Self-powered Neutron Detectors Research Articles

Preliminary validation of computational model for neutron flux prediction of Thai Research Reactor (TRR-1/M1)

This study is a part of an on-going work to develop a computational model of Thai Research Reactor (TRR-1/M1) which is capable of accurately predicting the neutron flux level and spectrum. The computational model was created by MCNPX program and the CT (Central Thimble) in-core irradiation facility was selected as the location for validation. The comparison was performed with the typical flux measurement method routinely practiced at TRR-1/M1, that is, the foil activation technique. In this technique, gold foil is irradiated for a certain period of time and the activity of the irradiated target is measured to derive the thermal neutron flux. Additionally, the flux measurement with SPND (self-powered neutron detector) was also performed for comparison. The thermal neutron flux from the MCNPX simulation was found to be 1.79×1013 neutron/cm2s while that from the foil activation measurement was 4.68×1013 neutron/cm2s. On the other hand, the thermal neutron flux from the measurement using SPND was 2.47×1013 neutron/cm2s. An assessment of the differences among the three methods was done. The difference of the MCNPX with the foil activation technique was found to be 67.8% and the difference of the MCNPX with the SPND was found to be 27.8%.

Clustering of Self Powered Neutron Detectors: Combining Prompt and Slow Dynamics

The focus of this work is on clustering Self Powered Neutron Detectors (SPNDs) with different dynamic characteristics into smaller groups, with each group containing highly correlated SPNDs. In order to cluster the SPNDs correctly, we propose novel ways to compensate for the effect of different dynamic response characteristics. In particular, two types of SPNDs: (i) cobalt, which give a prompt response, and (ii) vanadium, which give a delayed response, are considered. We propose and compare three compensation methods to cluster both types of SPNDs together using their measurement data: (i) pure delay applied to cobalt SPND data, (ii) cobalt SPND data `slowed down' to match the vanadium SPND dynamics by passing it through vanadium SPND transfer function, and (iii) vanadium SPND data `speeded up' to match the cobalt SPND dynamics by passing it through the inverse of vanadium SPND transfer function. Based on extensive simulations, it is found that slowing down cobalt SPND measurements to match the vanadium SPND dynamics yields the best results. This method is then used to obtain clusters from data obtained from a nuclear reactor in India for both vanadium and cobalt SPNDs and the resulting clusters appear reasonable.

New calculational model for self-powered neutron detector based on Monte Carlo simulation

A new calculational model of detector sensitivity, which is defined as the ratio of electric current to neutron flux, for self-powered neutron detector (SPND) is presented in this paper. Since Warren developed a sensitivity calculation model based on a multi-step analytic approach in 1972, there have been a lot of researches aimed at improving the accuracy of his model by partially employing Monte Carlo (MC) simulations in some of the calculation steps. However, there still exist assumptions such as mono-energy neutron in those models which could increase the uncertainty of the calculated detector sensitivity. In this paper, a new model which overcomes those limitations is developed by adopting the MC method more thoroughly and thereby reducing the number of calculation steps. The verification of the new method has been performed by comparing the results against the ones from the existing methods and measured data. A detailed analysis of SPND depletion by ORIGEN-S and BETA-S codes has shown that the proposed model can predict accurately the measured depletion data in a commercial pressurized water reactor (PWR) plant.

Robust filtering for dynamic compensation of self-powered neutron detectors

Self-powered neutron detectors (SPNDs), which are widely used in nuclear reactors to obtain core neutron flux distribution, are accurate at steady state but respond slowly to changes in neutron flux. Dynamic compensation methods are required to improve the response speed of the SPNDs and make it possible to apply the SPNDs for core monitoring and surveillance. In this paper, three digital dynamic compensation methods are proposed. All the three methods are based on the convex optimization framework using linear matrix inequalities (LMIs). The simulation results show that all three methods can provide a reasonable balance between response speed and noise suppression.

Dynamic compensation of Vanadium self powered neutron detectors based on Luenberger form filter

Vanadium Self-Powered Neutron Detectors (SPNDs) are widely used in a nuclear reactor for flux measurement, but the slow response characteristics of Vanadium SPNDs preclude their direct use for reactor protection and regulation applications. In this paper, three different dynamic compensators based on optimal filtering schemes are designed to improve the response time of Vanadium SPNDs using the dynamic mathematical model of the SPNDs. All these compensator design problems are formulated in a convex optimization framework using linear matrix inequalities (LMIs). The simulation results show that three compensators all can provide a reasonable balance between response speed and noise suppression.

Instrumentation for Neutron Flux and Tritium Production Rate Monitoring in the European TBM in ITER

Test blanket modules (TBMs) will be installed in the ITER with the aim to investigate the nuclear performance of different breeding blanket designs for fusion power reactors. Here, we present an overview of our ongoing work on three types of neutron flux monitors under development at Karlsruhe Institute of Technology (KIT) for the TBMs. A foil activation system with pneumatic sample transport was constructed by KIT utilizing the intense DT neutron generator of Technical University of Dresden. It is used as a test bench for the development of a neutron activation system for the TBMs. Self-powered neutron detectors (SPNDs) are widely applied in fission reactor monitoring, and the commercially available SPNDs are sensitive to thermal neutrons. We are investigating novel materials for SPND which would be sensitive also to the fast neutron flux expected in the TBMs. Within the I_SMART project, funded by KIC InnoEnergy, KIT is developing an online detector based on silicon carbide electronics for the TBMs.

In-Core Flux Sensor Evaluations at the ATR Critical Facility

As part of an Idaho State University (ISU)–led Advanced Test Reactor (ATR) National Scientific User Facility (NSUF) collaborative project that includes Idaho National Laboratory (INL) and the French Alternative Energies and Atomic Energy Commission (CEA), flux detector evaluations were completed to compare their accuracy, response time, and long-duration performance. Special fixturing, developed by INL, allows real-time flux detectors to be inserted into various Advanced Test Reactor Critical Facility (ATRC) core positions to perform lobe power measurements, axial flux profile measurements, and detector cross-calibrations. Detectors initially evaluated in this program included miniature fission chambers, specialized self-powered neutron detectors (SPNDs), and specially developed commercial SPNDs. Results from this program provide important insights related to flux detector accuracy and resolution for subsequent ATR and CEA experiments and yield new flux data required for benchmarking models in the ATR Life Extension Program (LEP) Modeling Update Project.

Engineering factors of the macrocode MOBY-DICK

Abstract In this paper a process of determination of methodological component of power distribution engineering factors of the macrocode MOBY-DICK (MD) is presented. This process is based on a direct comparison of measured and calculated data. The directly measured data analyzed in this article are thermocouple (TC) and Rhodium self-powered neutron detector (SPND) readings. Also the normality of deviations between measured and calculated TC temperature rises is discussed and methodological uncertainties of the macrocode MD are then checked by 97.5th percentile. Absolute and relative uncertainties of volumetric (KV) and radial (Kq) power distributions are determined by the comparisons of power distributions calculated by MD code with either power distributions provided by the Dukovany NPP core monitoring system (SCORPIO) or with power distributions obtained from direct TC and SPND readings. Final determination of engineering factors supposes hyperbolic dependence of MD methodological uncertainties on power peaking.

Comparison of Thermal Neutron Flux Measured by <formula formulatype="inline"><tex Notation="TeX">$^{235}{\rm U}$</tex> </formula> Fission Chamber and Rhodium Self-Powered Neutron Detector in MTR

Thermal neutron flux is one of the most important nuclear parameter to be measured online in material testing reactors (MTRs). In particular two types of sensors with different physical operating principles are commonly used: self-powered neutron detectors (SPND) and fission chambers with coating. This work aims to compare on one hand the thermal neutron flux evaluation given by these two types of sensors and on the other hand to compare these evaluations with activation dosimeter measurements, which are considered as the reference for absolute neutron flux assessment. This study was conducted in an irradiation experiment, called CARMEN-1, performed during 2012 in OSIRIS reactor (CEA Saclay-France). The CARMEN-1 experiment aims to improve the neutron and photon flux and nuclear heating measurements in MTRs. In this paper we focus on the thermal neutron flux measurements performed in CARMEN-1 experiment. The use of fission chambers to measure the absolute thermal neutron flux in MTRs is not very usual. An innovative calibration method for fission chambers operated in Campbell mode has been developed at the CEA Cadarache (France) and tested for the first time in the CARMEN-1 experiment. The results of these measurements are discussed, with the objective to measure with the best accuracy the thermal neutron flux in the future Jules Horowitz Reactor.

Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum.This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14MeV neutrons at the Frascati neutron generator (FNG).The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra.

Influence of the gamma-ray fraction in the reactor radiation on the total signal of a self-powered neutron detector

For controlling the linear power density in the reactor core, the Khortitsa-M software program as a part of the in-core instrumentation system (ICIS) employs only self-powered neutron detector (SPND) data with the neutronic calculation for the consistent determination of the power density in unmeasurable fuel assemblies (FAs). The confidence of the interpretation of the SPND data essentially determines the safe and efficient operation of a reactor. Previously, it was assumed that the gamma-ray fraction in the reactor radiation does not exceed one percent and is independent of the fuel enrichment and the FA and SPND burnups. Since it is difficult to estimate the contribution of the reactor gamma radiation to the SPND current experimentally, in this work, we present a calculated estimate using modern software and libraries of constants. On the basis of the results of this study, the question is discussed whether it is appropriate to take into account the reactor gamma radiation in the transfer function from the SPND current to the power density of six fuel elements surrounding the SPND with allowance for both the type of FA and the FA and SPND burnups.

Allowance for the burnup of a rhodium self-powered neutron detector in a VVER-1000 reactor

Methods for determining the dependence of the burnup of a rhodium self-powered neutron detector (SPND) from the charge flow in it are described.

Analysis of vanadium Self Powered Neutron Detector’s signal

Self Powered Neutron Detectors (SPNDs) have been widely used for monitoring the neutron flux, either in Nuclear Power Plants (NPPs) as well as in irradiation facilities and medical treatments. Nevertheless, the physical meaning of the parameter used to relate the detector signal (an electric current) to the neutron flux, i.e., the detector sensitivity, was not sufficiently analysed. Since the sensitivity, defined as ∊=iϕ, is calculated for given reactor conditions, thermal neutron flux at room temperature, it does not take into account the deviation originated at higher temperature conditions, like are found in nuclear reactors at NPPs, for example. The aim of this work is to explicitly exhibit the nonproportionality between the detector current signal and the neutron flux at different temperatures. The calculation of the neutron flux weighted microscopic cross section, definedσˆ=∫dεσϕ∫dεϕ,for a temperature interval has revealed the nonconstant behaviour of the sensitivity as a function of the temperature, showing the problem that SPNDs have to represent the neutron flux of a nuclear reactor at a NPP. As a consequence of this fact, it is not possible to use the sensibility as an absolute constant, but rather it must be calculated for each nuclear reactor and corrected under determined conditions.

Hot spot identification by sensitivity analysis and probabilistic inference methods: Demonstration exercise

The method for identifying the presence of hot spots in the core of a nuclear reactor described in a previous note (Gandini, 2011) is applied in a demonstration exercise relevant to a simplified, medium size PWR reactor. By this method the information obtained on-line through a system of neutron measuring devices such as self-powered neutron detectors (SPNDs, or collectrons) inserted in the core of a nuclear power reactor allows the on-line quick detection of a possible hot spot during plant operation. The method is based on the generalized perturbation method (GPT) techniques for the calculation of the sensitivity coefficients of the integral quantities measured with the collectrons with respect to parameters representative of the hot spot, and on the use of probabilistic inference techniques, taking into account the errors associated with the measurements. The methodology allows to assess the effect on the quality of the hot spot detection system following possible failures of the measuring devices during the core life cycle. Such an assessment may be useful for defining an adequate protection strategy in terms of quality, number and distribution of the collectrons. The results obtained with the exercise demonstrate the validity of the proposed method.

Measurement and Detection of Radiation.

Introduction to Radiation Measurements What Is Meant by Radiation? Statistical Nature of Radiation Emission The Errors and Accuracy and Precision of Measurements Types of Errors Nuclear Instrumentation Statistical Errors of Radiation Counting Introduction Definition of Probability Basic Probability Theorems Probability Distributions and Random Variables Location Indexes (Mode, Median, Mean) Dispersion Indexes, Variance, and Standard Deviation Covariance and Correlation The Binomial Distribution The Poisson Distribution The Normal (Gaussian) Distribution The Lorentzian Distribution The Standard, Probable, and Other Errors The Arithmetic Mean and Its Standard Error Confidence Limits Propagation of Errors Goodness of Data-x2 Criterion-Rejection of Data The Statistical Error of Radiation Measurements The Standard Error of Counting Rates Methods of Error Reduction Minimum Detectable Activity Counter Dead-Time Correction and Measurement of Dead Time Review of Atomic and Nuclear Physics Introduction Elements of Relativistic Kinematics Atoms Nuclei Nuclear Binding Energy Nuclear Energy Levels Energetics of Nuclear Decays The Radioactive Decay Law Nuclear Reactions Fission Energy Loss and Penetration of Radiation through Matter Introduction Mechanisms of Charged-Particle Energy Loss Stopping Power Due to Ionization and Excitation Energy Loss Due to Bremsstrahlung Emission Calculation of dE/dx for a Compound or Mixture Range of Charged Particles Stopping Power and Range of Heavy Ions (Z > 2, A > 4) Interactions of Photons with Matter Interactions of Neutrons with Matter Gas-Filled Detectors Introduction Relationship between High Voltage and Charge Collected Different Types of Gas-Filled Detectors Ionization Chambers Proportional Counters Geiger-Muller Counters Gas-Flow Counters Rate Meters General Comments about Construction of Gas-Filled Detectors Scintillation Detectors Introduction Inorganic (Crystal) Scintillators Organic Scintillators Gaseous Scintillators The Relationship between Pulse Height and Energy and Type of Incident Particle The Photomultiplier Tube Assembly of a Scintillation Counter and the Role of Light Pipes Dead Time of Scintillation Counters Sources of Background in a Scintillation Counter The Phoswich Detector Semiconductor Detectors Introduction Electrical Classification of Solids Semiconductors The p-n Junction The Different Types of Semiconductor Detectors Radiation Damage to Semiconductor Detectors Relative and Absolute Measurements Introduction Geometry Effects Source Effects Detector Effects Relationship between Counting Rate and Source Strength Introduction to Spectroscopy Introduction Definition of Energy Spectra Measurement of an Integral Spectrum with a Single-Channel Analyzer (SCA) Measurement of a Differential Spectrum with an SCA The Relationship between Pulse-Height Distribution and Energy Spectrum Energy Resolution of a Detection System Determination of the Energy Resolution-The Response Function The Importance of Good Energy Resolution Brief Description of a Multichannel Analyzer (MCA) Calibration of an MCA Electronics Introduction Resistance, Capacitance, Inductance, and Impedance A Differentiating Circuit An Integrating Circuit Delay Lines Pulse Shaping Timing Coincidence-Anticoincidence Measurements Pulse-Shape Discrimination Preamplifiers Amplifiers Analog-to-Digital Converters (ADC) Multiparameter Analyzers Data Analysis Methods Introduction Curve Fitting Interpolation Schemes Least-Squares Fitting Folding and Unfolding Data Smoothing Photon (Gamma-Ray and X-Ray) Spectroscopy Introduction Modes of Energy Deposition in the Detector Efficiency of X-Ray and Gamma-Ray Detectors: Definitions Detection of Photons with NaI(Tl) Scintillation Counters Detection of Gammas with Ge Detectors CdTe and HgI2 Detectors as Gamma Spectrometers Detection of X-Rays with a Si(Li) Detector Charged-Particle Spectroscopy Introduction Energy Straggling Electron Spectroscopy Alpha, Proton, Deuteron, and Triton Spectroscopy Heavy-Ion (Z > 2) Spectroscopy The Time-of-Flight Spectrometer Detector Telescopes (E dE/dx Detectors) Position-Sensitive Detectors Neutron Detection and Spectroscopy Introduction Neutron Detection by (n, Charged Particle) Reaction Fission Chambers Neutron Detection by Foil Activation Measurement of a Neutron Energy Spectrum by Proton Recoil Detection of Fast Neutrons Using Threshold Activation Reactions Neutron Energy Measurement with a Crystal Spectrometer The Time-of-Flight Method Compensated Ion Chambers Self-Powered Neutron Detectors (SPND) Concluding Remarks Activation Analysis Introduction Selection of the Optimum Nuclear Reaction Preparation of the Sample for Irradiation Sources of Radiation Irradiation of the Sample Counting of the Sample Analysis of the Results Sensitivity of Activation Analysis Interference Reactions Advantages and Disadvantages of the Activation Analysis Method Prompt Gamma Activation Analysis Neutron Depth Profile Neutron Radiography Health Physics Fundamentals Introduction Units of Exposure and Absorbed Dose The Relative Biological Effectiveness-The Dose Equivalent Dosimetry for Radiation External to the Body Dosimetry for Radiation Inside the Body Internal Dose Time Dependence-Biological Half-Life Biological Effects of Radiation Radiation Protection Guides and Exposure Limits Health Physics Instruments Proper Use of Radiation Applications of Radiation Detection Introduction Health Physics within Nuclear Power Plants and Radiological Facilities Portal Monitors and Passive Detection Interactive Radiation Detection Systems Unmanned Aerial Vehicles for Radiation Detection Coincidence and Anti-Coincidence Detection Systems Nuclear Medicine Detection of Nuclear Materials/Nonproliferation Issues Appendix A: Useful Constants and Conversion Factors Appendix B: Atomic Masses and Other Properties of Isotopes Appendix C: Alpha, Beta, and Gamma Sources Commonly Used Appendix D: Tables of Photon Attenuation Coefficients Appendix E: Table of Buildup Factor Constants Index Problems, Bibliography, and References appear at the end of each chapter.

Data reconciliation and gross error analysis of self powered neutron detectors: comparison of PCA and IPCA based models

Self powered neutron detectors (SPNDs) are used for measuring neutron flux in a nuclear reactor and hence are essential for safe operation of the reactor. The main objective in this paper is to develop models on representative groups of SPNDs using principal component analysis (PCA) based techniques. In particular, we identify models using regular PCA and the iterative PCA (IPCA) technique proposed in Narasimhan and Shah (Control Eng Pract 16:146–155, 2008). These models are then used for performing data reconciliation, gross error detection, identification and estimation for the SPNDs. Based on these performances, we compare the PCA and IPCA models. The proposed models can be used in the reactor to ensure that faulty detectors can be detected quickly and the corresponding values estimated in real-time thereby ensuring continuous reactor operation.

Burnup of rhodium SPND in VVER-1000: Method for determination of linear energy release by SPND readings

A method for determination of linear energy release of a VVER fuel assembly near a rhodium self-powered neutron detector (SPND) is described. The dependence of SPND burnup on the charge passing through it is specified.

Rhodium self‐powered neutron detector as a suitable on‐line thermal neutron flux monitor in BNCT treatments

A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Local mixed-field thermal neutron sensitivities and global thermal and mixed-field thermal neutron sensitivities derived from measurements performed at the RA-6 were compared and no significant differences were found. Global RA-6-based thermal neutron sensitivity showed agreement with pure thermal neutron sensitivity measurements performed in the RA-3 spectrum. Additionally, the detector response proved nearly unchanged by differences in neutron spectra from real (RA-6 BNCT beam) and ideal (considered for calibration calculations at RA-3) neutron source descriptions. The results confirm that the special design of the Rh SPND can be considered as having a pure thermal response for neutron spectra with epithermal-to-thermal flux ratios up to 12%. In addition, the linear response of the detector to thermal flux allows the use of a mixed-field thermal neutron sensitivity of 1.95 ± 0.05 × 10(-21) A n(-1)[middle dot]cm² [middle dot]s. This sensitivity can be used in spectra with up to 21% epithermal-to-thermal flux ratio without significant error due to epithermal neutron and gamma induced effects. The values of the measured fluxes in clinical applications had discrepancies with calculated results that were in the range of -25% to +30%, which shows the importance of a local on-line independent measurement as part of a treatment planning quality control system. The usefulness of the CNEA Rh SPND for the on-line local measurement of thermal neutron flux on BNCT patients has been demonstrated based on an appropriate neutron spectra calibration and clinical applications.

Prediction of Power Signal in Nuclear Reactors with Neural Network Based Intelligent Predictors in the Presence of 1/fα Type Sensor Noise

This paper describes the prediction of the power output from the Self-Powered-Neutron-Detector (SPND) in the liquid zone control compartment (LZC) of nuclear reactors, which are important in online global power measurement in a large Pressurized Heavy Water Reactor (PHWR). Noisy measured data from the SPNDs have been smoothened out with the help of an ARMA filter and then the smoothed data is used as the input for the neural networks for training purpose. These typical intelligent predictors have been studied with its variation considering different dynamic neural network structures with integer and fractional order noise considerations in the measured sensor data. The paper reports the best found network structure for the prediction of measured set of noisy SPND data.

Hot point detection method

Abstract A method is described by which the information obtained on-line through a system of neutron measuring devices such as self-powered neutron detectors (SPNDs, also called collectrons) inserted in the core of a nuclear power reactor (in particular, a PWR) allows the on-line detection of a possible hot spot during plant operation. The method is based on the generalized perturbation method (GPT) techniques, for the calculation of the sensitivity coefficients of the integral quantities measured with the collectrons with respect to parameters representative of the hot spot, and on the use of statistical inference techniques, taking into account the errors associated with the measurements. The methodology allows to assess the effect on the quality of the hot point detection system following possible failures of the measuring devices during the core life cycle. Such an assessment may be useful for defining an adequate protection strategy in terms of quality, number and distribution of the collectrons.