Self-powered Neutron Detectors Research Articles

Study on structural and material dependence of Rhodium Self-powered Neutron Detector performance

The neutron measurement properties of Rh-SPNDs are greatly affected by their geometric constructions, which needs to be optimized while designing. This effect is reflected in the sensitivities for gamma and neutron detection. However, the influence on gamma sensitivity has not been fully analyzed before, which is of the same importance as neutron sensitivity. Therefore, it is necessary to quantitatively analyze the relationship between detector sensitivity and geometric dimensions. The effect of Rh-SPND structural and material parameters on neutron and gamma sensitivities has been calculated and analyzed through numerical simulations. The research results show that detector sensitivities are strongly depending on the detector size and its emitter shape. Moreover, the detector response to exogenous reactor gamma can be eliminated by optimizing the detector collector size.

A simulation study of the ability to detect power distribution perturbations in the texas A&M TRIGA reactor with self-powered neutron detectors

Given the variety of ways that nuclear reactor core power may be perturbed, reactor operators and developers are keen on understanding the accuracy and convergence time during which perturbations in reactor power distribution may be synthesized (i.e., inferred) from an array of in-core radiation detectors. A simulation study was conducted as described herein using a highly detailed model of the Texas A&M Training, Research, Isotopes, General Atomics Reactor, in which an array of self-powered neutron detectors (SPNDs) was considered for input to the power synthesis methodology. The core power synthesis is conducted using a point-based iterative method with an iterative loop built in to ensure working equation consistency. The forward problem of SPND response to simulated perturbations in reactor power was solved for Gaussian peak-type perturbations in the reactor power distribution. These perturbations varied in variance, amplitude, and core location to assess their impact on synthesis error and to determine the number of iterations required for convergence. A relation between the unique resolvability limit and perturbation width was identified such that the maximum synthesis error increased rapidly when the peak width went beneath this limit (a width approximating half the reactor’s fuel pin-to-pin pitch); this resolvability limit is specific to the SPND configuration and fuel segmentation considered herein. The synthesis error increased linearly with perturbation peak amplitude, whereas the convergence time increased nonlinearly. Perturbations located closer to the center of the core were synthesized more accurately, albeit with a higher number of required iterations. These findings provide a qualitative and quantitative understanding of the accuracy and speed at which different types of spatial power perturbations can be resolved in light-water reactors.

Applicability to DEMO breeding blankets of neutron measurements techniques from fission reactors

Neutron measurements are routinely performed in fission reactors. A lot of experience has been gained at CEA in designing, prototyping and testing neutron sensors and associated acquisition systems in several nuclear facilities with a wide range of demanding conditions. In this paper, the applicability of this experience to neutron diagnostics in the breeding blankets of DEMO is addressed, as a complement to the already planned diagnostic strategy.The breeding blankets are a key component of a fusion reactor such as DEMO. The expected performances and operational constraints are demanding, nevertheless some of those, as the neutron flux range and high temperature, are close to what is encountered in sodium-cooled fast neutron reactors.The strengths and drawbacks of neutron sensors are presented with respect to the performances and constraints. Self-powered neutron detectors and fission chambers benefit from a vast return of experiment in fission reactors and received recent developments in the framework of studies for sodium-cooled fast neutron reactors. The innovative optical fission chamber is discussed for its electromagnetic immunity and high dynamic range demonstrated up to seven decades. Dosimetry with activation foils is also considered for its ability to provide an absolute calibration and spectral unfolding.

An Improved Numerical Model for Quantitatively Computing the Space Charge Effect of Self-Powered Neutron Detector

An Improved Numerical Model for Quantitatively Computing the Space Charge Effect of Self-Powered Neutron Detector

Simulation study of a rhodium self-powered neutron detector for irradiation of nuclear fuel and material reactor

To accurately judge the irradiation test cycle of nuclear fuel and material in-pile irradiation, and provide precise design inputs for the post-irradiation performance evaluation of fuel and material, the NIT-Rh self-powered neutron detector has been designed and developed. The sensitivity characteristics of the detector have been simulated and verified. This study introduces the basic structure and working principle of the rhodium self-powered neutron detector for irradiation of nuclear fuel and materials in the reactor. Additionally, a new computational model has been constructed by the Monte Carlo method, and the current sensitivity of different sizes of emitters and insulators has been analyzed. The model considers a more realistic neutron source and β-electron energy spectrum, accounts for the space charge effect on the signal in the insulator, and compares the simulated sensitivity with the reactor test sensitivity. The results show that the new rhodium self-powered neutron detector calculation model is more accurate, and the relative deviation between the simulated and reactor test sensitivities is only 3.06 %. It can be used for the in-heap irradiation test and provides a method for the application of the irradiation of nuclear fuel and material reactors.

Fault Identification Method for In-Core Self-Powered Neutron Detectors Combing Graph Convolutional Network and Stacking Ensemble Learning

Fault Identification Method for In-Core Self-Powered Neutron Detectors Combing Graph Convolutional Network and Stacking Ensemble Learning

A Monte Carlo method for quantitatively calculating the neutron sensitivity of rhodium self-powered neutron detectors in reactors

A Monte Carlo method for quantitatively calculating the neutron sensitivity of rhodium self-powered neutron detectors in reactors

Study of neutron sensitivity for vanadium self-powered neutron detector in nuclear reactors

Vanadium self-powered neutron detectors (SPNDs) are widely used as in-core unmovable detectors in nuclear power plants to continuously carry out data monitoring of neutron flux and nuclear power. In this study, a theoretical calculation model is presented for determining the signal current and thermal neutron sensitivity of vanadium SPNDs. First, the physical characteristics and reaction mechanisms in vanadium SPNDs irradiated with neutrons were discussed thoroughly. Then, a computational approach to determine the neutron sensitivity of the vanadium SPND based on a Monte Carlo method without assumptions and approximations was introduced. Finally, in order to verify the performance of the vanadium SPND and our methodology, the calculated values were compared with experimental measurements and the previous literature. It can be found that all the results agree well with the experimental measurements and that the maximum differences do not exceed 3%, which demonstrates the reliability and effectiveness of the method and further improves the calculation accuracy.

Monte Carlo analyses and experimental investigation of the vanadium self-powered neutron detector in 60Co source and research pulsed reactor

As the online monitoring apparatus in the reactor, self-powered neutron detectors (SPNDs) have a great potential for development in nuclear energy. To further characterize the SPND response in the actual radiation field, some irradiation experiments and Monte Carlo analyses based on the 60Co gamma-rays source and a research pulsed reactor were performed. During the experiments, the current signals were measured with various reactor irradiation facilities. A series of experimental phenomena, such as the negative current effect of the gamma rays, shutdown attenuation current, and the difference in detector sensitivity, are described and theoretically explained by using a numerical model GEANT4-SPND, an in-house Monte Carlo toolkit for SPND simulation based on Geant4. The numerical model proves to be highly effective in restoring the experimental results for the detector response current, the signal prompt proportion, and the current trend. This work has not only validated the existing calculation method but also provided reference radiation field parameters and detector experimental data for the future development and verification of sensor models. Besides, the analyses results have provided some insights into the study of the mechanism and optimization of detector design.

Analysis of signal cable noise currents in nuclear reactors under high neutron flux irradiation

Cables are indispensable in nuclear power plants for transmitting data measured by various types of detectors, such as self-powered neutron detectors (SPNDs). These cables will generate disturbing signals that must be accurately distinguished and eliminated. Given that the cable current is not very significant, previous research has focused on SPND, with little attention paid to cable evaluation and validation. This paper specifically focuses on the quantitative analysis of cables and proposes a theoretical model to predict cable noise. In this model, the reaction characteristics between irradiated neutrons and cables were discussed thoroughly. Based on the Monte Carlo method, a comprehensive simulation approach of neutron sensitivity was introduced and long-term irradiation experiments in a heavy water reactor (HWR) were designed to verify this model. The theoretical results of this method agree quite well with the experimental measurements, proving that the model is reliable and exhibits excellent accuracy. The experimental data also show that the cable current accounts for approximately 0.2% of the total current at the initial moment, but as the detector gradually depletes, it will contribute more than 2%, making it a non-negligible proportion of the total signal current.

Development of a complete simulation package for Self-Powered Neutron Detectors using the Monte Carlo Method and Geant4 toolkit

In order to provide guidance on the design and use of the Self-Powered Neutron Detector (SPND) and improve measurement accuracy, accurate simulation of SPND is required. Therefore, a complete SPND simulation package was developed, including a particle transport calculation module, an electric field calculation module, a visualization interface and a data post-processing module. Due to electric field produced by space charge in the insulator, Geant4 has been chosen as the particle transport calculation module and an iterative code for the electric field was written to implement the calculation of the electric field in the insulator. To help researchers simplify the process of writing codes, a wizard-like visual interface was developed based on QT for configuring the input files and for performing the calculations. The accuracy of the simulation package was verified by comparing it with other calculation codes. With the help of the simulation package, the user can very easily achieve an accurate simulation of the SPND.

Simulation and Analysis of the Space Charge Effect in a Self-Powered Neutron Detector

Simulation and Analysis of the Space Charge Effect in a Self-Powered Neutron Detector

Twin model-based fault detection and tolerance approach for in-core self-powered neutron detectors

Twin model-based fault detection and tolerance approach for in-core self-powered neutron detectors

A general method for delay compensation of self-powered neutron detectors in reactors

Self-Powered Neutron Detectors (SPNDs) have long been widely used as reactor in-core neutron detectors. However, delayed type SPNDs, a typical type SPND with delayed current component, can not be directly used for real-time neutron flux monitoring. To solve this problem, several researchers have established compensation methods to reconstruct neutron flux. For the various SPNDs with delayed currents, the existing compensation methods are usually case-dependent. Therefore, a general compensation method for delayed type SPNDs with high portability is proposed in this paper. Based on linear equations in the s domain, this method is applicable for all the delayed SPNDs, giving the general compensation solution with definite iterations. Moreover, this compensation method is tested by two typical delayed SPNDs, indicating its high portability and applicability.

Валідація MCNP-моделі формування струму фонової жили детектора прямого заряду у ВВЕР-1000

The article presents the results of the numerical simulation of the signal formation process of the background wires of self-powered neutron detectors (SPND) under the action of gamma radiation in the VVER-1000 core using MCNP code. The validation of the MCNP model was carried out on the results of experimental determination of the current of the background wires of the SPND, obtained at three different power units with VVER-1000 during the fuel campaign. The article also proposes a new gamma-ray method for determining the thermal power of the VVER-1000 reactor (TPR) based on the signals from the background wires of the SPND. TPR is an important safety parameter of VVER-1000, therefore, increasing the accuracy of determining TPR with the introduction of an additional gamma method for its determination is an urgent task, given the plans to increase the TPR of VVER-1000. The results of the experimental determination of the VVER-1000 TPR by the traditional neutron method based on the SPND signals are presented, and problematic issues regarding the error in determining the TPR by the neutron method are pointed out. The article presents the results of modeling to study the influence of the main factors affecting the change in the proportionality coefficient Kgm between the actual TPR and the TPR determined by the gamma method. To improve the accuracy of determining the TPR by the gamma method, a correction model for Kgm is proposed, which takes into account the effect of nuclear fuel burnup on the change in the signal of the background wires of the SPND. Taking into account that the signal of the background wires of the SPND is inertialess with respect to the change in the neutron power of the reactor, the introduction of the method for determining the TPR by the gamma method is promising for the implementation of an additional alternative channel for generating an emergency protection signal in terms of both power and the period of the reactor.

CONDITION MONITORING OF SELF-POWERED NEUTRON DETECTORS IN THE INCORE NUCLEAR INSTRUMENTATION SYSTEM UNDER TRANSIENTS

The paper describes a new method for condition monitoring of self-powered neutron detectors (SPND) used to monitor neutron flux in the reactor core. The method measures currents in the SPND measuring circuit under static and dynamic conditions. A new diagnostic indicator of SPND condition is introduced allowing for significant enlargement of the detector operating range and change in the method that is used to monitor the value of SPND emitter burnout. The introduction of the additional diagnostic indicator, namely the measuring circuit capacitance, can be used for assessing the neutron flux in the reactor core under transient conditions.

Sorted compressive sensing for reconstruction of failed in-core detector signals

Self-Powered Neutron Detectors(SPNDs) are used to calculate core power distributions, an essential factor in the safe operation of nuclear power plants. Some detectors may fail during normal operation, and signals from failed detectors are isolated from intact signals. The calculated detailed power distribution accuracy depends on the number of available detector signals. Failed detectors decrease the operating margin by enlarging the power distribution measurement error. Therefore, a thorough reconstruction of the failed detector signals is critical. This note suggests a compressive sensing based methodology that rationally reconstructs the readings of failed detectors. The methodology significantly improves reconstruction accuracy by sorting signals and removing high-frequency components from conventional compressive sensing methodology.

A Coupled Deterministic and Monte-Carlo Method for Modeling and Simulation of Self-Powered Neutron Detector

To improve the efficiency of numerical calculation of self-powered neutron detector (SPND) current in pressurized water reactors (PWRs), a coupling method based on high-fidelity neutronics code and an SPND direct response model is proposed. On the detector scale, a Monte-Carlo analysis toolkit, GEANT4-SPND, is developed and verified. The toolkit, which calculates the effective contribution of insulator materials to current, has been proven to be conducive to the calculation of gamma sensitivity. On the reactor scale, detectors modeled using the deterministic method provide information of the nuclear density for the depletion calculation. The particle flux information based on the energy and spatial angular distribution is regarded as the medium of coupling strategy, thus enabling an efficient simulation of SPND current for PWR problems. This work has confirmed the necessity of particle anisotropy for current calculation in real reactor environments. The numerical code and its coupling method are verified based on the experimental values of rhodium, vanadium, platinum, and hafnium SPNDs and the K1 assembly benchmark as well. Moreover, the method is used to calculate a real VVER-1000 reactor operating in China. The numerical results are in good agreement with the values measured in realistic circumstances. What is more, the spatial distribution of the neutron sensitivity of PWR is characterized.

Simulating self-powered neutron detector responses to infer burnup-induced power distribution perturbations in next-generation light water reactors

Understanding how 3D power distribution will be monitored throughout reactor core volumetric space in next-generation nuclear power reactors is crucial to the design, deployment, and licensing of these reactors. Although numerous techniques exist for 3D power distribution monitoring based on the response of both in situ and ex situ sensors currently implemented or proposed for use in the US reactor fleet, crucial details about these techniques are often unclear. The publicly available documentation does not include information such as how well these techniques are characterized and optimized in their implementations and the levels of uncertainty in the inferred 3D power distribution. The work described herein investigated a recently developed 3D power distribution inferencing method as applied to two next-generation reactor simulations: (1) the NuScale small modular reactor design and (2) the Westinghouse AP1000 design, both of which contain in-core strings of vanadium self-powered neutron detectors (SPNDs). This investigation considered a range of SPND string sensor densities, as well as a range of 3D power distribution axial segment sizes. SPND response simulation is informed by neutron flux calculations in representative homogenized cores. For the different sensor densities and power distribution axial segment sizes in these simulations, the average solution error, solver iterations, and run time were tracked to parameterize the sensor-core configuration.

Numerical simulation of Delayed-SPND response current in NECP-Bamboo

Self-Powered Neutron Detectors (SPNDs) are widely used in Gen-III commercial PWR cores to simultaneously and continuously measure the in-core neutron flux density. A response current simulation module has been designed, developed and validated in the NECP-Bamboo code for delayed-SPNDs. First, neutron transport and nuclide depletion simulation are carried out in Bamboo-Lattice to obtain the variation of nuclide composition and the few-group constants. Next, a Monte-Carlo code is employed to calculate the electron escape probability and gamma sensitivity of SPND for each depleted lattice. Finally, the electron escape probability, gamma sensitivity and the few-group constants are regarded as input parameters for the core simulation. The response current is calculated in the Bamboo-Core code during the simulation of core operation. The methodology and the code were verified against measured values on Vanadium and Rhodium SPND. In addition, the response current of Vanadium SPND in the AP1000 core was simulated and validated.