Thermal Neutron Flux Research Articles

Conceptual Design of a Fast Periodically Pulsed Reactor

The IBR-2M pulsed nuclear reactor operating at the JINR Laboratory of Neutron Physics delivers an intense neutron flux on the order of 1013 n/(cm2 s). By the end of the 2030s, it will outlive its useful lifetime and will have to be replaced with a new machine. The alternative neutron sources proposed so far require extensive and costly R&D efforts, and positive results cannot be guaranteed. In this paper, we present a conceptual design of the IBR-4 reactor running on plutonium-dioxide fuel which is aimed to replace the IBR‑2M machine as a neutron source. This design fully builds on proven technical solutions tested at other nuclear facilities and, therefore, requires no extensive R&D and guarantees reliable and safe operation of the proposed neutron source. Running at a 5-MW power and a pulse repetition rate of 5 Hz, IBR-4 will deliver a thermal-neutron flux of 5.8 × 1013 cm–2 s–1. When running at a 10-MW power and a pulse repetition rate of 10 Hz, the proposed source is expected to deliver a record-high thermal-neutron flux on the order of 1.1 × 1014 cm–2 s–1.

The comparison of five proton linac beams in order to set up a thermal neutron radiography installation

Radiography without doubt is maybe the most famous non-destructive technique with many applications in a number of scientific fields. In the present study, the performance of five photo-neutrons which delivered by proton linear accelerators with energy equal with 8, 12, 16, 20 and 24 MeV for a thermal neutron radiography unit has been designed and simulated with the help of the MCNPX 2.5.0 software which based on the Monte Carlo statistical method. The geometrical arrangement of the unit has been simulated with primary goal to provide the highest thermal neutron flux in the location of the examined item. For every energy of the proton beam all the significant parameters with the thermal neutron radiography theory like the flux of the thermals neutrons, the (collimator length)/(aperture diameter) ratio and the percentage of the thermal neutron within the neutron beam simulated for a wide series of values. Owing to the fact that the there are many fast neutrons inside the beam, it is obligatory to decrease the quantity of the fast neutrons. For this reason, the addition of a single sapphire for fast neutron filtering has as a consequence a neutron beam with better quality. According to the simulations, the presence of the filter improves the % presence of the thermal neutrons with no substantial abatement in the flux of the slow neutrons.

VALIDATION OF THERMAL SCATTERING LAWS FOR LIGHT WATER AT ELEVATED TEMPERATURES WITH DIFFUSION EXPERIMENTS

Light water is the most important neutron moderator in many reactor applications. Thermal neutron scattering kernels for H bound in H2O are represented by ENDF-format thermal scattering laws (TSLs) evaluated at discrete temperatures T. Nuclear data evaluations are conventionally validated utilizing a combination of critical benchmarks and experimental cross section data. Existing public critical benchmarks may be inadequate for testing water TSLs over the full range of T of interest in reactor applications, and experimental thermal scattering cross section data for water is sparse at elevated T. In this work, MC21 is used to simulate the decay of the equilibrium thermal neutron flux in light-water spheres of several radii. The fundamental-mode time eigenvalue α is calculated at 22 °C and 227 °C (at saturation pressure) for each sphere using three different H-H2O TSLs. Fitting α to a polynomial function of geometric buckling allows calculation of the thermal neutron diffusion length L. Extrapolation lengths are treated as a function of the transport mean free path and geometry. Experimental L data from 25 publications at 49 temperatures (from 10 °C to 295 °C), computed by several different time-dependent and space-dependent decay methods, is used to develop an empirical fit of L vs. T. The MC21-calculated L for the TSLs tested are within ±1% of the predicted values at 22 °C and 227 °C. This validation approach, which may be repeated at arbitrary T, constitutes an integral benchmark specific to and characterizing the detailed physics of the thermal scattering kernel applied.

Initial trials of a dose monitoring detector for boron neutron capture therapy

In this study, we present results of initial trials of a scintillator-over-fiber detector system with a silicon photomultiplier readout designed at Budker Institute of Nuclear Physics. The results demonstrate that the proposed system, using a pair of boron-enriched and boron-free scintillators, could be successfully used for monitoring of thermal neutron flux and estimation of irradiation dose. Nevertheless, it is necessary to optimize the design of the detector head components to ensure long time detector stability with heavy radiation background.

High resolution measurements with miniature neutron scintillators in the SUR-100 zero power reactor

Three 1-mm3 miniature fiber-coupled scintillators have been used to perform cm-wise resolution measurements of the thermal neutron flux within experimental channels of the SUR-100 facility, a zero power thermal reactor operated by the Institute of Nuclear Technology and Energy Systems at the University of Stuttgart. The detection system is developed at the École Polytechnique Fédérale de Lausanne in collaboration with the Paul Scherrer Institut. Thermal neutrons count rates were measured along the experimental channels I and II, which cross the reactor at the center and tangentially to the core, respectively. The reactor was modelled with the Monte Carlo neutron transport code Serpent-2.1.31. The comparison of experimental and computed reaction rate distributions showed a good agreement within the core region, with discrepancies within 2σ. An unexpected discrepancy, probably caused by a geometric inconsistency in the computational model of the reactor, was observed in the reflector region of the experimental channel I, where a 20% difference (i.e. 8σ) was found between experimental and simulated results. Significant discrepancies, respectively worth 10σ and 15σ, were noticed at distance, in the lead shielding region, for both experimental channels I and II. In addition, reaction rate gradients across the 2.6 cm and 5.4 cm diameters of both channels were measured. A horizontal reaction rate gradient of (9.09 ± 0.20) % was measured within 2.4 cm across the diameter of the experimental channel II, with a difference from computed results of 2%. The absence of a vertical reaction rate gradient inside the experimental channel I was confirmed by measurements.

AXIAL FLUX PROFILE IN THE ADVANCED TEST REACTOR

This work demonstrates an approach to determine probability of perturbation of the axial profile of the thermal neutron flux in the Advanced Test Reactor. The axial flux profile is expected to follow a theoretical cosine shape, due to the minimal use of vertically-withdrawn shims. Reactivity is normally controlled by rotation of Outer Shim Control Cylinders, uniformly affecting neutron flux at all axial locations. The Advanced Test Reactor routinely accepts for irradiation experiments of a variety of designs. Among the analyses required by the safety basis approved by the United States Department of Energy is the characterization of a new experiment’s potential for perturbing the axial flux, which could exacerbate power peaking in the driver fuel. However, this perturbation can be more or less severe in different locations within the fuel. Therefore, the best characterization of axial flux perturbation requires knowledge of baseline axial flux. Such information is obtained by measuring decay in activated uranium flux wires irradiated at known positions in cooling channels in plate-type fuel elements. Due to variability in measured axial flux, it is not usually clear whether a given anomalous measurement is caused by an actual perturbation. Assuming normality in random measurement errors, the probability of an actual perturbation is quantified.

DESIGN OF NEUTRONIC PARAMETERS OF MTR REACTOR USING WIMSD-5B/BATAN-FUEL CODES

BATAN has three aging research reactors, so it is necessary to design a new, more modern MTR type reactor using high-density, low enrichment uranium molybdenum fuel. The thermal neutron flux at the irradiation position is an important concern in the design of research reactors. This analysis is performed using standard computer codes WIMSD-5B and Batan-FUEL. The purpose of this study is to analyze the effect of the core configuration with safety control rods and neutronic parameters using the diffusion method calculation. The reactor core consists of 16 fuel elements and four control rods placed in the 5 x 5 position of the grid plate and is loaded the reflector elements outside the core. The cycle length is also a concern, not less than 20 days, and the reactor can be operated safely with a power of 50 MW. The calculation results show that for the highest fuel loading, which is 450 grams of U7Mo/Al fuel with D2O as a reflector, it will provide the lowest thermal neutron flux at the center of the core irradiation position, namely 1.0 x1015 n/cm2s. The core fuel cycle length will be up to 39 days, meeting the expected acceptance and safety criteria.

CALCULATION OF THE POWER DIAGRAM OF AN EXPERIMENTAL DEVICE WITH A NEUTRON CONVERTER

The article presents a method for calculating the power of the elements of an experimental device, the main purpose of which is to convert the thermal neutron flux of the IGR. This technique is an improvement of various calculation methods for obtaining data on energy release in structural elements of tested devices at the stage of planning experimental studies.

Anisotropy in the hardness of single crystal tungsten before and after neutron irradiation

In this work, the hardness of single crystal tungsten (W) after low, medium and high temperature neutron irradiation, reaching up to 1200°C, is studied. Micro-hardness tests with Vickers pyramid indenter are performed on reference and neutron irradiated samples by varying the orientation of the indenter on the single crystal tungsten contact surface. Due to the availability of multiple slip systems, the anisotropy of the hardness is characterized to allow the construction of uncertainty bands as well as dependence of the absolute value of the hardness on the orientation of the indenter wedges and principal crystallographic axes of the crystal. The applied neutron fluence and flux is representative for the ITER reactor, where tungsten is selected as armour material for plasma facing components. The neutron irradiation is performed in the BR2 reactor with mixed neutron spectrum up to 1 dpa (end of life fluence for ITER) but with appropriate measures to reduce the thermal neutron flux, such that the spurious transmutation into rhenium is reduced. The hardness study is accompanied with a detailed microscale examination to deduce the true indenter surface area. As a result, the true values of the hardness accounting for the pile-up effects are obtained as a function rotation angle of the indenter. The variation of the hardness as a function of indenter orientation before and after irradiation is discussed based on the indenter-slip system geometry. A simple but efficient methodology is proposed to deduce the true hardness and its angle-dependent variation for single crystal tungsten.

Assessment of a polymeric composite as a radiation attenuator and a restoration mortar for cracking in biological shields

This work is dedicated to figuring out robust epoxy/magnetite/boron carbide (EP/Mag/B4C) composite for radiation attenuation at multiple applications related to nuclear installations, as well as restoration mortar for cracking developed in concrete biological shields. The mechanical properties (flexural, compressive, and impact strengths) and the physical properties (water absorption, porosity, and dry bulk density), each, have been performed to label the composite integrity for practical application. In practice, attenuation properties have been performed by using a collimated beam emitted from spontaneous fission 252Cf (100 μg) neutron source and neutron gamma spectrometer with stilbene scintillator. The pulse shape discrimination technique which would come of the zero cross over method was used to measure the fast neutron and gamma-ray spectra. Thermal neutron fluxes have been measured by using the thermal neutron detection system and the BF-3 detector. The attenuation parameters: precisely, macroscopic effective removal cross-sections ΣR (cm-1), macroscopic cross-sections Σ (cm-1), and total attenuation coefficients μ (cm-1) of fast and thermal neutrons and total gamma-rays respectively were evaluated using the attenuation relations. Also, the MCNP5 code and MERCSF-N program have been used to compute the parameters theoretically. When applicable, measured and calculated results were compared, and it tells us a comprehensive agreement.

High count rate pulse shape discrimination algorithms for neutron scattering facilities

The performance of EJ-270, a 6Li loaded pulse shape discriminating plastic scintillator was tested for use in applications with high thermal and epithermal neutron fluxes such as neutron scattering facilities. The short decay time of EJ-270 make it of interest for high count rate applications. To realize this, 4 PSD algorithms were tested. The algorithms were selected based on the potential for high rate applications and simplicity. These algorithms were the charge integration method with and without the addition of a digital low pass filter, a measurement of the time to 10% of peak amplitude and a method we call the “tail sum” which utilizes a digital low pass filter and sums a small number of samples in the tail of each pulse. These algorithms were benchmarked using the figure of merit, the γ-sensitivity and potential rate capability. The charge integration method gave the highest figure of merit of 1.37 using a long window of 272.5 ns but had a γ-sensitivity of 2×10−6 which was poorer than the tail sum algorithm. The tail sum algorithm was able to achieve a figure of merit of 1.36 with a window of 250 ns and a γ-sensitivity on the order of 10−7. Reducing the integration windows to match the fastest algorithm of time to 10% resulted in the tail sum outperforming the other algorithms with a figure of merit of 1.26 and a γ-sensitivity of 6×10−7. The short charge integration method and tail-sum were compared on the EMMA beamline at the ISIS pulsed neutron and muon source. The tails sum demonstrated better separation between the γ-rays and thermal neutron at an incident peak neutron rate of 9.6×105 neutrons per second.

Analysis of lutetium-177 production at the WWR-K research reactor

Production of lutetium-177 using direct nuclear reaction 176Lu(n,γ)177Lu by WWR-K reactor neutrons on enriched LuCl3 (up to 82% of 176Lu) is described. Calculations were performed by MCNP6 transport code. Two different irradiation positions of the WWR-K research reactor were considered. Estimates of the maximum specific activity of the luthetium-177 are obtained for the reactor irradiation positions located: (a) in the reactor core centre, (b) in the core periphery. In these positions, thermal neutron flux is two times different.Experimental data was shown that k-factor is 1.5 for considered irradiation positions. The study shows that for the position located in the core center, the estimated maximum specific activity of lutetium-177 is 819 GBq/mg, is to be achieved after 15 days of irradiation. For the position located in the core periphery, specific activity of lutetium-177 is 561 GBq/mg, is to be achieved after 20 days of irradiation. Ratio of Lu-177m to Lu-177 specific activity is not more than 0.025 for both irradiation positions.

First experimental verification of the neutron field of Nagoya University Accelerator-driven neutron source for boron neutron capture therapy

As the commissioning phase of the Nagoya University Accelerator-driven Neutron Source for boron neutron capture therapy, in-phantom thermal neutron flux measurements were conducted using a small ⁶LiF/Eu:CaF₂ scintillator detector and activation foils. The spatial distribution of the measured thermal neutron flux agreed with the Monte Carlo simulation results. Based on this comparison, the free-in-air neutron spectrum, calculated at the exit aperture, was verified and the epithermal neutron flux, at a 0.25 mA proton current, was evaluated as (1.49 ± 0.10) × 10⁷ n/(cm2 s).

A NMR and SANS study of alkali-borosilicate behaviour under thermal neutron irradiation

The internal structure of several alkali-borosilicate glasses has been studied when exposed to a high thermal neutron flux. More specifically, the different glasses we are interested in are widely used around the globe for neutron guide manufacturing and show drastically different resistance to irradiation. Samples were irradiated with the thermal flux from T4 experimental tube in ILL’s High Flux Reactor up to a fluence of several 1017 n/cm2. The experimental tools employed for the structural analysis were Nuclear Magnetic Resonance (NMR) and Small Angle Neutron Scattering (SANS). NMR measurements allowed to detect the modification of atoms’ close environment and these results have been correlated to evolutions at macroscopic scale. In addition, neutron scattering measurements have demonstrated the possibility to detect phase segregation in a zinc-alkali-borosilicate.

Reactor studies of tritium release from lead-lithium eutectic Li15.7Pb with deuterium over the sample

Abstract At the moment, the lead–lithium eutectic was chosen as a breeding blankets material for a DEMO-type fusion reactor. During the reactor operation radioactive tritium will be produced in the blanket, but the interaction parameters of hydrogen isotopes with lead–lithium eutectic are poorly studied. The most significant processes affecting the release of tritium from the lead–lithium eutectic is its interaction with lithium atoms, as a result of which lithium tritide is formed. The main published studies of eutectics were usually carried out with the samples containing 17% of lithium, but even a small deviation in the lithium–lead concentration will lead to significant changes in the properties of the eutectic with respect to hydrogen isotopes. That is why there is an urgent need to research the generation and release of tritium from a lead–lithium eutectic with a lithium content of 15.7% directly under neutron irradiation. This paper presents the results of experimental studies of the lead–lithium eutectic with 15.7% Li content under neutron irradiation. The investigations were carried out at the IVG1.M research reactor (Kurchatov, Kazakhstan) at thermal power of 6 MW (thermal neutron flux ~1014 n/(cm2·s)). The experiments consisted in following: the deuterium flux (~10−9 mole/s) was supplied into the experimental ampoule with eutectic under irradiation and the release of DT-molecules was registered under various sample’s temperatures. The obtained experimental data allowed us to estimate the parameters of tritium release from Li15.7Pb.

Development of a Neutron Radiography System based on a 10 MeV Electron Linac

A thermal neutron radiography unit using the neutrons which emits a 10 MeV electron linac compact has been designed and simulated via MCNPX Monte Carlo code. The facility was carried out for an extensive range of values for the collimator ratio L/D, the main parameter which describes the quality of the produced radiographic images. The results show that the presented facility provides high thermal neutron flux; while with the use of single sapphire filter fulfills all the suggested values which characterize a high quality thermal neutron radiography system. A comparison with other similar facilities indicates that the use of a photoneutron source using a 10 MeV electrons beam is a useful substitutional for radiographic purposes.

Study of Forbush decreases and lunar cycles effects in the thermal neutron flux registered near the Earth surface by means of “Neutron” setup data

The experimental data on thermal neutron flux registered by the “Neutron” setup (MEPhI, Moscow) on the Earth's surface for the period from May 2015 to February 2019 were analyzed. The effects of Forbush decrease (FD) and the lunar cycles were studied. The comparison with the results of FD studies in data of two other setups: the Moscow neutron monitor (MNM) and the muon hodoscope URAGAN (MEPhI, Moscow) was made. Comparison showed that the FD amplitudes of “Neutron” are comparable with those of MNM (on average, less by about 30%), and about 1.5 times more than for MH URAGAN. The counting rate recovery of “Neutron” detectors is much faster than for MNM and MH URAGAN.

Pattern Recognition–Based Technique for Control Rod Position Identification in Pressurized Water Reactors

This paper proposes the application of a pattern recognition–based technique to enhance the process of control rod position identification in pressurized water reactors (PWRs). The proposed technique employs a multivariant analysis technique, namely, principal component analysis (PCA) and clustering analysis (CA) to identify the position of the PWR control rod using its impact on the core radial thermal neutron flux along the axial track of motion. The results of these investigations have shown that the proposed technique successfully removed the limitation on the data size and any limitations imposed by outlier samples, extracted the noise, and provided near-instantaneous analytical and visual ways for position identification process with excellent generalization fitting and prediction efficiencies. In the context of this paper, multiple in-depth simulations are conducted to ascertain the efficiency of the proposed technique in identifying the control rod positions. These simulations have been conducted using a Westinghouse 2772-MW(thermal) PWR benchmark at 100% thermal power generation, where a three-dimensional TRITON FORTRAN-code has been utilized to simulate the radial thermal neutron flux of the PWR core. The PCA model is developed, tested, and generalized using the SIMCA software package. In addition, CA is also performed via the Minitab statistics software package in order to confirm the efficiency of the proposed technique.

Neutron irradiation hardening across ITER diverter tungsten armor

In this work, we have performed neutron irradiation and sub-sequent hardness measurements on a series of tungsten grades to screen the irradiation-induced hardness as a function of irradiation temperature reaching up to 1200 °C. The selected irradiation temperatures were chosen by performing temperature analysis of the expected irradiation temperature on tungsten monoblock during the steady state operation in ITER, where 1200 °C corresponds to the surface temperature at 10 MW/m2 flux density expected during normal operational conditions. The applied neutron fluence and flux (using BR2 material test reactor, up to 1 dpa) is representative of ITER irradiation conditions except the neutron spectrum. However, the measures were taken to reduce the thermal neutron flux to limit the transmutation closer to the fusion conditions. The irradiation-induced hardness measured in single crystal after irradiation at 600–800 °C agrees very well with the earlier data reported after HFIR irradiation experiments. The new irradiation data obtained in the temperature range 900–1200 °C show that even at one third melting point the neutron exposure raises the hardness by 40% to 70%, depending on the selected grade. Screening measurements by transmission electron microscopy, applied to clarify the origin of the hardening at 1200 °C, have proven the presence of the dislocation loops and high density of voids. The presence of those defects should imply the reduction of thermal conductivity, fracture toughness as well as alteration of hydrogen isotope permeation and trapping.

Neutron background measurement for rare event search experiments in the YangYang underground laboratory

Several experiments have been conducted in the YangYang Underground Laboratory in the Republic of Korea such as the search for dark matter and the search for neutrinoless double beta decay, which require an extremely low background event rate due to the detector system and the environment. In underground experiments, neutrons have been identified as one of the background sources. The neutron flux in the experimental site needs to be determined to design a proper shielding system and for precise background estimation. We measured the neutron spectrum with a Bonner sphere spectrometer, with Helium-3 (3He) proportional counters. The neutron flux at the underground laboratory was so low that the radioactive decays from the radioisotopes contained in the detector created a significant background interference to the neutron measurement. Using Monte Carlo simulations, the intrinsic α background distribution due to the radioactive isotopes in the detector materials, was estimated. The neutron count rate of each Bonner sphere was measured from the pulse height spectrum of the 3He proportional counter, after subtracting the α particle background. The neutron flux and the energy spectrum were determined using the unfolding technique. The total neutron flux measured was (4.46 ± 0.66) × 10−5 cm−2 s−1, and the thermal and fast neutron flux (in the range 1–10 MeV) were (1.44 ± 0.15) × 10−5 cm−2 s−1 and (0.71 ± 0.10) × 10−5 cm−2 s−1, respectively.