High-precision and wide-range real-time neutron flux monitor system through multipoint linear calibration

The neutron flux monitor(NFM)system is an important diagnostic subsystem introduced by large nuclear fusion devices such as international thermonuclear experimental reactor(ITER),Japan torus-60,tokamak fusion test reactor,and HL-2 A.Neutron fluxes can provide real-time parameters for nuclear fusion,including neutron source intensity and fusion power.Corresponding to different nuclear reaction periods,neutron fluxes span over seven decades,thereby requiring electronic devices to operate in counting and Campbelling modes simultaneously.Therefore,it is crucial to design a real-time NFM system to encompass such a wide dynamic range.In this study,a high-precision NFM system with a wide measurement range of neutron flux is implemented using realtime multipoint linear calibration.It can automatically switch between counting and Campbelling modes with variations in the neutron flux.We established a testing platform to verify the feasibility of the NFM system,which can output the simulated neutron signal using an arbitrary waveform generator.Meanwhile,the accurate calibration interval of the Campbelling mode is defined well.Based on the above-mentioned design,the system satisfies the requirements,offering a dynamic range of 10~8 cps,temporal resolution of 1 ms,and maximal relative error of 4%measured at the signal-to-noise ratio of 15.8 dB.Additionally,the NFM system is verified in a field experiment involving HL-2 A,and the measured neutron flux is consistent with the results.

Measurement of Neutron Flux at the Initial Phase of Discharge in EAST

Neutron emission in EAST was investigated by a time-resolved monitor system which consists of four 3He proportional counters and a 235U fission chamber. The D-D neutron flux increased approximately an order of magnitude during the 27 MHz ion-cyclotron radio frequency （ICRF） heating, demonstrating that the ICRF wave heated the plasma effectively. In addition in lower hybrid wave （LHW） experiment with higher plasma parameters D-D neutrons were also detected. However, masses of photoneutrons were generated in Ohmic discharges with low plasma density. Effect of plasma density on the production of photoneutrons was studied, and it is found that LHW can suppress the generation of the runaway electrons and reduce the share of the photoneutrons effectively.

Development of a wide-range and fast-response digitizing pulse signal acquisition and processing system for neutron flux monitoring on EAST

The neutron count rate fluctuation reaches six orders of magnitude between the ohmic plasma scenario and high power of auxiliary heating on an experimental advanced superconducting tokamak(EAST).The measurement result of neutron flux monitoring(NFM)is a significant feedback parameter related to the acquisition of radiation protection-related information and rapid fluctuations in neutron emission induced by plasma magnetohydrodynamic activity.Therefore,a wide range and high time resolution are required for the NFM system on EAST.To satisfy these requirements,a digital pulse signal acquisition and processing system with a wide dynamic range and fast response time was developed.The present study was conducted using a field-programmable gate array(FPGA)and peripheral component interconnect extension for instrument express(PXIe)platform.The digital dual measurement modes,which are composed of the pulse-counting mode and AC coupled square integral's Campbelling mode,were designed to expand the measurement range of the signal acquisition and processing system.The time resolution of the signal acquisition and processing system was improved from 10 to 1 ms owing to utilizing highspeed analog-to-digital converters(ADCs),a high-speed PXIe communication with a direct memory access(DMA)mode,and online data preprocessing technology of FPGA.The signal acquisition and processing system was tested experimentally in the EAST radiation field.The test results showed that the time resolution of NFM was improved to 1 ms,and the dynamic range of the neutron counts rate was expanded to more than 10^(6) counts per second.The Campbelling mode was calibrated using a multipoint average linear fitting method;subsequently,the fitting coefficient reached 0.9911.Therefore,the newly developed pulse signal acquisition and processing system ensures that the NFM system meets the requirements of high-parameter experiments conducted on EAST more effectively.

Development of a data acquisition and control system for the International Thermonuclear Experimental Reactor neutron flux monitor

To satisfy high-precision,widc-rangc,and real-time neutron flux measurement requirements by the International Thermonuclear Experimental Reactor(ITF.R),a data acquisition and control system based on fission chamber detectors and fast controller technology,has been developed for neutron flux monitor in ITER Equatorial Port#7.The signal processing units which arc based on a field programmable gate array and the PXI Express platform arc designed to realize the neutron flux measurement with I ms time resolution and a fast response less than 0.2 ms,together with real-time timestamps provided by a timing hoard.The application of the wide-range algorithm allows the system to measure up to 10^10cps with a relative error of less than 5%.Furthermore,the system is managed and controlled by a software based on the Experimental Physics and Industrial Control System,compliant with COntrol.Data Access and Communication architecture.

A novel digital neutron flux monitor for international thermonuclear experimental reactor

A novel full-digital real-time neutron flux monitor（NFM） has been developed for the International Thermonuclear Experimental Reactor.A measurement range of 10~9 counts per second is achieved with 3 different sensitive fission chambers.The Counting mode and Campbelling mode have been combined as a means to achieve higher measurement range.The system is based on high speed as well as parallel and pipeline processing of the field programmable gate array and has the ability to upload raw-data of analog-to-digital converter in real-time through the PXIe platform.With the advantages of the measurement range,real time performance and the ability of raw-data uploading,the digital NFM has been tested in HL-2 A experiments and reflected good experimental performance.

Fusion Neutron Flux Monitor for ITER

Neutron flux monitor （NFM） as an important diagnostic sub-system in ITER （international thermonuclear experimental reactor） provides a global neutron source intensity, fusion power and neutron flux in real time. Three types of neutron flux monitor assemblies with different sensitivities and shielding materials have been designed. Through MCNP （Mante-Carlo neutral particle transport code） calculations, this extended system of NFM can detect the neutron flux in a range of 10^4 n/（cm^2.s） to 10^14 n/（cm^2.s）. It is capable of providing accurate neutron yield measurements for all operational modes encountered in the ITER experiments including the in-situ calibration. Combining both the counting mode and Campbelling （MSV; Mean Square Voltage） mode in the signal processing units, the requirement of the dynamic range （107） for these NFMs and time resolution （1 ms） can be met. Based on a uncertainty analysis, the estimated absolute measurement accuracies of the total fusion neutron yield can reach the required 10% level in both the early stage of the DD-phase and the DD-phase, the absolute measurement accuracy full power DT operation mode. In the advanced would be better than 20%.

Determining absolute value of thermal neutron flux density based on monocrystalline silicon in nuclear reactors

A new type of neutron detector based on monocrystalline Si is developed to measure the fluence and flux density of thermal and fast neutrons.The principle of this detector is based on the relationship between changes in electrical conductivity and neutron fluence during irradiation.Therefore,the absolute values of thermal neutron fluence and flux density are measured in a facile manner with high reliability.Compared with activation methods,our method not only possesses a similar accuracy,but also demonstrates superior application potential for the investigation of neutron fields in nuclear reactors owing to its suitable half-life.

Development of Prototype Neutron Flux Monitor for ITER

The prototype neutron flux monitor consists of a high purity ^(235)U fission chamber detector and a'blank'detector,which is a fissile material free detector with the same dimension as the fission chamber detector to identify noise issues such as noise coming from gamma rays.The main parameters of the fission chamber assembly that have been measured in the laboratory are confirmed to approach the technological level of the International Thermonuclear Experimental Reactor(ITER)in the near future.This prototype neutron flux monitor will be further developed to become a neutron flux monitor suitable for the operation phase of D-D fusion on the ITER.

Fusion Neutron Flux Detector for the ITER

Two fission chambers with different amounts of fissile material （enriched 90% uranium-235） have been manufactured for neutron flux detection in a thermonuclear fusion device. The characteristics of neutron signal and its discrimination from other signals, and a plateau of high voltage between the anode and cathode have been validated in a thermal neutron source. The energy responses of the two fission chambers at seven energy levels have been calibrated in an accelerator fast neutron source and the results agree well with the simulations.

Measurement of the Energy Spectrum of the Neutrons inside the Neutron Flux Trap Assembled in the Center of the Reactor Core IPEN/MB-01

This paper presents the neutron energy spectrum in the central position of a neutron flux trap assembled in the core center of the research nuclear reactor IPEN/MB-01, obtained by an unfolding method. To this end, we have used several different types of activation foils （Au, Sc, Ti, Ni, and plates） which have been irradiated in the central position of the reactor core （setting number 203） at a reactor power level （64.57±2.91 watts）. The activation foils were counted by solid-state detector HPGe （high pure germanium detector） （gamma spectrometry）. The experimental data of nuclear reaction rates （saturated activity per target nucleus） and a neutron spectrum estimated by a reactor physics computer code are the main input data to get the most suitable neutron spectrum in the irradiation position obtained through SANDBP （spectrum analysis neutron detection code-version Budapest University） code： a neutron spectra unfolding code that uses an iterative adjustment method. the integral neutron flux, （2.41 ± 0.01） × 10^9 n·cm^-2·s^-1 for the thermal The adjustment resulted in （3.85 ± 0.14） × 10^9 n·cm^-2·s^-1 for neutron flux, （1.09 ±0.02） × 10^9n·cm^-2·s^-1 for intermediate neutron flux and （3.41 ± 0.02） × 10^8 n·cm^-2·s^-1 for the fast neutrons flux. These results can be used to verify and validate the nuclear reactor codes and its associated nuclear data libraries, besides, show how much effective it can be that the use of a neutron flux trap in the nuclear reactor core to increase the thermal neutron flux without increase the operation reactor power level. The thermal neutral flux increased 4.04 ± 0.21 times compared with the standard configuration of the reactor core.

Reconstruction of the Neutron Flux in a Slab Reactor

In this electronic article we use the one-dimensional multigroup neutron diffusion equation to reconstruct the neutron flux in a slab reactor from the nuclear parameters of the reactor, boundary and symmetry condition, initial flux and?keff. The diffusion equation was solved analytically for one single homogeneous fuel region and for two regions considering fuel and reflector. To validate the method proposed, the results obtained in this article were compared using reference methods found in the literature.

Variation of environmental neutron flux with altitude and depth of both water and soil

Applying the extreme low-level γ-ray spectroscopic analysis the environmental neutron flux is measured using different moderator construction and environment through the reaction 197Au (n, γ) 198Au. The contribution of thermal and resonance neutrons is separated using the cadmium difference technique, while fast neutrons are measured by the paraffin moderator. The results of altitude dependence of the neutron flux are discussed. The thermal neutron flux near the surface of sea water is less than its value at 100 cm over ground near sea water, while the value over the surfaces of fresh water is higher than that near the surface of sea water. Also the thermal neutron flux at 5 cm soil depth increases, then decreases to its original value at 10 cm depth and still constant until 25 cm, then decreases rapidly to reach 27% of its original value at 60 cm depth. The soil compositions, corresponding neutron temperatures and effective absorption cross sections of earth are the most effective factors on the equilibrium region of thermal neutrons in the ground.

Dynamic linear calibration method for a wide range neutron flux monitor system in ITER

As a key part of the diagnosis system in the International Thermonuclear Experimental Reactor(ITER),the neutron flux monitor(NFM),which measures the neutron intensity of the fusion reaction,is a Counting-Campbelling system with a large dynamic counting range.A dynamic linear calibration method is proposed in this paper to solve the problem of cross-over between the different counting and Campbelling channels,and improve the accuracy of the cross-calibration for long-term operation.The experimental results show that the NFM system with the dynamic linear calibration system can obtain the neutron flux of the fusion reactor in real time and realize the seamless measurement area connection between the two channels.

Application of global variance reduction method to calculate a high-resolution fast neutron flux distribution for TMSR-SF1

The solid fuel thorium molten salt reactor(TMSR-SF1) is a 10-MWth fluoride-cooled pebble bed reactor. As a new reactor concept, one of the major limiting factors to reactor lifetime is radiation-induced material damage. The fast neutron flux(E > 0.1 MeV) can be used to assess possible radiation damage. Hence, a method for calculating high-resolution fast neutron flux distribution of the full-scale TMSR-SF1 reactor is required. In this study,a two-step subsection approach based on MCNP5 involving a global variance reduction method, referred to as forward-weighted consistent adjoint-driven importance sampling, was implemented to provide fast neutron flux distribution throughout the TMSR-SF1 facility. In addition,instead of using the general source specification cards, the user-provided SOURCE subroutine in MCNP5 source code was employed to implement a source biasing technique specialized for TMSR-SF1. In contrast to the one-step analog approach, the two-step subsection approach eliminates zero-scored mesh tally cells and obtains tally results with extremely uniform and low relative uncertainties.Furthermore, the maximum fast neutron fluxes of the main components in TMSR-SF1 are provided, which can be used for radiation damage assessment of the structural materials.

Real-time wide-range neutron flux monitor for thorium-based molten salt reactor

A novel full-digital real-time neutron flux monitor(NFM) has been developed for thorium-based molten salt reactor(TMSR).The system is based on the highspeed,parallel,and pipeline processing of the field programmable gate array as well as the high-stability controller area network platform.A measurement range of 10~8 counts per second is achieved with a single fission chamber by utilizing the normalization of the count and Campbell algorithms.With the advantages of using the measurement range,system integrity,and real-time performance,digital NFM has been tested in the Xi'an pulsed reactor fission experiments and was found to exhibit superior experimental performance.

Influence of the Neutron Flux Characteristic Parameters in the Irradiation Channels of Reactor on NAA Results Using k₀-Standardization Method

An approximation method using to estimate the influence of the uncertainties of the neutron flux characteristic parameters in the irradiation positions on the NAA results using k0-standardization technique was presented. Those are the epithermal reactor neutron spectrum shape-factor α, the effective resonace energy Ε for a given nuclide and the thermal to epithermal neutron flux ratio f. The method is applied to estimate the effect of the uncertainties in the determination of α Ε, and f on final NAA results for some irradiation channels of the Dalat reactor. It also shows that presented method is suitable in practical use for the estimation of the errors due to the uncertainty of the neutron flux characteristic parameters at the irradiation position.

Determination of Neutron Fluxes and Spectrum Shaping Factors in Irradiation Sites of Ghana’S Miniature Neutron Source Reactor (mnsr) by Activation Method After Compensation of Loss of Excess Reactivty

Accurate neutron flux values in irradiation channels of research reactors are very essential to their usage. The total neutron flux of the Ghana Research Reactor-1(GHARR-1) was measured after a beryllium reflector was added to its shim to compensate for excess reactivity loss. The thermal, epithermal and fast neutron fluxes were determined by the method of foil activation. The experimental samples with and without a cadmium cover of 1-mm thickness were irradiated in the isotropic neutron field of the irradiation sites of Ghana Research Reactor-1 facility. The induced activities in the sample were measured by gamma ray spectrometry with a high purity germanium detector. The necessary correction for gamma attenuation, thermal neutrons and resonance neutron self-shielding effects were taken into account during the experimental analysis. By defining cadmium cutoff energy of 0.55eV, Al-0.1% Au wires of negligible thickness were irradiated at 3kW to determine the neutron fluxes of two irradiation channels, outer channel 7 and inner channel 2 whose Neutron Shaping Factor (α) were found to be (0.037 ± 0.001) and (–0.961 ± 0.034). The neutron flux ratios at the inner irradiation site 2 were found to be, (25.308 ± 3.201) for thermal to epithermal neutrons flux, (0.179 ± 0.021) for epithermal to fast neutrons flux and (4.528 ± 0.524) for thermal to fast neutrons flux, in the outer irradiation site 7, the neutron flux ratios were found to be, (40.865 ± 3.622) for thermal to epithermal neutrons flux, (0.286 ± 0.025) for epithermal to fast neutrons flux and (11.680 ± 1.030) for thermal to fast neutrons flux.

FPGA-based position reconstruction method for neutron beam flux spatial distribution measurement in BNCT

A new measurement method for the spatial distribution of neutron beam flux in boron neutron capture therapy(BNCT)is being developed based on the two-dimensional Micromegas detector.To address the issue of long processing times in traditional offline position reconstruction methods,this paper proposes a field programmable gate array based online position reconstruction method utilizing the micro-time projection chamber principle.This method encapsulates key technical aspects:a self-adaptive serial link technique built upon the dynamical adjustment of the delay chain length,fast sorting,a coordinate-matching technique based on the mapping between signal timestamps and random access memory(RAM)addresses,and a precise start point-merging technique utilizing a circular combined RAM.The performance test of the selfadaptive serial link shows that the bit error rate of the link is better than 10-12 at a confidence level of 99%,ensuring reliable data transmission.The experiment utilizing the readout electronics and Micromegas detector shows a spatial resolution of approximately 1.4 mm,surpassing the current method’s resolution level of 5 mm.The beam experiment confirms that the readout electronics system can obtain the flux spatial distribution of neutron beams online,thus validating the feasibility of the position reconstruction method.The online position reconstruction method avoids traditional methods,such as bubble sorting and traversal searching,simplifies the design of the logic firmware,and reduces the time complexity from O(n2)to O(n).This study contributes to the advancement in measuring neutron beam flux for BNCT.

Measurement of neutron and charged particle fluxes toward earthquake prediction

In this paper, we describe a possible method for predicting the earthquakes, which is based on simultaneous recording of the intensity of fluxes of neutrons and charged particles by detectors, commonly used in nuclear physics. These low-energy particles originate from radioactive nuclear processes in the Earth＇s crust. The variations in the particle flux intensity can be the precursor of the earth- quake. A description is given of an electronic installation that records the fluxes of charged particles in the radial direction, which are a possible response to the accumulated tectonic stresses in the Earth＇s crust. The obtained results showed an increase in the intensity of the fluxes for 10 or more hours before the occurrence of the earthquake. The previous version of the installation was able to indicate for the possibility of an earthquake （Maksudov et al. in Instrum Exp Tech 58：130-131, 2015）, but did not give information about the direction of the epicenter location. In this regard, the installation was modified by adding eight directional detectors. With the upgraded setup, we have received both the predictive signals, and signals determining the directions of the location of the forthcoming earthquake, starting 2-3 days before its origin.

Monitoring method for neutron flux for a spallation target in an accelerator driven sub-critical system

In this paper, we study a monitoring method for neutron flux for the spaUation target used in an accelerator driven sub-critical （ADS） system, where a spallation target located vertically at the centre of a sub-critical core is bombarded vertically by high-energy protons from an accelerator. First, by considering the characteristics in the spatial variation of neutron flux from the spallation target, we propose a multi-point measurement technique, i.e. the spallation neutron flux should be measured at multiple vertical locations. To explain why the flux should be measured at multiple locations, we have studied neutron production from a tungsten target bombarded by a 250 MeV-proton beam with Geant4-based Monte Carlo simulations. The simulation results indicate that the neutron flux at the central location is up to three orders of magnitude higher than the flux at lower locations. Secondly, we have developed an effective technique in order to measure the spallation neutron flux with a fission chamber （FC）, by establishing the relation between the fission rate measured by FC and the spallation neutron flux. Since this relation is linear for a FC, a constant calibration factor is used to derive the neutron flux from the measured fission rate. This calibration factor can be extracted from the energy spectra of spallation neutrons. Finally, we have evaluated the proposed calibration method for a FC in the environment of an ADS system. The results indicate that the proposed method functions very well.