Measurement Of Neutron Spectra Research Articles

Estimation of Bonner sphere cross-talking with Monte Carlo method and spectrometer calibration with 241Am-Be neutron source

There are usually two kinds of experiment methods for neutron spectrum measurement by Bonner sphere spectrometer (BSS). The first is to use only one Bonner sphere (BS) at a time to complete the neutron experiment by repeated measurements; The other is to measure multiple Bonner spheres (BSs) together, and this multi-sphere simultaneous measurement method can complete the neutron experiment only once. The single sphere measurement method cannot guarantee that the measurement environment is completely consistent every time, which leads to inevitable measurement errors. The multi-sphere simultaneous measurement method can ensure the same experimental environment for each sphere, but there will be cross-talking. This paper studies the influence of cross-talking on BSS. Firstly, we use Geant4 to simulate the corresponding neutron response function with and without cross-talking, and analyze the error between the two neutron response functions. Secondly, the neutron count of ▪ neutron source experiment is simulated. Considering that the BSS is close to the floor, the neutron counting error of BS with and without cross-talking and floor scattering is calculated. Finally, we actually measure the ▪ neutron source using a BSS and obtain 8 neutron counts. The spectrum of ▪ neutron source is obtained by calibrating the influence of cross-talking. The results show that the scattered neutrons which affect the count of BS are mainly the primary scattered neutrons, which account for more than 90% of all scattered neutrons. The scattered neutrons mainly affect the neutron response function of the large BS in the high energy domain and the small BS in the low energy domain. The cross-talking has a great influence on the neutron count of large and small BS, while the floor scattering has a great influence only on the small BS. The ▪ neutron spectrum obtained by calibrated neutron counts is more accurate than that obtained by the original experiment counts. Especially in the range of 1∼5 MeV, the calibrated ▪ neutron spectrum is much closer to the international standard spectrum. The mean square error (RMSE) of calibrated ▪ neutron spectrum is reduced by 3.86%. In this paper, by analyzing the influence of cross-talking, the neutron count of BSS is calibrated, and the accuracy of ▪ neutron spectrum is improved, which provides a strong basis for unfolding neutron spectrum.

Comparison of fast-neutron-induced reaction cross-sections in [formula omitted], [formula omitted], [formula omitted], [formula omitted], [formula omitted], [formula omitted], and [formula omitted] in quasi-monoenergetic fields of p+Li and d+TiT neutron sources

This paper describes cross-section measurements of fast-neutron-induced reactions using the neutron activation method. The study employed two accelerator-driven fast neutron sources, each possessing distinct characteristics during irradiation of analogous sets of samples. Specifically, the neutron source at the NPI Řež relied on the p+Li(C) reaction, while MONNET at JRC Geel utilized the d+TiT(Ag) reaction for neutron production. Building upon prior experiments conducted at NPI, we present new cross-sections for reactions, some of which suffer from a dearth of experimental data. The results presented in this paper originate from samples of Al, Au, Bi, Co, NaF, and Y that were irradiated with neutrons at energies of 17.6(3) MeV, and 18.5(4) MeV at MONNET, and an energy of 22.5(8) MeV at the NPI. Following neutron irradiation, the samples were analysed using gamma-ray spectrometry. Cross-sections of several fast-neutron-induced reactions were determined on the basis of the measurement of neutron spectra and corresponding reaction yields. By comparing the results of analogous experiments conducted in different experimental setups, it becomes possible to assess systematic uncertainties. The obtained cross-sections may contribute to the further development of both advanced nuclear reactors and to a more precise fast-neutron dosimetry.

Measurement of neutron spectra in spent fuel storage

The neutron spectrum was measured at two locations in the spent fuel storage facility of the Temelín nuclear power plant. The measurement had two primary objectives: to map the neutron -γ field by quantifying the ambient dose equivalent H∗(10) and to identify methods that could improve the quality of the adjusted neutron spectrum using a Bonner Sphere Spectrometer (BSS). Three spectrometers were used: a BSS and two proton recoil spectrometers. Hydrogen-filled proportional counters and an EJ309 scintillator were used to construct the a priori spectrum for BSS adjustment. The details of this process and its results are discussed. The a posteriori spectrum was used to calculate the ambient dose equivalent H∗(10). The resulting spectrum is highly thermalised, but the predominant contribution to H∗(10) was in the 100 keV-1.3 MeV range. The use of hydrogen-proportional counters in combination with the BSS is recommended.

Production and measurement of a stellar neutron spectrum at 30 keV

A few years ago, we theoretically studied the production of a stellar neutron spectrum at kT = 30 keV using a shaped proton beam impinging on a thick lithium target. Here, we first measure the proton distribution to better control the produced neutron spectrum. Then, we measure the forward-emitted angle-integrated neutron spectrum of the 7Li(p,n)7Be reaction via time-of-flight neutron spectrometry with such proton distribution. The result resembles a stellar neutron spectrum at kT = 30 keV. This method avoids in activation experiments the need for spectrum correction. In the case of spherical samples, no knowledge of the cross-section of the isotope being measured by activation would be necessary. Therefore, the present method is of interest for isotopes with unknown or poorly known cross-sections, such as branching points in astrophysics. The key point of our method is the experimental control of the proton distribution that impinges on the lithium target.

Overview of the neutron diagnostic systems for the SPARC tokamak.

Neutron measurement is the primary tool in the SPARC tokamak for fusion power (Pfus) monitoring, research on the physics of burning plasmas, validation of the neutronics simulation workflows, and providing feedback for machine protection. A demanding target uncertainty (10% for Pfus) and coverage of a wide dynamic range (>8 orders of magnitude going up to 5 × 1019 n/s), coupled with a fast-track timeline for design and deployment, make the development of the SPARC neutron diagnostics challenging. Four subsystems are under design that exploit the high flux of direct DT and DD plasma neutrons emanating from a shielded opening in a midplane diagnostic port. The systems comprise a set of ∼15 flux monitors, mainly ionization chambers and proportional counters for measurement of the neutron yield rate, two independent foil activation systems for measurement of the neutron fluence, a spectrometric radial neutron camera for poloidal profiling of the plasma emissivity, and a high-resolution magnetic proton recoil spectrometer for measurement of the core neutron spectrum. Together, the four systems ensure redundancy of sensors and methods and aim to provide high resolutions of time (10ms), space (∼7cm), and energy (<2% at 14MeV). This paper presents the broader objectives behind the preliminary design of the SPARC neutron diagnostics and discusses the ongoing studies on neutronics, detector comparisons, prototyping, and integration with the unique infrastructure of SPARC. Engineering details of the four subsystems and the concepts for in situ neutron calibration are also highlighted.

Theoretical analysis of neutron spectra measurement with SiC detectors using a machine learning technique

Detecting neutrons is crucial for various nuclear applications. Currently, there exists a technology gap in detecting fast neutrons under high-flux conditions, primarily due to some limitations solid detectors face in supporting high flux. Addressing these challenges requires a balanced approach that combines a detector capable of providing exploitable signals with an effective method for reconstructing the neutron spectrum. This work explores the potential of utilizing machine learning techniques to address this gap by determining neutron spectra using SiC detectors. The machine learning algorithm described in this paper has been tested on measured spectra stored in the IAEA Compendium database. The outcomes of this theoretical study demonstrate the feasibility of the proposed approach in determining neutron spectra from various neutron sources. However, experimental validation is necessary to fully validate the methodology and ensure its applicability in practical settings.

Neutron spectrum measurements inside a self-shielded cyclotron during 18F production using a single-cylindrical neutron spectrometer with gold and indium activation foils

The Single-Cylindrical Neutron Spectrometer (SCNS) is a single moderated spectrometer in the form of a cylinder with a diameter of 8 cm and a length of 16 cm. Seven Au activation foil type neutron detectors are installed on the moderator axis with varying distances. This spectrometer is used to measure the neutron spectrum in self-shielded cyclotron during 18F production using a proton of 11 MeV and a beam current of 25 µA for 70 min. The detector response was calculated using MCNPX 2.7. The measurement results obtained activity of 198Au and 116mIn and were validated with simulation using MCNPX. With the detector response values and detector activity from the measurements, a neutron spectrum was obtained using the unfolding technique using the UMG program. The neutron spectrum from measurements using SCNS with Au and In activation foil detectors has a difference of 8 %. Meanwhile, with the MCNPX calculation results, the measurement results using the Au foil are 1 % while with the In foil it is 10 %.

Bonner sphere measurements of high-energy neutron spectra from a 1 GeV/u 56Fe ion beam on an aluminum target and comparison to spectra obtained by Monte Carlo simulations

The goal of this work is to characterize the secondary neutron spectra produced by 1 GeV/u56Fe beam colliding with a thick cylindric aluminum target and to perform a quantitative comparison with simulated results obtained with Monte Carlo codes. The measurements were performed using extended-range Bonner sphere spectrometers at two positions (15° and 40°) with respect to the beam direction. The secondary radiation field was simulated using four Monte Carlo codes (FLUKA, MCNP6, Geant4 and PHITS) and several physical models of nuclei transport and interaction. Neutron and proton energy distributions were simulated for the experimental measurement positions. The simulated neutron spectra, together with data measured with Bonner sphere spectrometers, after carrying out the correction of the contributions induced by the secondary protons, were used as input for the MAXED spectrum unfolding code to obtain the measured neutron spectra. Unfolded neutron spectra were compared with simulated ones to carry out a quantitative analysis of the performance of the chosen Monte Carlo codes and their corresponding physical models. This comparison showed that, because of experimental uncertainties and physical models, there are no unique solutions for each measurement location, but a range of solutions where the true experimental neutron spectra probably lie. The results showed deviations between 4.23% and 8.42% for some simulated spectra. Regarding the total integral values of neutron fluence and ambient equivalent dose, the unfolded neutron spectra showed deviations lower than 2%.

Determination of neutron spectrum based on artificial neural network using liquid scintillation detector EJ-301.

This paper focuses on the neutron spectrum measurement using a liquid scintillation detector, where the neutron spectrum could be identified and unfolded from the light output distribution of the EJ-301 liquid scintillation detector through a linear artificial neural network (ANN). The response functions of the EJ-301 detector for monoenergetic neutron sources, as well as the light outputs, have been simulated and calculated by Monte Carlo procedure FLUKA. The linear ANN was trained and tested through the simulated data, where response functions were set as the input of ANN and the corresponding neutron spectra were output. Therefore, the neutron spectrum-unfolding model was created. This spectrum-unfolding model was tested through the light outputs induced by monoenergetic neutrons and the random superposition of them. Unfolding results show that this model could identify the information of the neutron spectrum accurately from the light outputs of a liquid scintillation detector. Moreover, the EJ-301 detector was used to measure the radioactivity of 252Cf, and the pulse height distribution induced by neutrons was derived through the charge-comparison method to remove the influence of gamma rays. The measured pulse height distribution was unfolded by the trained model, and measured results show that the unfolded neutron spectrum of 252Cf was consistent with the reference one. This paper presents the feasibility that the unknown neutron spectrum could be identified and confirmed through a linear neural network trained by simulated monoenergetic neutron response functions, which could be a candidate of choice for the determination of the neutron spectrum.

Characterization of neutron spectrum of polyethylene-moderated AmBe neutron source using a passive single-cylindrical neutron spectrometer

The changes in the neutron spectrum produced by the AmBe neutron source moderated in the polyethylene sphere need to be characterized. A passive single-moderator neutron spectrometer with indium foil activation neutron detector has been developed to characterize the neutron spectrum of an AmBe source inside a polyethylene sphere. The detector response function was calculated using the MCNPX 2.7 program with IRDFF-II nuclear data library. Measurements were conducted by placing the Single-Cylindrical Neutron Spectrometer (SCNS) on the outer surface of the sphere for 5 h. The 116mIn activity due to neutron activation in each indium foil was measured using a gamma spectrometer with an HPGe detector. By identifying the count value at the peak energy of 1294 KeV and considering an HPGe detector efficiency of 3.2% at the foil position 1 mm above the detector surface, the activity of 116mIn was obtained. The activity value of 116mIn from each indium foil was compared with the MCNPX simulation results. The neutron spectrum was unfolded using the UMG 3.3 program with activity data input of 116mIn for each foil and detector response. A neutron spectrum was obtained with a total neutron flux of 634 ± 60 n/cm2·s, consisting of 25% thermal neutrons, 16% epithermal neutrons, and 61% fast neutrons. When compared with the simulation results, the total neutron flux in the spectrum produced by SCNS-In showed only a small difference of 1%. Based on these neutron spectrum measurements, it was determined that placing the AmBe neutron source inside a 15″ diameter PE-sphere will reduce the fast neutron flux by 78%.

Neutron spectrum measurements near KAMINI reactor south beam vault door using nested neutron spectrometer.

KAlpakkam MINI reactor (KAMINI) is a 233U fuelled research reactor has various neutron irradiation locations for experimental purposes. The pit at the south beam end of KAMINI reactor is being extensively utilised for neutron attenuation experiments in prospective shielding materials as well as for neutron radiography. During reactor operation, it will be closed by a movable shield. A vault door is located above the shield and the movable shield is used to attenuate streaming neutrons and gamma-rays during reactor operation. Even with the shield, there exists significant dose because of streaming neutrons and gamma rays. Its variation depends on the power of the reactor. The neutron and gamma dose rates close to the south beam vault door have recently been found to be 275-300μSv/h and 175-200μSv/h, respectively, when the reactor is operating at 10kW. In order to characterise the streaming neutron spectra of vault door place for the first time, measurements are done using the Nested Neutron Spectrometer. Along with the neutron flux, neutron mean energy and ambient dose-equivalent rate are also measured and compared with earlier measurements carried out inside the south beam pit. It is observed that the presence of paraffin shield reduces the neutron average energy from 370 to 178keV. Apart from energy reduction, 10kW normalised neutron flux of south beam pit is also attenuated by the shield by 25000 times and it is found that the neutron spectrum of the measured location is also more thermalized. Neutron reference data of the location are generated.

Generation of neutron spectra and emission yields from proton-induced reactions on natC and natLi targets using CR-39 detectors

Fast neutron spectrometry and dosimetry at a specified location in a compact radiation environment is difficult due to large size of the conventional neutron spectrometers. In such situations, a small-size bare CR-39 detector can be used as a neutron detector to quantify the fast neutron fraction. In the present study, CR-39 detectors were placed very close to natural lithium and carbon targets during the proton irradiations to acquire the emission neutron spectra. The neutron spectra and the ambient dose equivalents in the forward and lateral directions with respect to the incident proton beam were estimated for both targets at incident proton energies between 8 and 20 MeV. The measured neutron spectra using CR-39 detectors could identify the quasi mono-energetic neutron features from the Li(p,n) and 13C(p,n) reactions effectively. An important observation of the present study is the identification of the fast neutron signature from the 13C(p,n) system from the discrete state de-excitations of excited 14N composite nuclei. The theoretical evaluation of the neutron spectral features and relative neutron energy distributions were performed using the FLUKA: FLUktuierende KAskade, a Monte Carlo simulation package, and the estimates agreed with the experimental results for both systems. The neutron ambient dose equivalents were also estimated from the measured spectra. These neutron fluence and dose estimates at close vicinity to the target can serve as an essential basis for shielding calculations and planning the pertinent radiation protection strategies.

Spectra measurement of 238Pu-13C source

The precise description of the neutron spectrum is crucial for the measurement of cross sections and well defined integral experiments. Radioisotope sources such as PuC are one of the most widely used neutron sources for different purposes. The neutron spectrum of this source was measured with narrow energy bins for energies greater than 1.3MeV in this experiment with the method of stilbene scintillation spectrometry and the NGA-01 system. The measured neutron spectrum was also compared to simplified simulations in MCNP6 with satisfactory agreement. Effect of the internal structure of the neutron source was tested thoroughly with simulations and is also presented here.

A passive single-cylindrical neutron spectrometer for photoneutron spectrometry in 10 MV Elekta Precise Linac head

Measuring the neutron spectrum on the Linac head using a passive type neutron detector is relatively easier than an active type detector because it does not require the involvement of operators who work in a high radiation area. Neutron spectrum measurements with passive foil activation neutron detectors in a single-moderator are more practical and easier than multi-foil techniques. This technique is quite simple, easy to make, low cost, and the measurement time is relatively short. This spectrometer is designed by placing seven passive Au detectors at varying depths in a single-cylindrical moderator. The response function for each detector is calculated by MCNPX 2.7 program and used ENDF7 as a nuclear data library. Detector foil activity was measured using a gamma spectrometer, and the neutron spectrum was obtained using the unfolding technique using the UMG 3.3 program. The results of measuring the neutron spectrum measurements have been carried out at a position 10 cm below the head of the Elekta Precise 10 MV Linac was 2.73 × 105 n/cm2·s, where the fast neutron from these results are 4% higher when compared to the simulation results using PHITS.

Performance of deconvolution codes (MAXED, GRAVEL, MLEM) for neutron spectrometry of radioactive source using plastic scintillator simulated data

Neutron spectrometry plays a crucial role in decommissioning of nuclear sites and in ensuring the safety of workers from radiation exposure. In this study, we associate a proton recoil spectrometer made by a homemade plastic scintillator with triple discrimination capabilities (fast neutrons, thermal neutrons, and gamma rays) with deconvolution methods to obtain accurate measurements of neutron spectra. The purpose of this work is to test and evaluate the performance of three distinct deconvolution codes, namely MAXED (Maximum Entropy Deconvolution), GRAVEL, and MLEM (Maximum Likelihood Expectation Maximization), using two types of input spectra (flat and Watt spectrum). These deconvolution codes are applied to simulated data using the reference MCNP6.2 Monte Carlo code. Comparing the calculated mean squared error (MSE) performed by the three unfolding methods, we find that MLEM seems to perform better than MAXED and GRAVEL. Furthermore, given the calculated MSE values, the unfolded spectrum of Cf-252 is in better agreement with the standard reference spectrum by using Watt spectrum as input spectrum (MSE<1.2 × 10−6) than using a flat spectrum (MSE<1.3 × 10−6).

Calibration and validation measurements of the Extended-range Bonner sphere spectrometer

An Extended-range Bonner sphere spectrometer (EBSS) has been developed to investigate the neutron spectra of the China initiative Accelerator Driven System. This paper presents the design, calibration, and validation measurements of the EBSS system using a standard 241Am-Be neutron source and the cosmic ray neutrons. The EBSS system was simulated using the PHITS code, and the geometric structures were designed using the response functions obtained by simulations. The EBSS system comprises of seven polyethylene-only spheres and seven extended-range spheres encased in lead, copper, or tungsten shells along with a bare 3He proportional counter. To verify the accuracy of the simulated response functions, the EBSS system was calibrated with monoenergetic neutron beams of 565 keV at the China Institute of Atomic Energy in Beijing. The EBSS system was used to measure neutron count rates of the 241Am-Be neutron source and cosmic ray neutrons. Subsequently, the measured neutron count rates were converted into neutron spectra by the unfolding software. Experimental findings demonstrate the precise capability of the EBSS system to measure neutron spectra across a wide range of neutron radiation fields.

Measurements of AmBe and PuC spectra using EJ309 and stilbene scintillators

Two organic scintillators (EJ309 and stilbene) with their different detection chains and elaborate software were used to measure neutron spectra of AmBe and PuC sources. Method of shadow cone ensures minimization of scattered neutron influence. To separate neutrons and γ, a modern method of pulse shape discrimination was employed using fast digitizers. Calibration was performed by γ sources and three different methods for the identification of the Compton edge. Spectrometric systems also used different light output functions and unfolding methods. The same spectra were calculated using the MCNP6 code and the JENDL-5 library. The measured and calculated results were mutually compared, and also with the published results available in the literature. Generally, the results agree on the main peaks of neutron spectra but differ in their abundance. The lower limit for the ability to distinguish between neutrons and γ in spectrometry with organic scintillators was found at about 1.5 MeV.

Design and validation of a single cylindrical neutron spectrometer using a gold activation foil

A single cylindrical neutron spectrometer (SCNS) with a gold activation foil was designed for mono-directional high-flux neutron spectrum measurement applications. A total of seven gold activation foils with a diameter of 11 mm and a thickness of 0.2 mm were placed in the moderator as detectors with a depth variation of 15 mm. The moderator was cylindrical with a diameter of 8 cm and a length of 16 cm and made of polyethylene. Outside the moderator is a neutron shield made of 3 mm boron-rubber and 0.5-mm cadmium and covered with an aluminum casing of 5 mm thickness. The response value of each detector was determined by simulation using the Monte Carlo MCNPX 2.7 program. The simulated response detectors was then partially validated by comparing the simulation and the measurement result with an AmBe neutron source, where the difference is around 3%. The obtained results showed that SCNS with a gold activation foil could be used to measure the neutron spectrum in a high-flux neutron field.

Measurement of secondary neutron spectra induced by 480 MeV proton and 430 MeV/u 4He beams with a thick aluminum target

Knowledge of the characteristics of secondary neutrons produced by the interaction of Galactic Cosmic Radiation with spacecraft shielding materials is becoming increasingly important for predicting and mitigating biological risks of space explorers during deep-space travel. Hadron accelerators for medical applications are well suited to reproduce part of the conditions found in deep-space in terms of ion species and energies. The objectives of this work are to measure the secondary neutron spectra produced by proton and helium ion beams hitting an aluminum target with energies that correspond to the Galactic Cosmic Radiation peak during solar minimal activity and to validate and compare physical models of Monte Carlo simulations. Neutron spectra were measured with the extended-range Bonner sphere system NEMUS at two positions, 0° and 90° relative to the direction of the primary ion beam. The experimental setup consisted of 480 MeV proton and 430 MeV/u 4He beams colliding with a 30×30×63.5 cm3 aluminum target. The experimental neutron spectra were analyzed using the MAXED unfolding code and compared to several Monte Carlo simulation codes. The results show deviations in terms of the shape of the neutron energy distributions ranging between 1% and 14% and of the integral quantities of fluence and ambient dose equivalent ranging between 1% and 5.2%.

Neutron capture reaction cross-section data processing and resonance parameter analysis of &lt;sup&gt;197&lt;/sup&gt;Au based on white light neutron source

Neutron capture reaction is one of the neutron reactions and plays an important role in using reactor control rods and shell materials, designing nuclear device structures, and studying nuclear astrophysics S processes and element origins. The 4π BaF&lt;sub&gt;2&lt;/sub&gt; detection device has advantages such as high time resolution, low neutron sensitivity, and high detection efficiency, thus making it suitable for measuring neutron radiation capture reaction cross-section data. In order to fill the gap in our neutron capture reaction data in the keV energy range and improve their accuracy, the Key Laboratory of Nuclear Data at the Chinese Institute of Atomic Energy (CIAE) has established a Gamma Total Absorption Facility (GTAF), which consists of 28 hexagonal BaF&lt;sub&gt;2&lt;/sub&gt; crystals and 12 pentagonal BaF&lt;sub&gt;2&lt;/sub&gt; crystals to form a spherical shell with an external diameter of 25 cm and an internal diameter of 10 cm, covering 95.2% of the solid angles. The Back-n beam line of the Chinese Spallation Neutron Source (CSNS) is a back-streaming white beam line that covers neutron energy ranging from a few eV to several hundred MeV, making it suitable for measuring neutron capture cross-sections. The reaction cross-section data of &lt;sup&gt;197&lt;/sup&gt;Au is measured by using GTAF on the Back-n beam line. The measurement data are preliminarily background deducted through energy screening, PSD method, and crystal multiplicity screening. Subsequently, the background is analyzed and deducted based on the measurement data of &lt;sup&gt;nat&lt;/sup&gt;C and empty samples, and the yield of &lt;sup&gt;197&lt;/sup&gt;Au capture reaction is obtained. Resonance parameters are a set of parameters extracted from experimental data to describe the resonance curve, which can eliminate the influence of experimental conditions on resonance data and are more important than the cross-section obtained from experiments. The resonance energy, neutron resonance width, and gamma resonance width parameters of &lt;sup&gt;197&lt;/sup&gt;Au at 1–100 eV are fitted by using the SAMMY program. From the comparison between the resonance curves obtained from experimental measurements and the resonance parameters obtained from fitting with the ENDF/B-VIII.0 database, it can follow that the experimental measurement results are in good agreement with the database, nevertheless, there exist some differences in the resonance parameter, which may be due to the GTAF energy resolution, Back-n neutron spectrum measurement accuracy, and the experimental background deduction method. Our next work is to identify the sources of difference.