Self-powered Neutron Detectors Research Articles

THERMOCOAX rhodium SPND sensitivity dispersion and validation of the sensitivity calculation model

This paper describes the neutron irradiation tests of 7 THERMOCOAX Self-Powered Neutron detectors (SPNDs) in the BR1 reactor at SCK•CEN. The SPNDs were fabricated according to the same specifications, but from different fabrication batches. The SPND signals were recorded during stepwise power-up of the reactor, proving the linearity of the SPND response within a wide thermal neutron flux range: from as low as 0.8·109 n/(cm2) to 2.6·1011 n/(cm2s). Intercomparison of the SPND signals shows a very small spread, confirming the repeatability of the THERMOCOAX fabrication processes. The overall neutron sensitivities of the seven SPNDs agreed within a 1% margin. The experimental data were analyzed in terms of prompt and various delayed responses. Prompt contributions to the signal are due to external gamma induced processes and to processes involving gamma rays emitted instantaneously upon neutron capture. The main contribution in a rhodium SPND is due to activation of the rhodium emitter and beta emission during decay of the activated rhodium and leads to a delayed response with a characteristic time of a few minutes. Activation and subsequent beta decay in other materials present in the SPND lead to additional minor delayed signal contributions. The partial SPND sensitivities due to all these processes were calculated using an MCNPX based model and were compared with experimental sensitivities based on the recorded data. The results were in fair agreement; for the overall SPND neutron sensitivity an agreement within a 1% margin was achieved.

Compton (prompt-response) neutron detectors: Comparison of emitter materials through the nuclear transmutation model

Reactors’ safety and efficiency are the determining factors in the modern nuclear power industry, and their required level can be achieved only by using the appropriate control and protection systems. The problem of the accurate and prompt control of energy release is especially important when the reactor is operated near its critical parameters.Currently, β-emission self-powered neutron detectors (SPNDs) are used to control the energy release field in the WWER-1000 core. For such SPNDs, the signal delay and the life-time are inconsistent with the tendency to optimize the reactor operating conditions (for example, with the desire to increase the burnup rate of nuclear fuel). That is why, in order to ensure the necessary safety and economy indicators, it is proposed to use inertialess (prompt-response; Compton) SPNDs in the In-core Instrumentation System. Cd, Co, Er, Hf and Pt are usually used as the neutron-sensitive elements of such detectors. Therefore, it is necessary to determine, based on certain criteria, which metal is the best suited for WWER-1000 conditions. Such studies are usually carried out in real conditions and are quite complex, expensive and require a lot of time. However, the same can be realized by a qualitative comparison, which is a simpler way. In this paper, it is proposed to compare the popular materials of Compton SPNDs through the Rutherford model of nuclear transformations and the ‘Bateman solution’. By considering the calculation results of the elemental composition, the ability to absorb neutrons, the activity, the signal level, and metal cost, the number of possible candidates for the role of the most suitable Compton SPND emitter material can be significantly narrowed down.

Remote monitoring of Molten Core-Concrete Interaction experiment with Optical Fibre Sensors & perspectives to improve nuclear safety – DISCOMS project

The DISCOMS project (Distributed Sensing for Corium Monitoring and Safety) aimed at providing innovative solutions not requiring local electrical power supplies, for remote monitoring of a severe nuclear accident. The solutions are based on both long length SPNDs (Self Powered Neutron Detectors) and on distributed OFSs (Optical Fibre Sensors) capable to detect the onset of a severe accident, the corium pouring on the containment building concrete basemat, and its interaction with the concrete floor under the reactor vessel, until it spreads in the core catcher (EPR case). This paper mainly focuses on these last three detection targets achievable with distributed OFSs. It is based on the results of a Molten Core & Concrete Interaction (MCCI) experiment, namely VULCANO, held in June 2018 with a concrete crucible equipped with overall ~ 180 m long optical fibre sensing cables. This small scale experiment (50 kg of prototypical corium) has demonstrated the ability of distributed OFSs to remotely provide useful data during the MCCI run: i) temperature profiles images up to about 580°C (single wavelength Raman DTS reflectometer) until cooling down to room temperature, ii) high spatial-resolution frequency shifts profiles, due to combined (non-selective) strain and temperature influences (Rayleigh OFDR and Brillouin reflectometers), and iii) cables lengths ablated by the corium on sections weakened by the temperature (Raman DTS, Rayleigh OFDR, telecom and photon counting reflectometers).

РАСЧЁТ ПО ПРОГРАММАМ MCU-PD И ТВС-М ПОКАЗАНИЙ ДЕТЕКТОРОВ ПРЯМОЙ ЗАРЯДКИ

For successful licensing and safe passage of pilot operation of a new type of fuel in the core of an operating VVER-1000 reactor, it is necessary to carry out a preliminary validation of design codes that are routinely used for developing fuel cycles, core safety analysis and calculating maintenance of power unit operation. Validation matrix of the design codes includes problems of calculating transition functions of self-powered neutron detector (SPND). SPNDs are the part of incore monitoring system, which allows to determine power density in the core of an operating power unit. Transition functions from the self-powered detector current to the power density are calculated by means of TVS-M spectrum code. For each type of FA and each instance of self-powered detectors, transition functions are calculated in advance. Transition functions are never corrected during operation, therefore, strict requirements are imposed to the accuracy of their calculation. The main objective of the work is a comparative analysis of the TVS-M spectrum code and the MCU-PD Monte Carlo code as applied to the calculation of transition functions of self-powered neutron detectors for VVER fuel assemblies. During the work, we created calculation models of several types of fuel assemblies, differ in type of fuel, presence of fuel rods with Gd, and placement of SPND. We performed several series of calculations of the created FAs at different burnup steps. The calculations were carried out in states with different coolant parameters, fuel temperatures and presence of control rods. As the target functionals, we considered SPND current, SPND transition function and power density factor of six fuel rods surrounding SPND. We determined maximum relative deviations of TVS-M from MCU-PD in the target functionals calculation. We confirmed acceptable accuracy of calculations of the target functionals by means of TVS-M code for wide range of reactor parameters. We also estimated the influence of homogenization of the SPND structure on the accuracy of transition function calculation.

Toward the implementation of self-powered, wireless, real-time reactor power sensing

The successful development of real-time, self-powered, wireless reactor power sensing technology would enable key operating parameters of every fuel assembly in a commercial reactor core to be continuously monitored, and potentially be utilized for applications such as closed-loop reactor control. This paper discusses the expansion of successful work on the application of Self Powered Neutron Detectors (SPND) for reactor power monitoring and the wireless transmission of these detector signals using advanced vacuum micro-electronic (VME) transmitter technology developed by Westinghouse in collaboration with the Penn State University Radiation Science and Engineering Center. Results demonstrate that this technology may be successfully applied via long-term testing in the Penn State TRIGA reactor. SPND are inherently slower than, for example, ion chambers due to their dependence on radioactive decay to generate a signal. Additional research developed delay compensation algorithms to accelerate the response of these detectors for use in near-real-time closed-loop reactor control. Initial results indicate that practical signal acceleration can be obtained digitally or with an equivalent analog circuit. Future research will apply an inverse model of the sensor’s balance differential equations to more-effectively accelerate the SPND signal. Use of inverse models will also prove to be adaptive in the sense that time-dependent burn-up of these detectors, due to neutron activation of the detector material, can be modeled and applied in the acceleration algorithm.

Determination of VVER-1000 Thermal Power Based on Background Signals of Self-Powered Neutron Detectors

The value of the reactor thermal power (RTP) is used in the VVER-1000 control systems in most algorithms for generation of control, blocking and protection signals. Besides, the technical and economic indicators of the power unit are determined by this parameter. Plans to increase VVER‑1000 RTP to 101.5%, and later to 104-107% of the nominal require additional justification of the accuracy of the RTP determination. Therefore, the task of increasing the accuracy of RTP determination is important. The paper describes the ways to improve the accuracy of weighted mean thermal power (WMTP) determination by selecting the optimal weight coefficient (that subsequently is used for WMTP determination) of each of the methods of RTP determination, namely: by thermotechnical parameters of the primary and secondary sides by neutron flux in the in-core monitoring system (ICMS) and in the neutron flux control equipment (NFCE). Another possibility of increasing the accuracy of WMTP determination, namely by increasing the number of methods of RTP determination, is also considered in the paper. The analysis of changes in the background signals of self-powered neutron detectors (SPNDs) during the VVER-1000 fuel campaigns shows the fundamental possibility of using the total background signal as a separate and independent method for RTP determination. The paper presents the results of the calculation of RTP determination error taking into account the coefficients of the components of the total RTP determination error: systematic, dynamic and random errors, which must be determined during the commissioning phase. The results of reduction of the error of WMTP determination in case of application of the additional method of RTP determination based on background signals of the SPNDs are presented. Theoretically, possible minimum values of the WMTP determination error are given depending on the values of the error of the RTP determination by separate methods.

VVER Thermal Power Measured Using the Readings of Rhodium Self-powered Neutron Detector (SPND)

A procedure for determining the linear heat generation rate in a fuel assembly of VVER reactor at the location of a rhodium self-powered neutron detector (SPND) and verification of this procedure are discussed. A method of measurements of the VVER thermal power based on SPND readings is described.

Current compensation for material consumption of cobalt self-powered neutron detector

Abstract Co Self-Powered Neutron Detector (SPND) is confronted with the problem of material consumption, which causes the response current can neither reflect the change of neutron flux in time nor be proportional to the neutron flux. In this paper, a deconvolution-based method is established to solve this problem. First of all, a step signal of neutron flux is taken as an example to analyze its performance. When the material consumption of Co SPND is 10%, after compensation, the response current can be in correspondence of neutron flux. Finally, the effects of this model in different Signal-to-Noise Ratio are analyzed, which fully confirms the truth of its excellent performance for compensating Co SPND's signal.

Development of Advanced Instrumentation for Transient Testing

A fuel safety research program centered on in-pile transient testing experiments is being developed to support assessment and qualification of advanced nuclear fuel systems using the recently restarted Transient Reactor Test (TREAT) facility at the Idaho National Laboratory. While resumption of transient testing at TREAT is crucial to enable these programs, full recovery and cutting-edge transient testing capability also require a well-coordinated and innovative instrumentation development and qualification program to support near-term and future objectives. This paper summarizes the experimental approach of transient testing to focus on measuring the response of nuclear fuel to off-normal (or power-cooling mismatch) conditions for modern and advanced reactor environments requiring capabilities extending over wide measurement and environment conditions. It also highlights unique attributes of transient testing of importance to in-pile instruments including relatively low total neutron fluence, high gamma heating, and the need for a well-defined and possibly short time response. Historical approaches to instrumentation for transient testing are also reviewed to provide context to the modern instrument strategy. The paper details the instrumentation needs of modern transient testing. It also summarizes several ongoing research and development (R&D) activities that support the development of state-of-the-art and advanced measurement technologies that will provide a baseline capability for light water reactor and sodium-cooled fast reactor (SFR) experiment objectives. This R&D will extend to other advanced reactor needs and advanced sensing technology opportunities. Examples of specific sensors planned for near-term deployment with ongoing development include prompt response self-powered neutron detectors, miniature fission chambers, optical fiber–coupled infrared pyrometers, cladding surface thermocouples, electrical impedance–based boiling detectors, and linear variable differential transformer–based sensors for fuel elongation and pressure measurement.

Uncertainty quantification of the SPND dynamic model based on non-intrusive polynomial chaos method

Self-Powered Neutron Detectors (SPNDs) are widely used in nuclear reactors to monitor the core neutron flux distribution and to provide meaningful information for regulation and protection of the reactor. In this paper we present the implementation and the application of Non-Intrusive Polynomial Chaos (NIPC) method for uncertainty analysis of the SPND dynamic model. The stochastic responses of the SPND model can be expanded as a series of orthogonal polynomial basis, and the corresponding coefficients are obtained using the regression-based approach. The variances of the SPND responses are determined using these spectral coefficients and are compared to those obtained by Monte Carlo Random Sampling (RS) method. The results show that for SPND dynamic response model NIPC method achieves the accuracy of RS method with a lower computational cost.

Numerical Simulation and Sensitivity Study of the Rhodium Self-Powered Neutron Detector Used in PWR

Focusing on the specific self-powered neutron detector (SPND) using rhodium as the emitter material, this paper performs the numerical simulation and sensitivity study for SPNDs used in a typical pressurized water reactor (PWR). The computational model was developed based on a PWR lattice code bamboo lattice and a Monte Carlo code Geant4. In order to validate the model, the neutron sensitivities with single-energy incident neutrons are calculated and compared against the experimental results. Influences of the incident neutron energy spectrum and the emitter burnup to the neutron sensitivity of the SPND are analyzed. Bamboo lattice is used to calculate the incident neutron energy spectrum and the emitter nucleus densities which are the environment of the SPND. Numerical results show that the neutron sensitivity decreases due to the increase of the enrichment, the assembly average burnup, the boron concentration, the moderator temperature, and the burnable poison rods. The increase in the emitter burnup will lead to the decrease of the neutron sensitivity. Taking the incident neutron energy spectrum and the nucleus density into account together, the neutron sensitivity decreases with the increase in the assembly average burnup.

EX-CORE NEUTRON FLUX MONITORING SYSTEM IN GRAPHITE PRISM FOR GEN. IV REACTORS

Nowadays Gen. IV reactors are attracting more attention, and so the same attention should be given to the safety and security requirements of these reactors. One of the most critical safety criteria is connected to the control of the chain reaction and thus to the neutron flux monitoring system. The neutron flux monitoring system is an integral part of the instrumentation and control system (I&C) on reactor that has to be able to provide a full monitoring range information throughout the whole operation cycle. Neutron flux can be in general monitored via ex-core or in-core instrumentation. The measurement can be conducted by fission chambers, selfpowered neutron detectors (SPND), proportional counters, ionization chambers and other detectors such as plastic organic scintillators etc. The accurate positioning of neutron flux detectors has to be determined for each type of Gen. IV reactor with respect to all technical and physical aspects. This paper is aimed on ex-core measurement of neutron flux of Gen IV graphite moderated reactors, while discussing the international standards, safety requirements, technical requirements, quality requirements etc. This research will be applied to the molten salt modular reactor. The specification of the neutron flux monitoring system together with possible positioning is discussed with respect to the modular reactor dimensions and power. In the final part, first steps towards to the ex-core measurements for Gen IV are made. In order to be able to determine the correct position of the detector, the model of the reactor, where preliminary test will be made is created in Serpent code and this model is validated.

In-Pile Qualification of a Fast-Neutron-Detection-System

In order to ensure the quality and the relevance of irradiation programs in the future Jules Horowitz Reactor (JHR), the French Alternative Energies and Atomic Energy Commission (CEA) has significantly increased its research and development effort in the field of in-pile instrumentation during the last decade. Major progress has been achieved in the capability to perform accurate in-pile measurements using reliable and updated techniques. A significant part of this effort has been conducted in the framework of the Joint Instrumentation Laboratory between the CEA and the Belgian Nuclear Research Centre (SCK $\cdot$ CEN). In order to improve measurement techniques for neutron flux assessment, a unique system for online measurement of fast neutron flux has been developed and recently qualified in-pile. The fast-neutron-detection-system (FNDS) has been designed to monitor accurately high-energy neutron flux (E > 1 MeV) in typical material testing reactor (MTR) conditions, where neutron flux levels can be as high as $10^{15}\,\,\text {N}\cdot \text {cm}^{-2}\cdot \text {s}^{-1}$ and are generally dominated by thermal neutrons. Moreover, the neutron flux is coupled with a high gamma flux of typically a few $10^{15} \gamma \cdot \text {cm}^{-2}\cdot \text {s}^{-1}$ , which can be highly disturbing for the online measurement of neutron fluxes. The patented FNDS system is based on two detectors allowing the simultaneous detection of both thermal- and fast-neutron flux. Thermal neutrons can be measured using a self-powered neutron detector or a 235U miniature fission chamber, while fast neutron detection requires a miniature fission chamber with a special fissile material presenting an energy threshold near 1 MeV, which can be 242Pu for MTR conditions. Fission chambers are operated in Campbelling mode for an efficient gamma rejection. FNDS also includes a specific software that processes measurements to compensate online the fissile material depletion and to adjust the sensitivity of the detectors, in order to produce a precise evaluation of both thermal and fast neutron flux even after long-term irradiation. FNDS has been validated through a two-step experimental program. A first set of tests was performed at BR2 reactor operated by SCK $\cdot$ CEN in Belgium. Two FNDS prototypes were operated in-pile during nearly 1000 h. These tests exhibited the consistency of the measurement of thermal to fast neutron flux ratio with Monte Carlo neutron and photon transport calculations, as well as the right compensation of fissile material depletion. This paper describes a second test recently completed at ISIS reactor operated by CEA in France. For this irradiation, FNDS signal was compared to reference thermal- and fast-neutron flux measurements using activation dosimeters analyzed under COFRAC Quality Certification. During this latter test, FNDS proved its ability to measure online the fast neutron flux with an overall accuracy better than 5%. FNDS is now operational and is assumed to be the first and unique acquisition system able to provide an online measurement of the fast neutron flux in MTR conditions. This system will be used to perform spectral neutron characterization of JHR channels, but it may also be implemented in future irradiation experiments, for a better and real-time evaluation of the fast neutron flux received by material and fuel samples.

Investigation on A-FNS neutron spectrum monitor system

We proposed the neutron spectrum measurement system for Advanced Fusion Neutron Source (A-FNS). It is configured with fission chambers, Self-Powered Neutron Detectors, neutron activation systems using water flow, and activation monitors. In this study, we mainly investigated the multiple activation foils which are sensitive to high energy neutrons. We selected 14 main dosimetry threshold reactions based on specific A-FNS usage conditions. The reactions are as follows: 55Mn(n,2n)54Mn, 59Co(n,p)59Fe, 59Co(n,2n)58Co, 59Co(n,3n)57Co, 75As(n,2n)74As, 89Y(n,2n)88Y, 93Nb(n,2n)92mNb, 169Tm(n,2n)168Tm, 169Tm(n,3n)167Tm, 197Au(n,2n)196Au, 197Au(n,3n)195Au, 209Bi(n,3n)207Bi, 209Bi(n,4n)206Bi, and 209Bi(n,5n)205Bi. The dosimetry reaction rates were calculated using Monte Carlo code MCNP-5.15 with McDeLicious-11 and nuclear data libraries FENDL-3.1c, IRDFF-1.05 and EAF-2010. As for neutron-induced cross section data of the reactions in the high energy region, we clarified that there were limited experimental data, and proposed new experiment for their evaluations.

Determination of control rod positions during fuel life-cycle using fixed in-core Self-Powered Neutron Detectors of Tehran Research Reactor

This paper deals with reconstruction of reactor states from measurements of fixed in-core Self-Powered Neutron Detectors (SPNDs) by Artificial Neural Networks (ANNs). Reactor states are defined by position of the control rods and burn-up of the fuel for Tehran Research Reactor. Position of control rods during fuel life-cycle should be monitored, because of their strong impact on neutron flux distributions and operational safety margins of the reactor. In the literature, position of control rods is reconstructed using axial distributions of neutron flux measured by in-core detectors in the vicinity of the control rods. In this research, ANNs are employed to determine position of the control rods and burn-up of the fuel, using 2D neutron flux measurements of the previously optimized SPNDs at central elevation of the reactor core. Also, sensitivity studies are performed on mapping performance of the ANNs with different architectures and training algorithms. The results indicate that a specific architecture of feed-forward neural networks, which are trained by Bayesian regularization back-propagation algorithm, has the best performance for the reactor states reconstruction. In addition, noise resistance capability of the developed ANN is evaluated with noisy data sets. The developed neural network has satisfactory noise resistance response, especially when trained by noisy data sets.

Construction of a Prototype Silver Self-powered Neutron Detector and Simulation of a Developed SPND using MCNP4C Code

Self-powered neutron detectors (SPNDs) are used widely in flux mapping, control, and protection of nuclear reactors. Compared with other neutron detector, SPND has the advantages of small size, low cost, and relatively simple electronics. In this work, a prototype SPND for flux measuring neutron generators has been designed and tested. Using the Monte Carlo simulation code, the output current of the detector was calculated and compared with the experimental results. The agreement between the results suggests that this detector is capable of flux monitoring of the neutron generator. Then, a developed SPND has been introduced with the provision of producing higher output current.

Self-Powered Neutron Detector Calibration Using a Large Vertical Irradiation Hole of HANARO

A calibration technology of the self-powered neutron detectors (SPNDs) using a large vertical irradiation hole of HANARO is developed. The 40 Rh-SPNDs are installed on the polycarbonate plastic support, and the gold wires with the same length as the effective length of the rhodium emitter of the SPND are also installed to measure the neutron flux on the SPND. They are irradiated at a low reactor power, and the SPND current is measured using the pico-ammeter. The external gamma-rays which affect the SPND current response are analyzed using the Monte Carlo simulation for various irradiation conditions in HANARO. It is confirmed that the effect of the external gamma-rays to the SPND current is dependent on the reactor characteristics, and that it is affected by materials around the detector. The current signals due to the external gamma-rays can be either positive or negative, in that the net flow of the current may be either in the same or the opposite direction as the neutron-induced current by the rhodium emitter. From the above procedure, the effective calibration methodology of multiple SPNDs using the large hole of HANARO is developed. It could be useful for the calibration experiment of the neutron detectors in the research reactors.

Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

The Self-Powered Detector Simulation ‘MATiSSe’ Toolbox applied to SPNDs for severe accident monitoring in PWRs

In the framework of the French National Research Agency program on nuclear safety and radioprotection, the ‘DIstributed Sensing for COrium Monitoring and Safety’ project aims at developing innovative instrumentation for corium monitoring in case of severe accident in a Pressurized Water nuclear Reactor. Among others, a new under-vessel instrumentation based on Self-Powered Neutron Detectors is developed using a numerical simulation toolbox, named ‘MATiSSe’. The CEA Instrumentation Sensors and Dosimetry Lab developed MATiSSe since 2010 for Self-Powered Neutron Detectors material selection and geometry design, as well as for their respective partial neutron and gamma sensitivity calculations. MATiSSe is based on a comprehensive model of neutron and gamma interactions which take place in Selfpowered neutron detector components using the MCNP6 Monte Carlo code. As member of the project consortium, the THERMOCOAX SAS Company is currently manufacturing some instrumented pole prototypes to be tested in 2017. The full severe accident monitoring equipment, including the standalone low current acquisition system, will be tested during a joined CEA-THERMOCOAX experimental campaign in some realistic irradiation conditions, in the Slovenian TRIGA Mark II research reactor.

Measurement and simulation of thermal neutron flux distribution in the RTP core

The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.