in the Penn State Breazeale Reactor while accounting for control rod movement and fuel temperature feedback. A fission matrix A, represents the rate of fission neutron production in a reactor core, where the aij entry of the matrix is equal to the neutrons produced in cell i per neutron born from cell j. The fission source distribution and multiplication factor keff are the principle eigenvector and principle eigenvalue, respectively, of the fission matrix. Given a varied core temperature distribution, a new fission matrix can be obtained by interpolating from a database of pre-calculated fission matrices. This approach shows promise in accurately and quickly calculating the reactivity and fission source distributions for a system with an arbitrary temperature distribution. In the present work, the largest resulting relative error of the calculated power distribution per pin was 1.52% compared to a Serpent criticality simulation, and all calculated eigenvalues were within 50 pcm, which corresponds to 0.1–0.4 cm of rod insertion depending on whether rods are inserted in the center or edge of the core. In contrast, the critical rod position varies by 13.49 cm for the cases considered (0–650 kW). The present database is composed of data from 78 Monte Carlo calculations, but enables subsequent neutronics calculations to be done in a fraction of a second. Future work would be to implement correction factors for neighboring temperature differences in three-dimensional models, incorporate xenon feedback, and investigate ways to decrease database size.

Quantification of system uncertainties in activation experiments at nuclear research reactors

Neutron flux measurements in reactor cores are often used for code validation studies, however reported experimental flux uncertainties remain limited in scope. This work presents both a detailed assessment of several flux measurement uncertainties and a guideline for reporting these quantities. The effects of the sample position, the presence of 135Xe, the control rod positions, reactor power, and nuclear data during activation procedures were individually quantified at the McMaster Nuclear Reactor. NiCr flux wires were used to study the 51Cr (n,γ) activation reaction experimentally and through a series of MCNP6 simulations, and the SCALE-6.2.2 252-group library was used to study cross-section data uncertainty for use in translating activation measurements into neutron flux values. The placement of the sample and the reactor power measurements were found to be the most significant contributors to the overall measurement uncertainty. A procedure for reporting the results and uncertainties of an irradiation procedure is provided.

A prototype for SANDD: A highly-segmented pulse-shape-sensitive plastic scintillator detector incorporating silicon photomultiplier arrays

We report the first clear observation of neutron/gamma-ray pulse-shape sensitivity of a fully-instrumented 8 $\times$ 8 array of plastic scintillator segments coupled to two 5 cm $\times$ 5 cm 64-channel SiPM arrays as part of a study of the key metrics of a prototype antineutrino detector module designed for directional sensitivity. SANDD (a Segmented AntiNeutrino Directional Detector) will eventually comprise a central module of 64 elongated segments of $^{6}$Li-doped pulse-shape-sensitive scintillator rods, each with a square cross section of 5.4 mm $\times$ 5.4 mm, surrounded by larger cross section bars of the same material. The most important metrics with the potential to impact the performance of the central module of SANDD are neutron and gamma-ray pulse-shape sensitivity using silicon photomultipliers (SiPMs), particle identification via scintillator rod multiplicity, and energy and position resolution. As a first step, we constructed a prototype detector to investigate the performance of a central SANDD-like module using two 64-channel SiPM arrays and rods of undoped pulse-shape-sensitive plastic scintillator.

Estimation of fuel temperature reactivity coefficient of a research reactor using ANFIS

Fuel Temperature reactivity coefficient is one of the important operational and safety parameters in nuclear reactors to monitor for normal operation and transient safety analysis in research as well as in power reactors. The aim of this work is to estimate the temperature reactivity coefficient by using adaptive neuro-fuzzy inference system (ANFIS) as a preliminary assessment for developing an online monitoring system in research reactors such as the Reactor TRIGA PUSPATI (RTP). The input for the learning mechanism is the fuel temperature recorded at Ch A of the instrumented fuel element (IFE) and control rod positions with the fuel temperature reactivity coefficient as the targeted output. The ANFIS model was validated using testing data that consisted of fuel temperature measured at Ch B and results on -0.00571 $°C−1 (-3.997 pcm°C−1) of actual output and -0.00592 $°C−1 (4.144 pcm°C−1) of predicted output. Through the developed model, it showed that the ANFIS also can be used as a tool to estimate the fuel temperature reactivity coefficient with good prediction and accuracy shown from the calculation of mean absolute error (MAE) of 0.00349.

Fractal fluctuations in exploratory movements predict differences in dynamic touch capabilities between children with Attention-Deficit Hyperactivity Disorder and typical development.

Children with Attention-Deficit Hyperactivity Disorder (ADHD) struggle to perform a host of daily activities. Many of these involve forceful interaction with objects and thus implicate dynamic touch. Therefore, deficits in dynamic touch could underlie functional difficulties presented by ADHD children. We investigated whether performance on a dynamic touch task (length perception by wielding) differ between children with ADHD and age-matched controls. We further examined whether this difference could be explained by fractal temporal correlations (wielding dynamics). Forty-two children (ADHD: 21; typically developing: 21) wielded unseen wooden rods and reported their perceived length in the form of magnitude productions. The rods varied in the magnitude of the first principal moment of inertia (I1). Three-dimensional displacements of hand and rod positions were submitted to Detrended Fluctuation Analysis to estimate trial-by-trial temporal correlations. Children with ADHD reported shorter length for rods with higher I1 than their typically developing peers, indicative of reduced sensitivity to mechanical information supporting dynamic touch. Importantly, temporal correlations in wielding dynamics moderated children’s usage of I1. This finding points to a role of exploratory movements in perceptual deficits presented by children with ADHD and, thus, should be considered a new potential target for interventions.