Rhodium Self-powered Neutron Detector 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.

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.

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

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.

Analysis and comparison of direct inversion and Kalman filter methods for self-powered neutron detector compensation

Many light-water reactors use self-powered neutron detectors (SPNDs) to monitor neutron flux. Some SPNDs suffer from delayed response. To compensate for this inherent delay, many methods such as the direct inversion method (DIM) and the Kalman filter method can be used. In the current investigation, these two methods were applied to the most common SPNDs, rhodium SPND (Rh-SPND) and vanadium SPND (V-SPND). The DIM, a leading compensation method, is used for V-SPND and Rh-SPND response compensation. The Kalman filter, which is a closed-loop compensation method, has also been developed for both SPNDs. The performance of the DIM and Kalman method for V-SPND and Rh-SPND was computed and surveyed with various sampling times. To evaluate the sensitivity of the compensator algorithm to miscalculations of the prompt fraction of the detector signal, computations were also conducted with various prompt fractions. Results show that the Kalman method is much more stable with different prompt fractions. Both methods improve the response time of V-SPNDs and Rh-SPNDs, and thus would allow them to be used for reactor protection, reactor control functions, reactor core monitoring, and for fulfilling surveillance requirements.

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.

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.

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.

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.

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.

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.

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.

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.

A Correlation-Based Signal Validation Method for Fixed In-Core Detectors

Recently, continuous power reactor core surveillance, which is based on fixed in-core detector readings, has exhibited a trend of growing significance. The fixed in-core sensors can only be replaced during shutdown periods; therefore, it is important to have reliable information on the quality of each detector in advance to be able to create an appropriate detector replacement schedule. During the operating cycle, the continuous core surveillance system should rely only on reliable measurements, and only an effective detector failure diagnosis can ensure avoiding falsified information. At the same time, most published signal validation methods are not well suited for an extensive set of fixed in-core detectors. A relatively simple, though powerful and robust, method is proposed that can be applied for both signal validation and early failure detection. The basic idea of the method is that inevitably there exist such process noise components in the detector signals that are characteristically correlated within some well-determined groups of sensors. The lack of such correlation most probably occurs due to some detector failure. When a smaller, localized subgroup of the detectors shows a decreased level of correlation with the majority, that is typically caused by some abnormal event in the technological process. In such cases the results of this method can be utilized as a target identification tool for the more sophisticated noise diagnostics methods. The method has been thoroughly tested with an extensive data set, including rhodium self-powered neutron detectors and assembly outlet thermocouple signals, which was collected throughout a complete operational cycle of a VVER-440/213-type pressurized water reactor.

Autoregressive Moving Average Inverse Dynamics for Rhodium Self-Powered Neutron Detector

AbstractThe autoregressive moving average (ARMA) solution is derived for the rhodium self-powered neutron detector inverse dynamics. Numerical performance is compared to the autoregressive (AR) inverse dynamics form. A potential difference in the dynamic range of the two methods would be masked by analog-to-digital converter quantizing noise. TheAR and ARMA inverse dynamics methods have equal performance for additive noise and require the same number of multiplications and additions per time step.

Dynamic Compensation Methods for Self-Powered Neutron Detectors

Among the three digital dynamic compensation methods that are developed for or applied to the rhodium self-powered neutron detector–the dominant pole Tustin method of the core operating limit super...

Instantaneous flux measurements using the background signal of the rhodium self-powered neutron detector

The background current produced by a rhodium self-powered neutron detector (RSPND) with an integral background lead has been investigated to determine its relationship to the neutron flux seen by the detector. The background current and steady state flux were measured at the core midplane for all fuel assemblies, and good linearity was seen. Under transient conditions, the background current, when delayed fission product decay contributions are included, follows the flux reasonably well although with a slight lag. This opens the possibility of using the RSPND during transient conditions.

Analog and Digital Dynamic Compensation Techniques for Delayed Self-Powered Neutron Detectors

Analog and digital methods have been developed to compensate for the time delay associated with rhodium self-powered neutron detector signals. This delay is caused by the decay of the neutron-activ...