Fission Chamber Research Articles

Studies of beam ion confinement to enhance plasma performance on EAST

A key physics issue for achieving steady-state high-performance plasmas on EAST tokamak is to decrease beam-ion losses to improve plasma confinement during neutral beam injections (NBIs). To decrease the beam losses, previous counter-I p NBI injections are upgraded to co-I p injections. Analysis shows that due to the reversed direction of drift across the flux surfaces caused by the pitch angle, the beam prompt loss fraction decreases from about 49% to 3% after the upgrade. Moreover, because of the change of entire beam path, beam shine-through (ST) loss fraction for counter-I p tangential and counter-I p perpendicular injections is reversed to co-I p tangential and co-I p perpendicular injections, respectively. Due to the change in the initial trapped-confined beam ion fraction caused by the peaked pitch profiles, the losses induced by toroidal ripple field are also reversed after the upgrade. To further improve the beam-ion confinement under the present NBI layout, the amplitudes of toroidal field are increased from 1.75 to 2.20 T. Result shows that, due to the smaller orbit width and peaked pitch angle profile, the beam prompt loss power is lower with higher toroidal field. Due to the synergy of higher initial trapped-confined beam ion fraction and narrower Goldston-White-Boozer (GWB) boundary, the loss induced by ripple diffusion is higher with higher toroidal field. The combined effect of beam ST loss, prompt loss and ripple loss, contributes to the increase in beam ion density. The decrease in beam loss power enhances beam heating efficiency, especially the fraction of beam heating ions. Finally, comparison between simulation and measurement by 235U fission chamber (FC) indicates that the increase in neutron rate is mainly contributed by improvement of beam-ion confinement. This study can provide potential support for beam operation and high-T i experiment on EAST tokamak.

Real-time monitoring and protection strategies for dense granular flow spallation target in Accelerator-Driven System

An innovative concept of a high-power, gravity-driven, dense granular spallation target has been integrated into the prototype design of the China initiative Accelerator-Driven System (CiADS), which is envisioned as a promising nuclear demonstration facility aimed at converting long-lived fission products and minor actinides into short-lived isotopes. This spallation target system, bridging the proton beam accelerator and the subcritical reactor, plays a crucial role in ensuring safety control and facilitating dynamic operational strategies. In this study, we have developed a system design for an online detector array, incorporating several neutron fission chambers and thermocouples installed in the gap between the spallation target and the subcritical reactor to monitor operations. Furthermore, we have devised real-time monitoring and protection methods to tackle the challenges of measuring beam position and the blockage condition associated with granular flow, considering the distribution of neutron flux and heat deposition under abnormal irradiation conditions. Monte Carlo simulations have demonstrated the applicability and feasibility of the control system for the dense granular spallation target in the accelerator-driven system. The numerical results confirm that the proposed control system meets the requirements for stable measurement with acceptable accuracy.

Accurate measurements of delayed neutron data for reactor applications: methodology and application to 235U(nth,f)

Large inconsistencies still exist in nuclear data libraries regarding the kinetic parameters of delayed neutron (DN) precursors. As an example, there is a 17% gap between ENDF-B/VIII.0 and JEFF-3.3 on the average lifetime T1/2 of DN precursors from thermal fission of 235U. This parameter is of major importance for reactivity predictions of nuclear reactors in nominal or accidental configurations. In this context, CEA is actively participating to the ALDEN project (Average number and Lifetime of DElayed Neutrons) which aims at providing the nuclear data community with new data sets of DN from thermal and fast neutron induced fission of various actinides. A dedicated experimental setup was designed and optimized for that purpose and is presented in this paper. It consists of a “long counter” detector containing 16 proportional counters filled with 3He, embedded in a high density polyethylene matrix. The detector surrounds a fissile target prepared in the form of a miniature fission chamber, containing a few hundreds of micro-grams of fissile material. This set-up is connected to fast and efficient neutron shutters that can produce step-irradiations of variable durations. The equations driving the DN counting following step-irradiations of the fissile target are established and discussed in the perspective of DN yield or group parameter measurement. A comprehensive analysis of the different steps of data reduction is detailed: dead time characterization, Region of Interest (ROI) determination, absolute and relative efficiency calibration, fission rate estimation, irradiation time and background determination, DN decay curve production and physical parameter fitting. Following a prototype experiment performed in 2018 at the PF1b cold neutron beam line of Institute Laue Langevin (ILL, Grenoble, France), we discuss here the analysis of two campaigns occurring in 2019 and 2021 in which significant improvements were achieved in terms of background minimization, counting statistics and fission rate determination. The achievements of this work are the measurement of the delayed neutron emission per fission for the thermal neutron induced fission of 235U, estimated at (1.625 ± 0.010) % and the group parameters leading to an estimated lifetime of T1/2\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$${T}_{1/2}$$\\end{document} = (8.87 ± 0.10) s. Those results are consistent with the values recommended by the IAEA/CRP work and they come with reduced uncertainties compared with previously published results.

Validation of MCNPX Calculational Model for Kartini Reactor Outer Core Neutron Flux

Abstract Kartini reactor in Yogyakarta is a nuclear reactor with a low thermal power of 100 kW that is mainly used for education, training, and research. For that purpose, measuring neutron flux and controlling the power are important operational parameters. The Fission Chamber (FC) type of detector is the main component in the reactor operation and safety system that measures the neutron flux in the outer reactor core. As the FC detector is placed in a fixed location, a numerical calculation model is needed to know the neutron flux in all outer core regions. In this study, the MCNPX code is used. The first step is to validate the model with experimental measurement. The objective of the present study is to obtain a validated model by comparing calculations using MCNPX with measurements using gold foil activation (Neutron Activation Analysis technique). The computational model tested showed promising results with an average difference compared to the gold foil activation method of about 23%. Further research can be done using more gold foil samples or foils from other materials to see the consistency of measurements or minimize the measurement difference.

Experimental validation of the 3-D neutron kinetics algorithm of RAPID using the JSI TRIGA Mark-II reactor

The high-fidelity simulation of reactor kinetics, with the ability of calculating detailed 3-D time-dependent distributions of the nuclear quantities of interest, is an important challenge in the field of nuclear engineering. As the industry moves towards innovative reactor designs and the financing and licensing of prototypes becomes increasingly complex, there is an increasing need of best-estimate and high-fidelity codes that are capable of replacing the computational tools that have been fine-tuned for LWRs using years of operation data and validation against benchmark facilities. Traditional neutron transport software either has to rely on significant modeling approximations or requires an impractical amount of computer time to calculate accurate and detailed solutions to the transport problems, failing to meet the need of the industry to simulate numerous scenarios over a relatively small time. RAPID is a software that relies on the Multi-stage Response-function Transport (MRT) methodology to achieve high-fidelity accurate solutions to the neutron transport problem by pre-calculating a database of response functions or coefficients that are then combined and interpolated to calculate the nuclear quantities of interest in seconds to minutes on a single computer core. In this article, the recently developed algorithms for 3-D time-dependent simulation of reactor kinetics and for control rods movement in the reactor core are validated using experiments performed at the Jožef Stefan Institute (JSI) TRIGA Mark-II reactor. The experiments were designed and performed with the goal of RAPID’s validation. In particular, time-dependent responses from four fission chambers located in the TRIGA core were collected and compared to RAPID-calculated values. The comparison of RAPID’s predictions to measured data shows very good agreement, demonstrating the ability of RAPID of calculating detailed 3-D time-dependent fission source distributions for a reactor kinetic problem that would be impractical to model with traditional neutron transport codes. In addition, it is demonstrated how RAPID is capable of obtaining a high level of accuracy within seconds to minutes on a single computer core, with a significant speedup with respect to state-of-the-art neutron transport methods.

Neutronics simulations for the design of neutron flux monitors in SPARC.

This paper presents the development and application of high-fidelity neutronic models of the SPARC tokamak for the design of neutron flux monitors (NFM) for application during plasma operations. NFMs measure the neutron flux in the tokamak hall, which is related to fusion power via calibration. We have explored Boron-10 gamma-compensated ionization chambers (ICs) and parallel-plate Uranium-238 fission chambers (FCs). We plan for all NFMs to be located by the wall in the tokamak hall and directly exposed to neutrons streaming through a shielded opening in a midplane port. Our simulations primarily use a constructive solid geometry-based OpenMC model based on the true SPARC geometry. The OpenMC model is benchmarked against a detailed CAD-based MCNP6 model. The B10 ICs are equipped with high-density polyethylene (HDPE) sleeves, borated HDPE housings, and borated aluminum covers to shield out scattered neutrons, optimize detector response levels, and make calibration robust against changes in the tokamak hall. The B10 neutron absorption branching ratio may cause the detectors' responses to be non-linear to neutron flux >200keV. However, our simulations unveil that, in the SPARC environment and with the proposed housings and sleeves, >99% of the detector responses are induced by <100keV neutrons. U238's insensitivity to slow neutrons makes this FC a promising candidate for direct fusion neutron measurements. Along with a borated HDPE sleeve, about 60% of the FCs' responses are induced by direct neutrons.

Gallium oxide (Ga2O3) energy dependent scintillation response to fast neutrons and flash gamma-rays.

Gallium oxide is a newly emerged ultrawide bandgap (4.9eV) semiconductor that is suitable as a combined electronics and radiation detection platform. We have experimentally demonstrated fast neutron and gamma-ray scintillation from Czochralski-grown β-Ga2O3 in a recent series (October 2023) of experiments at the unmoderated pulsed neutron spallation source located at the Los Alamos Neutron Science Center. Using the neutron time-of-flight (TOF) technique and a fast-gated intensified CCD camera, we observed energy-dependent neutron scintillation for neutron energies ranging from 1 to 400MeV, including the 14.1MeV neutron energy relevant to D-T fusion. Neutron flux is quantified and calibrated by cascading the scintillator after the fission chamber, enabling a detailed analysis of temporal and energy-dependent characteristics of the scintillation events. A pronounced scintillation signal from the spallation gamma flash with a temporal full width of half maximum of ∼4ns is indicative of the material's rapid response. Neutron energy dependent scintillation is observed using the TOF method at a 22.6-m distance from the neutron source. These results highlight the possibility of developing a Ga2O3 based fusion neutron diagnostic platform integrated with both scintillation and electronics functions on the integrated chip scale.

Prompt-critical power excursion measurements with optical fibers in the CABRI experimental reactor

This paper presents our experimental work to assess the capability of estimating the transient-induced power distribution evolution in the CABRI experimental reactor, using Cherenkov radiation in optical fibers. The CABRI reactor is designed to produce power transients from 100 kW up to 21 GW in less than 100 ms, in order to irradiate a test fuel pin under conditions representative of a Reactivity Insertion Accident in pressurized water reactors. Because of the wide response range and short response time required to follow the reactor power evolution during a complete transient, the usual means of detection, such as ionization or fission chamber, are rendered inoperable, especially for in-core or core vicinity measurements. For this reason, we suggest measuring the Cherenkov light produced within the optical fibers. The Cherenkov light emission is linked to local electron production, which is proportional to local gamma flux through the Compton or pair production cross-section. In other words, the intensity of Cherenkov radiation is related to the photon flux intensity. Knowledge of the fission photons emitted by the reactor gives direct insight into the fission rate, hence a spatial power density distribution could be reconstructed by using the measurement of Cherenkov light at different points in the reactor. The radio-luminescence measurements taken using photodiodes show very good linearity with the reactor power monitoring system despite the transient radiation induced attenuation observed in the optical fibers. The radio-luminescent light shows steady values with around 5% of variations over the 15 transients for low OH fibers and high OH fibers. Low OH tends to overestimate the FWHM of the power peak by 1–4%.

Measurement of the neutron total cross section of 9Be at the Back-n white neutron source of CSNS* *Supported by the National Key Research and Development Plan (2016YFA0401603), the National Natural Science Foundation of China (11675155, 11790321) and Foundation of President of China Academy of Engineering Physics (YZJJLX2016003)

The neutron total cross section data of 9Be are essential in the nuclear structure model research of light nuclei and nuclear power installations. The neutron total cross section of 9Be in the 0.3 eV−120 MeV energy region has been measured using time-of-flight and transmission methods with the Neutron Total Cross Sectional Spectrometer (NTOX) based on the multi-cell fast fission chamber at the China Spallation Neutron Source (CSNS)-Back-n white neutron source (Back-n). The fission count-neutron energy distributions of 235U and 238U without samples and with Be samples with three thicknesses were measured in the double-bunch operation mode for a beam power of 100 kW. The Bayesian method was used to eliminate the influence of the double-bunch problem on neutron measurement in the energy region above 10 keV. The neutron total cross section of 9Be results was consistent with ENDF/B-VIII.0 evaluation library data in the 0.3 eV−20 MeV energy region. In the energy ranges of 0.3 eV to 10 keV and 0.01 to 20 MeV, the deviations between our results and the evaluation results of ENDF/B-VIII.0 were within 2.5% and 15%, respectively. In the resonance energy region, the measured resonance energies in our experiment were 0.63, 0.82, and 2.8 MeV, respectively. The results showed that the total cross section uncertainties of three Be samples were within 2.2% in the energy region below 1 MeV. The total cross section uncertainty of 30 mm Be from 235U was the smallest and less than 5% in the energy region of 0.3 eV−120 MeV. The results of this experiment can provide technical support for further data analysis and related nuclear data evaluation.

Design and characterization of a new high-count wide-range fission chamber based on GEM

As one of the main detectors for monitoring neutron flux rate in a nuclear reactor, the fission chamber (FC) suffers from issues such as low sensitivity, limited counting rate dynamic range, and system mode switching is cumbersome. This study utilizes the advantages of gas electron multipliers (GEM), which are easy to fabricate in large areas and have high counting rates, to design a novel fission chamber. By conducting Monte Carlo simulations on parameters such as the thickness of the 235U coating, the drift distance, and the operating electric field strength, a wide-range fission chamber design that combines high sensitivity and a counting rate range based on GEM has been obtained. The simulation results demonstrated that using a GEM detector to build a fission chamber can significantly improve sensitivity and extend the counting rate range. Subsequently we built a proof-of-concept GEM-based fission chamber and tested it with a 55Fe low-energy X-ray source and an Am-Be neutron source. The results showed that the proof-of-concept detector had a good signal-to-noise ratio and energy linearity, as well as clear discrimination between alpha background and neutron pulse amplitudes.

A computational- experimental model of reactor kinetic for investigating the linearity response of in-core neutron detectors of a low power research reactor

In this study, the linear response of fission chamber (FC) neutron detectors in Isfahan miniature neutron source reactor (MNSR) is investigated through experiments and calculations at different powers of the reactor. The linearity of the response of the main neutron detector in the MNSR is important due to its safety functions in reactor power control. Additionally, determining the linear response region of this detector presents a linear neutron dosimetry field even outside of the reactor core, i.e. at the outlet of external beam tubes. Moreover, it is very difficult to remove and calibrate this neutron detector that has been installed in the reactor core for a long time. For this purpose, an alternative method for measuring the linearity of the response of a neutron detector is extended by investigating the kinetic behavior of reactor. The aim of this method is to calculate the delayed neutrons population in the reactor core and compare with the FC detector response. In this method, the numerical solution method is used to solve the point kinetic equation with 15 delayed neutron groups (6 delayed neutron groups and 9 photo-neutron groups) after inserting a large negative reactivity into the MNSR core by considering temperature and xenon reactivity feedbacks during a coupling procedure. To obtain the xenon reactivity feedback, xenon concentration is calculated by coupling the 135Xe and 135I production equations and the εpPFnlPTnl parameter is obtained by an experimental method. Furthermore, the value of fission cross section (Σf) is calculated using a validated model in the Dragon code. The large reactivities are inserted at two initial power of 150 W and 30 kW to investigate the response of the main FC neutron detector in a cold xenon-free condition and by considering temperature and xenon reactivity feedbacks, respectively. The results demonstrate that the response of the MNSR FC detector is linear in the neutron flux range between 5 × 107 to 1 × 1012 n.cm−2.s−1, which is equivalent to reactor power of 1.5 W to 30 kW. This innovative method can be applied to calibrate the neutron detectors in other research and power reactors.

JSI TRIGA fuel rod reactivity worth experiments for validation of Serpent-2 and RAPID fuel burnup calculations

Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient ΔρBU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.

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

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

Measurement and modelling signals from an optical fission chambers during reactor irradiation

Nowadays, fission chambers are commonly used for neutron diagnostic in industrial pressurized water reactors. However, with the diversity of emerging nuclear power plant prototypes such as fusion, fast neutron or high temperature reactors, innovative neutron diagnostic system are developed by the French Alternative Energies and Atomic Energy Commission (CEA). Indeed, in some case the operating range of existing fission chambers will not be adapted to the measuring conditions. Thus, the CEA is developing a technology called optical fission chambers, where photons are collected instead of charged particles. This paper presents a macroscopic model of the signal-to-noise ratio of optical fission chamber measuring in neutron and gamma fields. Due to the novelty of this study, we expect to have a better understanding about the intensity of the different light sources. Secondly we aim to validate the developed numerical model by comparing the simulated spectra to the experimental ones. Finally, we highlight some design features which may improves the signal to noise ratio of optical fission chamber.

Effects of Neutron Flux Distribution and Control Rod Shadowing on Control Rod Calibrations in the Oregon State TRIGA® Reactor

Control rod calibration experiment results for the Oregon State TRIGA® Reactor (OSTR) immediately following LEU conversion in 2008, and MCNP® 5 predicted rod worths from the 2008 LEU Conversion Safety Analysis Report (CSAR) are discussed. The reactivity worth of the four OSTR control rods is measured using the rod-pull method. Reactor power and period measurements in this method rely on the fission chamber power detector on the north side of the reflector. It is proposed that the location of the fission chamber and the neutron flux distribution in the core may result in an inaccurate reactor period measurement due to the asymmetry of the neutron flux distribution in the OSTR core. The asymmetry of the flux is believed to be more pronounced during super-criticality, resulting in errors in the time-of-power-rise measurements. As a result, control rod calibration experiments may under-predict or over-predict the reactivity worth of certain control rods. A time-independent Monte–Carlo method for the quantification of these effects is presented. Thermal flux maps at the core axial mid-plane are obtained from the model to inform discrepancies between predicted and observed results.

Fast simulation of neutron noise using the Transient Fission Matrix approach and validation on the CROCUS reactor

This article presents the development of a neutron noise calculation technique with a reduced calculation time based on the TFM (Transient Fission Matrix) approach. The latter contains different information such as the system transfer function (fission and absorption probability density functions and corresponding neutron multiplicity) of prompt and delayed neutrons, and the corresponding propagation time. This information is used to reconstruct neutron showers based on precursor decays up to the shower disappearance by absorption, leakage or delayed neutron precursor creation. Then pseudo-detectors such as fission chamber count rate histograms are reconstructed from these neutron showers and the αprompt is calculated from the cross-correlation function. A very good agreement has been obtained on the uranium-fueled light water reactor CROCUS between the model developed and the reference value.

Validating the simulation of beam-ion charge exchange in MAST Upgrade

Simulation of the impact of charge-exchange (CX) reactions on beam ions in the Mega Amp Spherical Tokamak (MAST) Upgrade was compared to measurements carried out with a fission chamber (neutron fluxes) and a fast ion deuterium-alpha (FIDA) diagnostic. A simple model was developed to reconstruct the outer-midplane neutral density based on measurements of deuterium-alpha emission from edge neutrals, and on Thomson scattering measurements of electron density and temperature. The main computational tools used were the ASCOT orbit-following code and the FIDASIM code for producing synthetic FIDA signals. The neutral density reconstruction agrees qualitatively with SOLPS-ITER modelling and yields a synthetic passive FIDA signal that is consistent with measurement. When CX losses of beam ions are accounted for, predicted neutron emission rates are quantitatively more consistent with measurement. It was necessary to account for CX losses of beam ions in simulations to reproduce the measured passive FIDA signal quantitatively and qualitatively. The results suggest that the neutral density reconstruction is a good approximation, that CX with edge neutrals causes significant beam-ion losses in MAST Upgrade, typically 20% of beam power, and that the ASCOT fast-ion CX model can be used to accurately predict the redistribution and loss of beam ions due to CX.

A detector system for `absolute' measurements of fission cross sections at n_TOF in the energy range below 200 MeV

A new measurement of the 235U(n,f) cross section was performed at the neutron time-of-flight facility n_TOF at CERN. The experiment focused on neutron energies from 20 MeV to several hundred MeV, and was normalized to neutron scattering on hydrogen. This is a measurement first of its kind at this facility, in an energy range that was until now not often explored, so the detector development phase was crucial for its success. Two detectors are presented, a parallel plate fission chamber (PPFC) and a recoil proton telescope (RPT), both dedicated to perform measurements in the incident neutron energy range from 30 MeV to 200 MeV. The experiment was designed to minimize statistical uncertainties in the allocated run time. Several efforts were made to ensure that the systematic effects were understood and under control. The results show that the detectors are suited for measurements at n_TOF above 30 MeV, and indicate the path for possible future lines of development.

Development and characterization of fission chamber neutron detectors in Isfahan miniature neutron source reactor

In this research, two fission chamber neutron detectors have been designed and constructed, and their performance has been evaluated in the radiation field of the Isfahan miniature reactor neutron beam line. Each detector consists of two coaxial cylinders called anode and cathode made of stainless steel. The outer layer of the anode is covered with a very thin layer of uranium-235 oxide as fissile material, and a mixture of argon and nitrogen gas is used as the filling gas. The thickness of the deposited layer on the anode is different, so that on one of them, named FC400A, the thickness of the uranium oxide layer was more than 5 μm, and on the other one, named FC400B, it was less than 1.5 μm. A spectrum collection software has also been designed and developed in LabVIEW. The developed detectors are irradiated in the neutron field resulting from the neutron radiography beamline of the miniature neutron source reactor in the range of neutron flux of about 2E+02 to 2E+05 n.cm−2.s−1. In order to determine the sensitivity of the detection response, the results of these two detectors are also compared with a commercial fission chamber neutron detector. The results showed that the working voltage range of these detectors is in the range of 400–600 V and both of these detectors have a suitable linear response in the mentioned neutron flux range. Also, the sensitivity of FC400A and FC400B detectors is 0.1 and 1 counts/1E+04 neutrons, respectively.

Determination of the linear behavior of FC detectors in Isfahan MNSR using ex-core offline and online experiments

The linear behavior of fission chamber (FC) neutron detectors in Isfahan miniature neutron source reactor (MNSR) was investigated experimentally. FC neutron detectors are installed near the reactor core to measure the produced neutron flux in the core. The linearity of the response of these detectors is important because the measured neutron flux by them is one of the important methods in determining and controlling the reactor power and therefore affects the safety of the reactor. For this purpose, the measured neutron fluxes by these detectors at different reactor powers were compared with the measured neutron flux at the output of the external beam tubes by different methods. In this study, two online and offline neutron flux detection methods were used to analysis the linear behavior of MNSR neutron detectors. The BF3 detector and solid-state nuclear track detector (SSNTD) were used to determine the neutron flux in the online and offline methods, respectively. The results of two diverse methods showed that the response of MNSR FC detectors for measuring the reactor power is linear with acceptable accuracy in the neutron flux range between 5×107 to 1012n.cm−2. s−1, which is equivalent to reactor power of 1.5 W to 30 kW.