Determination Of Neutron Flux Research Articles

Fast neutron induced reaction cross sections on natural manganese and tantalum

The cross sections for the 55Mn(n,2n)54Mn, 181Ta(n,2n)180gTa, and 181Ta(n,p)181Hf reactions were measured to be 705.1 ± 26.1 mb at 14.0 MeV, 1362.7 ± 87.2 mb at 13.6 MeV, and 2.31 ± 0.09 mb at 13.6 MeV, respectively, by using an off-line γ-ray spectroscopic technique. The neutrons were produced via the 3H(d,n)4He reaction. The monitor reactions 27Al(n,α)24Na and 93Nb(n,2n)92mNb were used for neutron flux determination. The results from the present work were compared with those of the literature and the evaluated data from ENDF/B-VIII.0, JEFF-3.3, JENDL-5, CENDL-3.2, and BROND-3.1 libraries. Besides, the cross sections were also estimated with the TALYS-1.96 nuclear model code using different level density models for a better description of the present work and literature data. The present experimental results were found to be in good agreement with most of the available literature data and with the evaluated data.

Computational and experimental determination of slow neutron flux by using Li-6 foil and nuclear track detector

In the present study, slow neutron flux was determined with an experimental method based on 6Li(n, t)4He reaction. 6Li coated foils were mounted on track etch detectors without any gap and they were irradiated with an Am–Be neutron source for 6 h. The track detectors were affixed on a polyethylene plate which has a thickness of 2.0 cm for neutron moderation. Following irradiation, the track detectors were etched, the tritium tracks were counted with an in-house developed track counting system, and the total number of reactions was computed. The difference in particle ranges and track diameters on the Cr-39 detector allowed for the precise distinguishing of tritium, alpha, and neutron particles. The neutron spectrum at the 6Li coated foil position was simulated with MCNP5 using Evaluated Nuclear Data File (ENDF) cross sections of 6Li(n, t)4He reaction. The total reaction numbers obtained with experimental and computational methods are compatible with each other.

Neutron Flux Determination for the NEAR Station at the CERN n_TOF Facility Using the SAND II Unfolding Code

The NEAR station, a new experimental area of the n_TOF facility, was established after the second long shut down of CERN in 2020. This new area was created in order to exploit in situ the high instantaneous neutron flux (originating from proton spallation into a lead target in bunches with momenta reaching up to 20 GeV/c). The reason for the creation of the NEAR station is to utilize the neutrons for experiments related to radiation damage on materials and nuclear astrophysics and the corresponding neutron flux was characterized via extensive neutron multiple foil activation measurements. The irradiated foils were subsequently measured using high purity germanium detectors (HPGe) to determine their induced activities. Finally, the widely used SAND II unfolding code was implemented for the characterization of the neutron flux, using the evaluated cross sections from the IRDFF library, along with the experimentally derived activities. The preliminary results concerning the neutron flux determined in the present work are compared with the corresponding FLUKA simulated ones.

Recoil Proton Telescopes and Parallel Plate Avalanche Counters for the 235U(n,f) cross section measurement relative to H(n,n)H between 10 and 450 MeV neutron energy

With the aim of measuring the 235U(n,f) cross section at the n_TOF facility at CERN over a wide neutron energy range, a detection system consisting of two fission detectors and three detectors for neutron flux determination was realized. The neutron flux detectors are Recoil Proton Telescopes (RPTs), based on scintillators and solid state detectors, conceived to detect recoil protons from the neutron-proton elastic scattering reaction. This system, along with a fission chamber and an array of parallel plate avalanche counters for fission event detection, was installed for the measurement at the n_TOF facility in 2018, at CERN.An overview of the performances of two RPTs — especially developed for this measurement — and of theparallel plate avalanche counters are described in this article. In particular, the characterizationin terms of detection efficiency by Monte Carlo simulations and response to neutron beam, the studyof the background, dead time correction and characterization of the samples, are reported. Theresults of the present investigation show that the performances of these detectors are suitable foraccurate measurements of fission reaction cross sections in the range from 10 to 450 MeV.

Measurement of thermal neutron flux and cadmium ratios using neutron activation analysis

Irradiation of known foil samples in a reactor and subsequent measurement of induced activity allows determination of neutron flux. This is an important information for many purposes. It can be used reciprocally to determine contents of unknown samples irradiated at the same position in the reactor. On the other hand, neutron flux information is used in the assessment of a safe operation of nuclear power reactors. In the frame of this work, bare and cadmium covered foils of gold and dysprosium have been irradiated at different positions of the reactor. Induced activity measurements were made using a standard procedure. Thermal neutron flux and cadmium ratio values were determined for the irradiation positions.

Nuclear Data Sheets for A=64

The evaluated experimental data are presented and evaluated for 13 known nuclides of mass 64 (Ti, V, Cr, Mn, Fe, Co, Ni, Cu, Zn, Ga, Ge, As, Se). For each nuclide, the best values combining all available data are recommended for spectroscopic properties. No excited states have been identified in 64Ti, 64As, and 64Se. Only one excited state in 64V as an isomer, and three in 64Cr have been identified. Data for excited states in 64Mn, 64Fe, and 64Co remain limited. 64Ni, 64Cu, and 64Zn are the most studied nuclides through various reactions and decays, followed by 64Ga and 64Ge. In the opinion of the evaluators, there are several incomplete or discrepant aspects of the high-spin portion of the level scheme for 64Zn above ≈5 MeV excitation which need to be resolved in further experiments using large γ-detector arrays. The decay schemes of 64Ti β−, 64As ε and 64Se ε are unknown, while very little information is available for 64V β− and 64Cr β−. The decay schemes of 64Mn β−, 64Fe β−, 64Co β− and 64Ge ε are somewhat better known, but still considered incomplete by evaluators. The decay scheme of 64Ga ε decay is known in detail, however there is the possibility of additional levels populated above 4713 keV, as the Q(ε) value is 7171 keV. The β−, β+ and ε decay modes of 64Cu are well known, as this radionuclide is of great importance in applications, for example, as a dosimeter for neutron flux determination in nuclear reactors, and as a radiopharmaceutical for positron emission tomography (PET). This work supersedes earlier full evaluations of A=64 by 2007Si04, 1996Si12, 1991Si03, 1979Ha35 and 1974Au04.

Measurement of the 124Xe (n,2n) reaction cross section induced by 14.8 MeV neutron

The measurement of the 124Xe(n, 2n)123Xe reaction cross section at 14.8 MeV neutron has been performed by the activation method. Quasi monoenergetic neutron beams were produced at the Cockcroft Walton Accelerator of China Institute of Atomic Energy, by the T(d, n)4He reaction. A cell made of polymethyl methacrylate (PMMA) was used as the container of 124Xe (99.95% abundance) gas, and the mass of 124Xe in cell was determined by weighing method. Two high-purity thin 93Nb foils were attached to the cell and performed as monitors of the neutron flux determination by 93Nb(n, 2n)92mNb reaction. After the irradiation, the activities of 123Xe and 92mNb were measured by off-line gamma-ray spectrometric technique using a pre-calibrated HPGe detector. As a result, the cross section has been deduced at 14.8 MeV, and the result was compared with the literature data and evaluations.

Determination of Neutron Flux Viewed from the Maxwell-Boltzmann’s Statistical Distribution Using Numerical Approach

This article aims to investigate the computational distribution of neutron flux using Maxwell-Boltzmann statistics. The reactor geometry used is a cylindrical shape. Identify the flux neutron only can when a neutron collides. Data analysis uses numerical analysis with the MATLAB 2013a program. The graph generated from a cylinder geometry is a histogram graph with the Poisson distribution by the Maxwell-Boltzmann distribution. There is a difference in the value of the neutron flux due to statistical uncertainty using a numerical approach. The purpose of writing this article is to determine the computational distribution of neutron fluxes in a cylindrical geometry reactor using Maxwell Boltzmann statistics. By writing this article it is hoped that it will become a reference material for further research or studies on computing the distribution of neutrons in other cases.

Cross section measurement of 241Am(n,f) reaction at the Experimental Area 2 of the n_TOF facility at CERN: First Results

The condition for the safe design and operation of fast neutron reactors and energy boosters (Generation-IV reactors, ADS systems [1]) is the accuracy of nuclear data. The 241Am isotope (T1/2 = 433 years) is highly present in nuclear waste, accounting for about 1.8% of the actinide mass in PWR UOx nuclear reactors’ waste [2]. In addition, the 241Am isotope is further produced by the β decay of the 241Pu isotope (T1/2 = 14.3 years). Given the high production rate of 241Am isotope, its incineration with concurrent energy production is considered to be of utmost importance for the design and implementation of the recycling of existing nuclear waste. Sensitivity studies of the proposed systems for energy production showed that high-precision measurements of the cross section of the 241Am(n,f) reaction are required. In the present work, the 241Am(n,f) reaction cross section was measured in the Second Experimental Area of the n_TOF facility at CERN, using an array of Micromegas detectors. For the measurement, six targets of 241Am with average activity of 17 MBq per sample were coupled with an equal number of detectors in a common chamber. Additionally two 235U and two 238U samples were coupled with Micromegas detectors utilizing the neutron flux determination. Within this work, an overview of the experimental set-up and the adopted data analysis technique is presented along with preliminary results.

Measurement of the 241Am(n,2n) Reaction Cross Section, by the Activation Method

The measurement of the cross section of the reaction 241Am(n,2n)240Am, has been performed, for the first time at neutron energies of 8.8, 9.6, 10.6 and 11.4 MeV, by the activation method. The monoenergetic neutron beam was produced at the 5.5MV TANDEM accelerator of NCSR "Demokritos", using the H{d,n) He reaction. During the 5-day long irradiation, the neutron flux was monitored by a BF3, counter. The neutron yield as well as the beam current, were recorded in 100 sec intervals by two multiscaling units. The radioactive target consisted of a 37GBq 241Am source enclosed in a Pb container. A natural Au foil, 27Al foil and a 93Nb foil were used as reference materials for the neutron flux determination. The activity of the irradiated targets, was measured off-line by a 56% relative efficiency, HPGe detector. The first preliminary results are presented and compared with evaluated libraries and existing data.

Investigation of neutron flux by thermoluminescence dosimeters in the neutron-gamma mixed field

Introduction: Neutrons have many applications in various fields, such as medicine. The important application of neutron in medical science is in Boron Neutron Capture Therapy (BNCT). Because of this, determination of neutron dose and flux is critical for the health maintaining of workers and patients exposed to this beam. The neutron radioisotope sources produce gamma rays in addition to the neutrons, and the neutron fields around these sources are mixed-field neutron-gamma fields. Since the separation of the dose of each radiation component in the compound field is difficult, the gamma-neutron mixed field dosimetry is a major challenge. International Commission on Radiation and Units (ICRU) recommends the use of a dosimeter pair for mixed-field dosimetry that have different responses to each component of radiation. One of the effective methods of neutron dosimetry is the use of the TLD-600/700 dosimeter pair. TLD-600 is 6Li enriched and TLD-700 is 7Li enriched. These dosimeters have the same sensitivity to gamma radiations but TLD-600 is sensitive to thermal neutrons because of high capture cross-section of 6Li, while in low fluxes, TLD-700 has no sensitivity to thermal neutrons. Materials and Methods: In this paper, in order to measure the neutron flux of the Ra-Be source in the core laboratory of the Faculty of Science of the University of Guilan. Six TLD-600/700 dosimeters were used in three intervals of 7, 14 and 20 cm from the source Results: After 24 hours of irradiation, the intensity of the thermoluminescence of TLDs was measured by TLD reader. Finally, by calculating the calibration coefficient, the measured flux values by the TLDs were determined and compared with the calculated fluxes Conclusion: Results show that the measured and calculated values are in good agreement.

Determination of the flux of a neutron radioisotope source through gold activation method

Introduction: In recent years, the use of neutrons has increased in various fields such as medicine and industry. One of the important applications of neutrons is in medical research. Neutrons have greater linear energy transfer than gamma rays and are considered more dangerous radiation. Hence, the detection and determination of neutron dose is a very important and vital issue. For determination of neutron dose, measurement of its intensity or flux is essential. One of the ways of neutron flux determination is using of gold foil activation method. In this method, the activity of the irradiated foil is proportional to the neutron flux. The gold foil is activated by reaction 197Au (n,γ) 198Au because of the thermal neutron capture,. In the decay of activated gold, a gamma radiation of 411.8 keV is released with a probability of more than 95%, that By measuring the area below the peak associated with this energy, by the NaI(Tl) detector, the activity of gold and then the neutron flux is determined. Materials and Methods: In this paper, an 18 carat gold foil was used to measure the flux of a neutron radioisotope source Ra-Be in the nuclear laboratory of the Faculty of Science of Guilan University. Results: After irradiation, the gold foil was counted by NaI(Tl) detector and its flux was determined Conclusion: Finally, by comparing the measured flux and the flux in the Ra-Be neutron source catalogue, it is resulted that the mentioned method has a good accuracy in determination of the flux of the neutron source.

Thermal and epithermal neutron flux distributions measurement in thermal column of TRR using an experimental-simulation method

For designing an appropriate neutron beam, the determination of neutron flux at any irradiation facility is an important key factor. Due to the importance of determining the thermal and epithermal neutron fluxes in a typical thermal column of a reactor, a simple and accurate technique is introduced in this study. Absolute thermal and epithermal fluxes were measured experimentally at a certain point using the foil activation method by neutron bombardment of bare and cadmium covered Au foils. The relative neutron fluxes were also derived simply by means of Monte Carlo simulation by accurate modelling of the reactor components. Finally, by normalization of the relative distribution flux with regard to information about the absolute neutron flux, the accurate thermal and epithermal neutron distributions were derived, separately.

DETERMINATION OF THE BACKGROUND OF 3He-FILLED PROPORTIONAL COUNTERS USED FOR LOW-LEVEL NEUTRON MEASUREMENTS

Low-level neutron measurements are required for applications such as environmental monitoring, measurements of weak sources and the determination of neutron fluxes at underground laboratories. When analyzing low-level neutron measurements, it is important to be able to distinguish the signal due to neutrons from any background term that may originate within the detector. A solution to this problem has been developed for the case of measurements carried out with 3He spherical proportional counters of the type which are often used in Bonner sphere spectrometers. To determine this background, measurements were carried out in the former UDO underground laboratory of the Physikalisch-Technische Bundesanstalt located in the Asse salt mine. The analysis of the data was carried out using Bayesian parameter estimation. The result of the analysis is a very general parameterised function that can be used to describe the pulse height spectrum due to the background of the proportional counters.

Determination of Reactivity and Neutron Flux Using Modified Neural Network for HTGR

Nuclear kinetic calculations based on point kinetic model have been generally applied as the standard method for neutronics codes. As the central control rod (C-CR) withdrawal test has demonstrated in a prismatic core type high-temperature gas-cooled reactor (HTGR) named High Temperature Engineering Test Reactor (HTTR), the transient calculation of kinetic parameter, reactivity, and neutron fluxes, requires a new method to shorten calculation-process time. Development of neural network method was applied to point kinetic model as the necessity of real-time calculation that could work in parallel with the digital reactivity meter. The combination of Time Delayed Neural Network ( TDNN ) and Jordan Recurrent Neural Network ( Jordan RNN ) named TD-Jordan RNN was the result of the modeling approach. The application of TD-Jordan RNN with adequate learning, tested offline, determined results accurately even when signal inputs were noisy. Furthermore, the preprocessing for neural network input utilized noise reduction as one of the equations to transform two of twelve time-delayed inputs into power corrected inputs.

Determination of 1-keV to 1-MeV Neutron Flux by Radiative Capture (n, $\gamma $ ) on92,94 Zr Dosimeters

The purpose of this paper is to demonstrate 92,94Zr application in neutron reactor dosimetry. We aim to detect neutron flux in 1-keV to 1-MeV energy range by irradiating 92,94Zr samples under boron nitride (BN) filter. The presence of BN is needed in order to cut off thermal neutron contribution. This paper describes Zr irradiation and experimental results that were obtained by $\gamma $ spectrometry and accelerator mass spectrometry. In parallel, Zr irradiation is modeled with Monte Carlo-based code for neutron transport—TRIPOLI-4v10 developed by CEA (Saclay). Experimental values of 95Zr activity and 93Zr/92Zr ratio are compared to calculation results. Since crosssection values are necessary for the neutron flux determination, we present here the comparison between several nuclear data libraries for radiative capture on 92,94Zr. Uncertainties for calculation/experience (C/E) ratios are obtained combining both experimental and nuclear data ones. This paper shows the feasibility of 1-keV to 1-MeV neutron detection by using 92,94Zr (n, $\gamma $ ) reaction and introduces a postirradiation analysis for reactor dosimetry based on accelerator mass spectrometry. It also reveals that more precise evaluation of 92,94Zr (n, $\gamma $ ) cross section is required in order to improve C/E values.

Characterization of a 6 MeV Accelerator Driven Mixed Neutron/Photon Source

There are many applications which require high yield radiation sources with mixed fluxes of photons and neutrons. In particular, such sources are necessary to test radiation detectors and materials. This study was concerned with the determination of photon and neutron fluxes generated by the interaction of a 6 MeV linear electron accelerator driven photon beam with a beryllium photoneutron converter. The double step procedure of an (e,γ) reaction followed by an (γ,n) emission results in a mixed radiation environment. The optimal converter geometry was determined by comparison of the computed neutron fluxes for each converter position. Computational results have shown that photon fluxes up to 1011 photons/cm2/s and neutron fluxes up to 107 neutrons/cm2/s are achievable with the optimal setup. This paper is focused on the results of the MCNPX modeling and experiments and discussion of the converter orientation which leads to the largest radiation fluxes.

Determination of the Secondary Neutron Flux at the Massive Natural Uranium Spallation Target

The flux of secondary neutrons generated in collisions of the 660 MeV proton beam with the massive natural uranium spallation target was investigated using a set of monoisotopic threshold activation detectors. Sandwiches made of thin high-purity Al, Co, Au, and Bi metal foils were installed in different positions across the whole spallation target. The gamma-ray activity of products of (n,xn) and other studied reactions was measured offline with germanium semiconductor detectors. Reaction yields of radionuclides with half-life exceeding 100min and with effective neutron energy thresholds between 3.6 MeV and 186 MeV provided us with information about the spectrum of spallation neutrons in this energy region and beyond. The experimental neutron flux was determined using the measured reaction yields and cross-sections calculated with the TALYS 1.8 nuclear reaction program and INCL4-ABLA event generator of MCNP6. Neutron spectra in the region of activation sandwiches were also modeled with the radiation transport code MCNPX 2.7. Neutron flux based on excitation functions from TALYS provides a reasonable description of the neutron spectrum inside the spallation target and is in good agreement with Monte-Carlo predictions. The experimental flux that uses INCL4 cross-sections rather underestimates the modeled spectrum in the whole region of interest, but the agreement within few standard deviations was reached as well. The paper summarizes basic principles of the method for determining the spectrum of high-energy neutrons without employing the spectral adjustment routines and points out to the need for model improvements and precise cross-section measurements.

Presence of neutrons in the low-level background environment estimated by the analysis of the 595.8 keV gamma peak

In order to explore possible improvements of the existing techniques developed to estimate the neutron fluence in low-background Ge-spectroscopy systems, gamma spectra were collected by a HPGe detector in the presence of the 252Cf spontaneous fission neutron source. The spectra were taken with and without a Cd envelope on the detector dipstick, with different thicknesses of plastic used to slow down neutrons. We have analyzed the complex 595.8keV gamma peak, as well as several more gamma peaks following the neutron interactions in the detector itself and surroundings materials. The investigation shows that some changes of the initial neutron spectra can be monitored by the analysis of the 595.8keV gamma peak. We have found good agreement in the intensity changes between the long-tail component of the 595.8keV and the 691keV gamma peak (72Ge(n,n′)72Ge reaction), usually used for the estimation of the fast neutron fluence. Results also suggest that the thermal neutrons can have a stronger influence on creation of the Gaussian-like part of 595.8keV peak, than on the 139keV one following 74Ge(n,γ)75mGe reaction and used in the standard methods (Škoro et al., 1992) [8] for determination of the thermal neutron flux.

The influence of field size and off-axis distance on photoneutron spectra of the 18 MV Siemens Oncor linear accelerator beam

At present, high energy electron linear accelerators (LINACs) producing photons with energies higher than 10 MeV have a wide use in radiotherapy (RT). However, in these beams fast neutrons could be generated, which results in undesired contamination of the therapeutic beams. These neutrons affect the shielding requirements in RT rooms and also increase the out-of-field radiation dose to patients. The neutron flux becomes even more important when high numbers of monitor units are used, as in the intensity modulated radiotherapy. Herein, to evaluate the exposure of patients and medical personnel, it is important to determine the full radiation field correctly. A model of the dual photon beam medical LINAC, Siemens ONCOR, used at the University Hospital Centre of Osijek was built using the MCNP611 code. We tuned the model according to measured photon percentage depth dose curves and profiles. Only 18 MV photon beams were modeled. The dependence of neutron dose equivalent and energy spectrum on field size and off-axis distance in the patient plane was analyzed. The neutron source strength (Q) defined as a number of neutrons coming from the head of the treatment unit per x-ray dose (Gy) delivered at the isocenter was calculated and found to be 1.12 × 1012 neutrons per photon Gy at isocenter. The simulation showed that the neutron flux increases with increasing field size but field size has almost no effect on the shape of neutron dose profiles. The calculated neutron dose equivalent of different field sizes was between 1 and 3 mSv per photon Gy at isocenter. The mean energy changed from 0.21 MeV to 0.63 MeV with collimator opening from 0 × 0 cm2 to 40 × 40 cm2. At the 50 cm off-axis the change was less pronounced. According to the results, it is reasonable to conclude that the neutron dose equivalent to the patient is proportional to the photon beam-on time as suggested before. Since the beam-on time is much higher when advanced radiotherapy techniques are used to fulfill high conformity demands, this makes the neutron flux determination even more important. We also showed that the neutron energy in the patient plane significantly changes with field size. This can introduce significant uncertainty in dosimetry of neutrons due to strong dependence of the neutron detector response on the neutron energy in the interval 0.1–5 MeV.