Self-powered Neutron Detectors Research Articles

Numerical optimization of Rhodium Self-Powered Neutron Detector

Self-Powered Neutron Detector (SPND) has been widely used in reactors to monitor neutron flux due to its adaptability for in-core severe environment. As a typical representative, SPND with Rhodium as its emitter has been modeled in the frame of Geant4, which is good at tracing the history of the particles in simulation and very appropriate for directly simulating SPND’s working mechanism. First of all, simulation model has been validated by simulating one existing typical SPND. Then, the performances’ dependences on the detector’s dimensions have been studied systematically. Finally, SPND’s geometries have been optimized and the suggested optimal design gives a neutron sensitivity of 1.03 × 10−19 A·cm2·s and a ratio of prompt current of 80.22%. Additionally, this work is also helpful to optimize all the SPNDs for customized requirements.

Deconvolution-based real-time neutron flux reconstruction for Self-Powered Neutron Detector

Self-Powered Neutron Detector (SPND) is useful for in-core neutron flux measurement in nuclear reactors due to its tiny size, simple structure, ruggedness and self-powered feature. One type of SPNDs with delayed current from instable intermediate nuclides cannot directly represent the real-time in-core neutron flux Φ(t) by their current I(t), which should be avoided during reactor control and protection. In this paper, we proposed a deconvolution-based method to reconstruct real-time neutron flux for SPND. Following the establishment of dynamic model, the unit-impulse response function h(t) was easily obtained when neutron flux was unit-impulse. Then, the iterative compensation relations were established for delay compensation according to the convolution relationship I(t) = Φ(t) * h(t). In the meanwhile, determination methods for initial values were also proposed and the compensation performance for jump neutron flux was demonstrated to be only 0.3 s. Furthermore, the dependences on initial conditions and sampling time interval were studied systematically, indicating our method is effective and robust. Finally, our method has been compared with a typical compensation method and validated with measured current, showing its advantages. This method is very attractive due to its obvious simplicity, high intuitiveness and general applicability.

Vanadium, rhodium, silver and cobalt self-powered neutron detector calculations by RAST-K v2.0

Vanadium, rhodium, silver and cobalt Self-Powered Neutron Detectors (SPNDs) are evaluated and compared in this study, including their characteristic response time. The module for response calculations of different SPNDs has been programmed and added to RAST-K v2.0 (Reactor Analysis code for Steady-state and Transient). For simplification purposes and running time minimization, previously studies have often adopted the simplified depletion chain of the SPNDs. The isomers are often neglected in the detectors studies because of their low levels of contribution to the final signal. Besides, few published documents can be found for comparisons of the behavior of different emitters in normal power changes, especially for silver and cobalt emitters. In this paper, the full depletion chains for different emitters are adopted, and the isomers’ contributions are calculated. A new method for detector response calculation based on Chebyshev Rational Approximation Method (CRAM) is used as the depletion method. It is shown that the response of silver SPND is very similar to that of rhodium SPND and can be considered as an alternate emitter material for light water reactors. This study can contribute to the selection of proper SPNDs in specific applications.

Application of Prompt Self-Powered Neutron Detectors to the Lead-Cooled Fast Reactor Demonstrator ALFRED: Validation of the Monte Carlo Model for Selected SPNDs

The advanced lead fast reactor European demonstrator (ALFRED) is a European research initiative into the framework of the Generation IV International Forum facilities. ALFRED is a scaled down reactor compared to the industrial prototype European lead fast reactor proposed in lead-cooled European advanced demonstration reactor. It has a relatively low power (125 MWe) with a compact design to reduce the cost but maintaining its representativeness and it is cooled by pure lead. One of the open issues is linked to the neutron flux in-core monitoring system because of the harshness of the environment the detectors should be installed in, due to high temperatures, and the neutron-gamma radiation field levels. Monte Carlo simulation is a possible way of facing the problem, reproducing into a virtual world the reactor core, the surrounding environment and radiation interactions. In previous works, neutron spectra and gamma doses at possible detectors' locations in ALFRED were retrieved, with consideration on the applicability of each suitable device currently available. Fission chambers (FCs) were found to be exploited at reactor start-up and intermediate power range. Prompt self-powered neutron detectors (SPNDs) seemed to be the best solution to monitor the reactor full power, becoming the main research target: their effective applicability on field has to be demonstrated. SPND applications do not include reactor control purposes usually. Moreover, their irradiation experience involved thermal and epithermal neutron spectra monitoring, mainly. The lack of data when SPNDs sense fast neutron fluxes in terms of prompt-response pushed the authors to deepen the study in such direction. The work herein shows the mathematical approach based on Monte Carlo simulation of SPNDs by the Monte Carlo N-particle eXtended code (MCNPX), so as to study the capability of the code in reproducing real devices' signals while experimented on field. Such a verification turned out to be the preliminary stage for studying new concepts for SPNDs, in terms of sensitive materials and geometries, envisaging the possibility for designing, prototyping, and testing new devices in suitable fast neutron-flux facilities.

A proposal for an alternative use of prompt-Self Powered Neutron Detectors: Online spectral-deconvolution for monitoring high-intensity neutron flux in LFRs

Self-Powered Neutron Detectors (SPNDs) are currently used in reactors’ environment to sense the magnitude of neutron-fluxes, usually for spatial-distribution mapping of the fuel region as to optimize burn-up strategies. During the development of tailored instrumentation for Lead-cooled Fast Reactors, the possibility to perform online spectral deconvolution of fast neutron-fluxes was recognized. Seven geometrically similar SPNDs with different neutron-sensitive materials, have been characterized by the Monte Carlo transport code MCNPX. Thanks to a database of spectral sensitivities vs. neutrons’ energy, a mathematical deconvolution process from 7 electric current values virtually measured by SPNDs was started, retrieving spectral information, in terms of 7 energy windows, of the neutron-flux the detectors are subjected by. This paper describes the procedure that led to those results, prefiguring future development to improve proposed method.

Application of data reconciliation for fault detection and isolation of in-core self-powered neutron detectors using iterative principal component test

In-core neutron detectors are used in large nuclear reactors for the purpose of acquiring knowledge about the spatial flux distribution, which helps in control, monitoring and protection of the reactors. Random errors and faults, which cause degradations of the signals from the true values, invalidate the in-core detector measurement data. A scheme utilizing the analytical redundancy among physically redundant in-core detectors greatly facilitates removing the ill-effects of these errors. Particularly, data-based or process-history-based schemes are suitable when there are a large number of variables as in the case of in-core detectors. In this paper, application of a Data Reconciliation (DR) scheme (with the aim of minimizing the effect of random errors) coupled with a Fault Detection and Isolation (FDI) technique (which makes the data free from faults) to the in-core detector measurement data of the Advanced Heavy Water Reactor (AHWR) is presented. This reactor uses Vanadium Self-Powered Neutron Detectors (VSPNDs) as in-core neutron detectors. Since the constraint model developed for all the in-core neutron detectors as a whole cannot yield satisfactory performance from the DR and FDI, they are developed for the individual clusters of VSPNDs using the Iterative Principal Component Analysis (IPCA), a multi-variate statistical technique. The FDI technique used is the Iterative Principal Component Test (IPCT) discussed in Sagar et al. (2015c). The effectiveness of this scheme is tested as follows: (i) the mathematical model of the AHWR is used to simulate measurement data from VSPNDs; (ii) the data from VSPNDs are grouped as clusters based on correlations among the variables; and (iii) the effectiveness of the DR-based FDI of VSPNDs in some clusters is established based on some performance indices, when the spatial control of the reactor is active.

Lifetime extension of in-core self-powered neutron detector using new emitter materials

Summary In this paper, new in-core self-powered neutron detector emitter candidate materials, ie, vanadium, cobalt, and silver, have been examined in their lifetimes compared with the commonly used rhodium emitter. Using a new quantitative lifetime evaluation model, the lifetimes of vanadium and cobalt were determined to be longer than that of rhodium, but these materials were also shown to have the disadvantage of low signal intensities. Under normal operating conditions, we showed that rhodium emitter can be used for 2 cycles of pressurized water reactors (PWRs) with lifetime of 4.35 years, whereas silver can be used for 5 cycles of PWRs with lifetime of 8.04 years. Three sensitivity tests were performed for rhodium and silver about (1) the emitter size, (2) the fuel assembly burnup, and (3) the emitter temperature variations. From the test results, we observed that the lifetimes of rhodium and silver emitters remained 2 and 5 cycles long, respectively. We concluded that silver can significantly extend the in-core detector's lifetime in PWR operation.

A Simple Iterative Method for Compensating the Response Delay of a Self-Powered Neutron Detector

The self-powered neutron detector (SPND) is popularly used as an in-core neutron flux monitor in reactors due to its simple structure, self-powered feature. However, its response delay needs compensation to obtain the actual real-time neutron flux for reactor control and protection. In this paper, a simple iterative method for compensating SPND response delay is proposed as well as noise filtering. Two favorable noise filtering methods were compared, and then, the moving average filter was chosen. The governing differential equations were established according to decay mechanism, and then, iterative compensation relations for delay compensation were established by discretization with simplicity and flexibility. The test result shows that the compensated response delay for a prompt jump of neutron flux is only 0.9 s, indicating its effectiveness. Furthermore, the dependence on initial conditions and sampling time interval was also studied, indicating that two initial condition determination modes for two typical detector start-up situations can be chosen correspondingly for delay minimization and critical sampling with a time interval of about , which shows consistency with the Von Neumann stability analysis. Finally, our method has been compared with the Z-transform method and verified with measured current, which showed its better performance.

Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction.In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

Neutron-gamma flux and dose calculations for feasibility study of DISCOMS instrumentation in case of severe accident in a GEN 3 reactor

The present paper presents the study carried out in the frame of the DISCOMS project, which stands for “DIstributed Sensing for COrium Monitoring and Safety”. This study concerns the calculation of the neutron and gamma radiations received by the considered instrumentation during the normal reactor operation as well as in case of a severe accident for the EPR reactor, outside the reactor pressure vessel and in the containment basemat. This paper summarizes the methods and hypotheses used for the particle transport simulation outside the vessel during normal reactor operation. The results of the simulations are then presented including the responses for distributed Optical Fiber Sensors (OFS), such as the gamma dose and the fast neutron fluence, and for Self Powered Neutron Detectors (SPNDs), namely the neutron and gamma spectra. Same responses are also evaluated for severe accident situations in order to design the SPNDs being sensitive to the both types of received neutron-gamma radiation. By contrast, fibers, involved as transducers in distributed OFS have to resist to the total radiation gamma dose and neutron fluence received during normal operation and the severe accident.

Analysis and solution of current spike occurred in dynamic compensation of self-powered neutron detectors

Dynamic compensation methods are required to improve the response speed of the Self-Powered Neutron Detectors (SPNDs) and make it possible to apply the SPNDs for core monitoring and surveillance. During the experimental test of the compensation method based on linear matrix inequality (LMI), spikes are observed in the compensated SPND current. After analyzing the measurement data, the cause is fixed on the unphysical change of the uncompensated SPND current due to range switching. Then some modifications on the dynamic compensation algorithms are proposed to solve the current spike problem.

On Capabilities and Limitations of Current Fast Neutron Flux Monitoring Instrumentation for the Demo LFR ALFRED

Among GEN IV projects for future nuclear power plants, lead-cooled fast reactors (LFRs) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant safety, and waste management. The novelty of this matter causes some open issues about coolant chemical aspects, structural aspects, monitoring instrumentation, etc. Particularly, hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e., source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal hydraulics and chemical conditions for the LFR core environment will be assumed, as the neutron flux will be studied extensively by the Monte Carlo transport code MCNPX (Monte Carlo N-Particles X-version). The core coolant’s high temperature drastically reduces the candidate instrumentation because only some kinds of fission chambers and self-powered neutron detectors can be operated in such an environment. This work aims at evaluating the capabilities of the available instrumentation (usually designed and tailored for sodium-cooled fast reactors) when exposed to the neutron spectrum derived from the Advanced Lead Fast Reactor European Demonstrator, a pool-type LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such a class of instrumentation does follow the power evolution, but is not completely suitable to detect the whole range of reactor power, due to excessive burnup, damages, or gamma interferences. Some improvements are possible to increase the signal-to-noise ratio by optimizing each instrument in the range of reactor power, so to get the best solution. The design of some new detectors is proposed here together with a possible approach for prototyping and testing them by a fast reactor.

Monte Carlo 방법을 이용한 로듐 및 바나듐 자발 중성자계측기의 연소에 따른 민감도 평가

Self-powered neutron detector (SPND) is a sensor to monitor a neutron flux proportional to a reactor power of the nuclear power plants. Since an SPND is usually installed in the reactor core and does not require additional outside power, it generates electrons itself from interaction between neutrons and a neutron-sensitive material called an emitter, such as rhodium and vanadium. This paper presents the simulations of the depletion sensitivity evaluations based on MCNP models of rhodium and vanadium SPNDs and light water reactor fuel assembly. The evaluations include the detail geometries of the detectors and fuel assembly, and the modeling of rhodium and vanadium emitter depletion using MCNP and ORIGEN-S codes, and the realistic energy spectrum of beta rays using BETA-S code. The results of the simulations show that the lifetime of an SPND can be prolonged by using vanadium SPND than rhodium SPND. Also, the methods presented here can be used to analyze a life-time of those SPNDs using various emitter materials.

A comparative study of Kalman filter and Linear Matrix Inequality based H infinity filter for SPND delay compensation

This paper deals with delay compensation of vanadium Self Powered Neutron Detectors (SPNDs) using Linear Matrix Inequality (LMI) based H-infinity filtering method and compares the results with Kalman filtering method. The entire study is established upon the framework of neutron flux estimation in large core Pressurized Heavy Water Reactor (PHWR) in which delayed SPNDs such as vanadium SPNDs are used as in-core flux monitoring detectors. The use of vanadium SPNDs are limited to 3-D flux mapping despite of providing better Signal to Noise Ratio as compared to other prompt SPNDs, due to their small prompt component in the signal. The use of an appropriate delay compensation technique has been always considered to be an effective strategy to build a prompt and accurate estimate of the neutron flux. We also indicate the noise-response trade-off curve for both the techniques. Since all the delay compensation algorithms always suffer from noise amplification, we propose an efficient adaptive parameter tuning technique for improving performance of the filtering algorithm against noise in the measurement.

Precise Control of Neutron Irradiation Fluence in the Neutron Transmutation Doping Process at HANARO Using SPND and Zirconium Foils

A neutron irradiation control method using a self-powered neutron detector (SPND) and zirconium foils was developed for neutron transmutation doping (NTD) application at HANARO. An SPND was installed at a fixed position of the upper part of the sleeve at the HANARO NTD hole for real-time monitoring of the neutron irradiation. The SPND signal output was significantly affected by the in-core conditions and surroundings of the facility. Furthermore, the SPND signal varied by $\sim 15\%$ throughout the reactor cycle according to the change of the control rod position. However, it was also confirmed that the variation of the neutron flux on the silicon ingots inside the irradiation rig was not significant for the control rod motion. In this procedure, the process for determining the thermal neutron flux by using zirconium foils was established for achieving efficient NTD irradiation. Accordingly, the relationship between the ratio of the neutron flux to the SPND output and the control rod position was deduced. Silicon irradiation experiments for transmutation doping were performed using the established relationship. The irradiated neutron fluence was controlled within ±1.3% of the target value. The mean irradiation/target ratio of the fluence was 0.9992, and the standard deviation was 0.0071. Thus, we confirmed that the produced irradiation was extremely accurate. This procedure can be useful in SPND applications requiring real-time high-accuracy neutron fluence measurements.

Monte Carlo 방법을 이용한 바나듐 자발 중성자계측기 초기 민감도 계산

Monte Carlo 방법을 이용한 바나듐 자발 중성자계측기 초기 민감도 계산

Material Irradiation Technology Using Capsules at HANARO

The Korean research program on nuclear technology requires various in-pile tests to evaluate the safety of currently operating nuclear power plants and develop a future nuclear system. HANARO (High-flux Advanced Neutron Application Reactor) provides a peak thermal and fast flux of 5.0 × 1014 n/cm2 · s (E < 0.625 MeV) and 2.1 × 1014 n/cm2 · s (E > 0.821 MeV), respectively, at 30 MW maximum thermal power. A capsule system has been developed for irradiation tests of nuclear materials and fuels at the core region. Extensive efforts have been made to establish the design and manufacturing technology for various capsules and a temperature control system, which should be compatible with HANARO’s characteristics. The capsule system has been actively utilized for various materials and fuel irradiation tests requested by users from research institutes, universities, and industries. Since 1995, more than 10,000 specimens have been irradiated using the developed capsule system. Presently, new irradiation technologies, such as a high-temperature irradiation capsule for a very-high-temperature reactor and a sodium-cooled fast reactor, and a low-temperature irradiation capsule for a research reactor, are being developed. These technologies will be completed with the development of instruments such as a linear variable differential transformer, a micro-electric heater, and a self-powered neutron detector in the near future, and they will be used for the Generation-IV nuclear system in Korea.

Research on the dynamic compensation algorithm of vanadium self-powered neutron detector based on robust filtering theory

Research on the dynamic compensation algorithm of vanadium self-powered neutron detector based on robust filtering theory

Online Fault Detection and Diagnosis of In-Core Neutron Detectors Using Generalized Likelihood Ratio Method

Vanadium self-powered neutron detectors (VSPNDs) have been proposed to be used in the Advanced Heavy Water Reactor for flux mapping purposes. However, response of VSPNDs to neutron flux variations is slow, and they might also develop faults. This paper proposes a hybrid scheme for state estimation based on the Kalman filtering approach and fault detection, diagnosis, and correction based on the generalized likelihood ratio (GLR) method for VSPNDs. The Kalman filter estimates prompt neutron flux variations from the delayed signal of VSPND, while the GLR method analyzes the innovations to detect fault in VSPND. Subsequently, the scheme also corrects for the step changes in the measurement emanating from fault in VSPND. Performance of the proposed hybrid scheme has been evaluated from simulation of neutron flux variations occurring due to simultaneous movement of regulating rods and demand power variations. It is shown that the magnitude and time of occurrence of step change in signal due to the fault are effectively estimated, and thus the step change can be corrected online.

Fault detection and isolation for self powered neutron detectors based on Principal Component Analysis

The self powered neutron detectors (SPNDs) play an important role in nuclear reactor monitoring. The 3-D power distribution and parameters used to evaluate the operation condition of reactor and the margin of safety can be determined using the measurement of SPNDs through power mapping procedure. Faulty SPNDs that are either completely or partially failed (hard fault or soft fault) provide incorrect information for monitoring. Correct detection and isolation of the faulty SPNDs are of primary importance to the efficient operation and management of the nuclear reactor. In this study, the methodology of Principal Component Analysis (PCA) is utilized to construct the mathematical models among various detectors at different axial location within the same string. The data used to build the mathematical models are generated by advanced neutronics code SMART rather than measurements. The square prediction error based on the model and the Detector Validity Index (DVI) based on the reconstruction are employed, respectively, to detect the SPND fault and to isolate the faulty SPNDs. The fault detection and isolation scheme is validated with four types of simulated SPND faults, i.e. bias, drifting, precision degradation and complete failure. The simulation results show that the proposed PCA based method can be used in the nuclear reactor to ensure that faulty SPNDs can be detected quickly.