Neutron Interaction Cross Sections Research Articles

Influence of neutron cross-section resonances on organ/tissue equivalent and effective dose coefficients for the ICRP voxel phantoms.

The materials which compose the ICRP Voxel phantoms used in the computation of conversion coefficients involve neutron interaction cross-sections that have resonances at specific energies. Depending on the energy bin structure used in the computations, these cross-section resonances may occur at energies that fall between energies at which dose coefficients are computed, thus their effects may not be completely accounted for in the reported coefficients. In the present study, a highly refined energy grid that closely follows the resonance structure in the phantom material cross-sections was identified and used to calculate dose coefficients. Both the equivalent organ/tissue doses for male and female voxel phantoms were computed as well as their summation to obtain the effective dose coefficients. The used refined energy grid tracks very closely the cross-sections in the vicinity of the resonances. The resulting refined energy grid coefficients are compared to coefficients for the coarser energy grid used in ICRP Publication 116. Additionally, reference spectra have been folded with both the fine and coarse sets of conversion coefficients. The resulting total effective doses for these reference spectra are used to assess the adequacy of the dose coefficients calculated on the original ICRP 116 energy grid. The dose coefficients were similarly computed for the local skin dose on the trunk of the body using the ICRU Report 95 phantom. The overall impact of the resonances on the organ/tissue equivalent dose, the effective dose, and the local skin dose are presented and discussed. In general, it was found that resonances can impact neutron dose coefficients, but in most cases the wide range of neutron energies encountered minimized this effect. The impact of resonances was further limited when computing effective dose due to organ/tissue summing and sex-averaging. For the neutron fields studied here, the impact was below 5%.

Synthesis and optimization of alkaline earth borate glasses doped with Fe2O3: Significance of BaO/MgO on the physical, structural features and radiation shielding performance

In this study, MgO–BaO–B2O3–Fe2O3 glasses were synthesized using the conventional melt and quench method, and the role of BaO and MgO in dictating their structural, physical, and optical properties was investigated. The fast neutrons, electrons, and heavy charged radiation transmission abilities of the glasses were delineated with the aim of evaluating their capacity to act as radiation insulators. The density of the glasses decreased from 3.40 g/cm3 to 2.89 g/cm3 as the MgO content increased. Also, the obtained optical bandgap increased from 1.90 eV to 2.20 eV with the increase of MgO in the MBBFe glasses. The stopping powers of electrons vary from 1.39 to 11.01 MeVcm2/g, 1.41–11.37 MeVcm2/g, 1.43–11.69 MeVcm2/g, 1.44–11.87 MeVcm2/g and 1.46–12.05 MeVcm2/g for glasses with MgO contents of 0 (MBBFe0), 5 (MBBFe1), 10 (MBBFe2), 12.5 (MBBFe3), and 15 (MBBFe4) mol%, respectively. Generally, the increase in the MgO content of the glasses led to increased mass stopping powers of the protons, alpha particles, and carbon ions. The replacement of BaO with MgO compromised the thermal neutron interaction probabilities of the present glassy specimens. The MBBFe-0 glass had the best thermal neutron interaction cross sections. The fast neutron removal cross-section of the glasses was within the range of 0.1037–0.1070 cm−1 with MBBFe-3 having the highest value and MBBFe-4 the least. The increase in the BaO suppresses photon buildup within the glass system. The MBBFe-glasses displayed good physical and optical properties, photon and neutron shielding abilities, which makes them attractive for diverse shielding and optical functions.

Gamma, neutron, and charged particle shielding performance of ABKT glass system

In this study, the gamma, charged, thermal neutron, and fast neutron radiation shielding parameters of (5+x)K2O-(25-x)TeO2-(10-x)As2O3-(60-x)B2O3 (where x ranged from 0 (ABKT1) to 20 (ABKT5) mol% in 5 mol% steps) were calculated and simulated. In addition, important optical parameters such as the reflection loss (Rloss), molar refractivity (Rm), molar polarizability (αm), optical transmission (TOpt), metallization criterion (M(n)), static dielectric constant (Ɛsta), optical dielectric constant (Ɛopt) of the glasses, were computed and juxtaposed with the chemical composition of the glasses. The values of Rm and αm decreased from 17.213 cm3/mol to 15.247 cm3/mol and 6.830 x10−24 cm3 to 6.05 x10−24 cm3, respectively, as K2O replaced TeO2 in the ABTK glass system. Similarly, Rloss reduced from 0.173 to 0.164, while TOpt increased from 0.705 to 0.719. The values of M(n), εsta, and εOpt were in the range 0.381–0.397, 5.565–5.87, and 4.565–4.871, respectively. The mass attenuation coefficients of ABKT1–ABKT5 were within the range of 0.0276–30.8789, 0.0269–29.6485, 0.0262–28.3373, 0.0255–29.9370, and 0.0249–25.4384 cm2/g, respectively. ABKT1 is the best gamma photon absorber among the ABKT glasses for narrow and broad beam transmission modes. For electrons, protons, alpha particles, and heavy carbon ions, the mass stopping powers of the glasses follow the order ABKT1>ABKT2>ABKT3>ABKT4>ABKT5. ABKT1 had the best photon and ion absorbing capabilities, while ABKT5 has the highest thermal and fast neutron interaction cross-sections. The ABKT1 glass demonstrated the potential to perform better than some commercial glass shields, standard shielding glasses, light and heavy concretes, and a host of other materials as a transparent gamma radiation shielding material.

Novel machine-learning method for spin classification of neutron resonances

The performance of nuclear reactors and other nuclear systems depends on a precise understanding of the neutron interaction cross sections for materials used in these systems. These cross sections exhibit resonant structure whose shape is determined in part by the angular-momentum quantum numbers of the resonances. The correct assignment of the quantum numbers of neutron resonances is, therefore, paramount. In this project, we apply machine learning to automate the quantum number assignments using only the resonances' energies and widths and not relying on detailed transmission or capture measurements. The classifier used for quantum number assignment is trained using stochastically generated resonance sequences whose distributions mimic those of real data. We explore the use of several physics-motivated features for training our classifier. These features amount to out-of-distribution tests of a given resonance's widths and resonance-pair spacings. We pay special attention to situations where either capture widths cannot be trusted for classification purposes or where there is insufficient information to classify resonances by the total spin $J$. We demonstrate the efficacy of our classification approach using simulated and actual $^{52}\mathrm{Cr}$ resonance data.

Determination of the nuclear radius parameter using the γ-ray spectrometer

The nuclear radius parameter of carbon, aluminium, iron, copper, and zinc nuclei has been determined by using (n,γ)-reaction. The neutrons from the americium-beryllium source are made to interact with the water moderator to produce the γ-rays of 2.2 MeV through (n,γ)-reaction. The γ-radiation emitted from the water medium is measured with a scintillation detector coupled to 8k multi-channel analyzer. The neutrons from the americium-beryllium source are allowed to transmit through carbon, aluminium, iron, copper, and zinc elemental targets of various thicknesses, and transmitted neutrons are again allowed to interact with water moderators to produce 2.2 MeV γ-radiation. By measuring the yield of γ-radiation produced in water moderators by neutrons transmitted through elemental targets of different mass number values, the total neutron interaction cross-sections are determined. By knowing the total neutron interaction cross-sections and mass number of the target nuclei, the radius parameter has been determined.

A microdosimetric analysis of the interactions of mono-energetic neutrons with human tissue

Nuclear reactions induced during high-energy radiotherapy produce secondary neutrons that, due to their carcinogenic potential, constitute an important risk for the development of iatrogenic cancer. Experimental and epidemiological findings indicate a marked energy dependence of neutron relative biological effectiveness (RBE) for carcinogenesis, but little is reported on its physical basis. While the exact mechanism of radiation carcinogenesis is yet to be fully elucidated, numerical microdosimetry can be used to predict the biological consequences of a given irradiation based on its microscopic pattern of energy depositions. Building on recent studies, this work investigated the physics underlying neutron RBE by using the microdosimetric quantity dose-mean lineal energy (y‾D) as a proxy. A simulation pipeline was constructed to explicitly calculate the y‾D of radiation fields that consisted of (i) the open source Monte Carlo toolkit Geant4, (ii) its radiobiological extension Geant4-DNA, and (iii) a weighted track-sampling algorithm. This approach was used to study mono-energetic neutrons with initial kinetic energies between 1 eV and 10 MeV at multiple depths in a tissue-equivalent phantom. Spherical sampling volumes with diameters between 2 nm and 1 μm were considered. To obtain a measure of RBE, the neutron y‾D values were divided by those of 250 keV X-rays that were calculated in the same way. Qualitative agreement was found with published radiation protection factors and simulation data, allowing for the dependencies of neutron RBE on depth and energy to be discussed in the context of the neutron interaction cross sections and secondary particle distributions in human tissue.

Fast neutron resonance radiography with full time-series digitization

Fast neutron resonance radiography (FNRR) is an emerging technology for resolving the in-depth elemental composition of samples. It relies on the modification of a broad neutron energy spectrum transmitted through the sample due to resonant features in its neutron interaction cross-sections. FNRR has yet to reach widespread use because of its high cost and the large size of high-flux neutron generators, and of the radiography setup. The realization of compact neutron generators based on high-intensity lasers motivate reducing the size of FNRR setups from 10- to few-meter scale. The challenge is posed by the fact that for a fixed temporal resolution, a shorter neutron flight path corresponds to reduced resolution in energy. Here we address this challenge by recording the full neutron-time-of-flight time-series for each and every pixel. The rich spectral information reveals unique neutron interaction features, even with low energy resolution. We present the characterization of a proof-of-principle detector, using a spallation neutron source. The results are used to assess the requirements for a fully applicable neutron imager for FNRR applications.

Semi-empirical analysis of the spatial variation of Pu quality ratio in a reactor core for application in pyroprocessing of used nuclear fuel

The axial variation of Pu quality ([Pu-239 + Pu-241]/Pu) in used nuclear fuel is analyzed using MCNP simulations. The Pu quality ratio variation is found to be due to the differences in axial neutron flux and neutron interaction cross-sections of U and Pu isotopes. The Pu axial non-uniformity is found to be larger for isotopes with smaller neutron cross-sections. Analysis shows initially, the axial neutron flux distribution dictated the Pu-239 concentration. When the fuel irradiation time increased, the ratio of U-238 neutron capture to Pu-239 neutron absorption cross-section is found to influence the Pu-239 concentration. A fuel rod concept with a varied axial uranium enrichment is modeled and simulated to understand the non-uniformity of Pu quality ratio. The results show that not only the peaks for Pu quality ratio reduced at the top (3.63%) and bottom (4.62%) of the assembly but also a maximum reduction of 6.25% in non-uniformity is obtained.

Evaluations of n + 27Al reaction in the energy range 0.1–200 MeV

As 27Al has important applications in reactor science and technology, perfect understanding of its various cross sections in the wide range of energy is indispensable. Scarcity of experimental data particularly in high energy makes theoretical evaluation essential. In the present work, different neutron interaction cross sections of 27Al in the energy range 0.1–200 MeV have been evaluated by TALYS 1.9. The reaction grid, outgoing light particle production and probable production routes are also analyzed. The present evaluations are compared with the available experimental and evaluated nuclear data: ENDF/B-VIII, JENDL-4.0, etc. This is generally in good agreement except 0.7–6 MeV of projectile energies. In this energy range, diffuseness and geometrical parameters are adjusted to overcome the disagreement. In the lower-energy region, mainly elastic scattering is observed. Shape of non-elastic curve is well produced; however, there is small energy shift between evaluated and experimental data.

Generation of the windowed multipole resonance data using Vector Fitting technique

The multipole representation provides an analytical way of Doppler broadening cross sections for neutron interactions and relies on a partial fraction decomposition that represents R-matrix resonance parameters by poles and residues. The windowed multipole method selectively broadens impactful poles and approximates the rest to increase the efficiency of the process. However, this process requires knowledge of the resonance parameters and certain limitations occur when additional channels are opened in the defined resolved resonance region such that only 2/3 of the ENDF/B-VII.1 nuclides can be processed. Some important nuclides needed in reactor applications such as Oxygen-16 and Hydrogen-1 are directly represented in point-wise form in the evaluation nuclear data library hindering the practicality of this approach. This paper presents a new fitting method for nuclear point-wise cross sections data that yields a pole and residue form analogous to the multipole representation. The Vector Fitting technique which originates from the field of signal processing was applied to the fitting of point-wise cross sections. Poles and residues from Vector Fitting were generated for Oxygen-16, Hydrogen-1, Boron-10 and Boron-11 and processed in the windowed multipole format. These new libraries were tested by direct comparison of the microscopic cross sections before and after Doppler Broadening, and by integral comparisons using a typical PWR pin cell from the BEAVRS benchmark. Results indicate that the new libraries are equivalent to the point-wise representation with the added benefit of allowing on-the-fly Doppler broadening without the need for temperature interpolation. Runtimes are approximately ∼28% slower using the multipole representation on this example problem compared to a single temperature ACE file.

Evaluation of cross sections for neutron interactions with 238U in the energy region between 5 keV and 150 keV

.Cross sections for neutron interactions with 238U in the energy region from 5keV to 150keV have been evaluated. Average total and capture cross sections have been derived from a least squares analysis using experimental data reported in the literature. The resulting cross sections have been parameterised in terms of average resonance parameters maintaining full consistency with results of optical model calculations by using a dispersive coupled channel optical model potential. The average compound partial cross sections have been expressed in terms of transmission coefficients by applying the Hauser-Feshbach statistical reaction theory including width-fluctuations. A generalized single-level representation compatible with the energy-dependent options of the ENDF-6 format has been applied using standard boundary conditions. The results have been transferred into a full ENDF-6 compatible data file.

Secondary neutrons inside a proton therapy facility: MCNPX simulations compared to measurements performed with a Bonner Sphere Spectrometer and neutron H*(10) monitors

Neutron spectrometry measurements with an extended-range Bonner Sphere Spectrometer (BSS), as well as neutron H*(10) measurements using an extended-range rem meter WENDI-2, a conventional rem meter LB 6411 and a tissue-equivalent proportional counter, were performed inside and around the Fixed-Beam Treatment Room at the proton therapy facility of Essen, in Germany. The WENDI-2 stood out as the easiest detector for making accurate neutron H*(10) measurements, since its direct measurements were equivalent to the H*(10) rates obtained with the BSS. The measurements were also compared to simulation results obtained with MCNPX 2.7.0 using two different selections of physics models for the hadron interactions above 150 MeV: the Bertini & Dresner models and the CEM03 model. For neutron H*(10) rates outside the treatment room, factors of 1.6–1.8 were obtained between the results of the two simulations, the Bertini & Dresner models yielding the largest values in all positions. The comparison of the simulation results with the WENDI-2 and BSS measurements for positions inside the treatment room showed that the Bertini & Dresner models reproduce the global neutron production in the water phantom relatively well, whereas the CEM03 model underestimates it by a factor of ∼1.3. At the most-forward angle, however, the Bertini model (unlike the CEM03 model) seemed to overestimate the production of neutrons with energies above 100 MeV. Outside the shielding, the simulated H*(10) overestimated the WENDI-2 measurements by factors of 2–3 with the Bertini & Dresner models, and 1.1–1.7 with the CEM03 model. Both simulations were thus conservative with respect to the neutron fluxes transmitted through the concrete walls. This conservative behaviour is probably caused by a combination of several uncertainties, including for instance uncertainties on the proton and neutron interaction cross-sections and uncertainties on the concrete composition and density.

A calibration method for realistic neutron dosimetry in radiobiological experiments assisted by MCNP simulation.

Many studies on biological effects of neutrons involve dose responses of neutrons, which rely on accurately determined absorbed doses in the irradiated cells or living organisms. Absorbed doses are difficult to measure, and are commonly surrogated with doses measured using separate detectors. The present work describes the determination of doses absorbed in the cell layer underneath a medium column (DA) and the doses absorbed in an ionization chamber (DE) from neutrons through computer simulations using the MCNP-5 code, and the subsequent determination of the conversion coefficients R (= DA/DE). It was found that R in general decreased with increase in the medium thickness, which was due to elastic and inelastic scattering. For 2-MeV neutrons, conspicuous bulges in R values were observed at medium thicknesses of about 500, 1500, 2500 and 4000 μm, and these were attributed to carbon, oxygen and nitrogen nuclei, and were reflections of spikes in neutron interaction cross sections with these nuclei. For 0.1-MeV neutrons, no conspicuous bulges in R were observed (except one at ~2000 μm that was due to photon interactions), which was explained by the absence of prominent spikes in the interaction cross-sections with these nuclei for neutron energies <0.1 MeV. The ratio R could be increased by ~50% for small medium thickness if the incident neutron energy was reduced from 2 MeV to 0.1 MeV. As such, the absorbed doses in cells (DA) would vary with the incident neutron energies, even when the absorbed doses shown on the detector were the same.

Recent Developments in Nuclear Data Measurement capabilities at the Gaerttner LINAC Center at RPI

The Gaerttner LINAC Center at RPI uses a 60 MeV electron linear accelerator to produce short pulses of neutrons with duration of 5-5000 ns. The main research thrust at the Center is nuclear data for nuclear reactors and criticality safety applications. The Center includes several setups for time-of-flight measurements including neutron transmission, capture and scattering detectors, and a lead slowing-down spectrometer. Experiments were designed to produce neutron interaction cross sections that cover the energy range of 0.01 eV to 20 MeV. Recently added experiments include: setups for keV and fast neutron transmission, a C 6 D 6 detector array for keV neutron capture measurements, and a fast neutron scattering system. Results discussed here include fast neutron scattering and angular distributions for nat Fe, iron capture measurements for incident neutrons from 1 keV to 2 MeV, fast neutron transmission through W and H 2 O samples, and keV transmission through Mo isotopes.

THE USE OF FAST NEUTRON DETECTION FOR MATERIALS ACCOUNTABILITY

For many years at LLNL, we have been developing time-correlated neutron detection techniques and algorithms for applications such as Arms Control, Threat Detection and Nuclear Material Assay. Many of our techniques have been developed specifically for the relatively low efficiency (a few percent) inherent in man-portable systems. Historically, thermal neutron detectors (mainly 3 He ) were used, taking advantage of the high thermal neutron interaction cross-sections, but more recently we have been investigating the use of fast neutron detection with liquid scintillators, inorganic crystals, and in the near future, pulse-shape discriminating plastics that respond over 1000 times faster (nanoseconds versus tens of microseconds) than thermal neutron detectors. Fast neutron detection offers considerable advantages, since the inherent nanosecond production timescales of fission and neutron-induced fission are preserved and measured instead of being lost in the thermalization of thermal neutron detectors. We are now applying fast neutron technology to the safeguards regime in the form of high efficiency counters. Faster detector response times and sensitivity to neutron momentum show promise in measuring, differentiating, and assaying samples that have modest to very high count rates, as well as mixed neutron sources (e.g., Pu oxide or Mixed Cm and Pu ). Here we report on measured results with our existing liquid scintillator array and promote the design of a nuclear material assay system that incorporates fast neutron detection, including the surprising result that fast liquid scintillator becomes competitive and even surpasses the precision of 3 He counters measuring correlated pairs in modest (kg) samples of plutonium.

Cross sections for neutron interactions in the CUORE neutrinoless double beta decay experiment

We describe experiments to determine the cross sections for neutron reactions on tellurium isotopes. These cross sections will be used to improve Monte Carlo simulations for CUORE, a next generation double beta decay experiment.

A Background-Free Direction-Sensitive Neutron Detector

The detection and measurements of properties of neutrons are of great importance in many fields of research, including neutron scattering and radiography, measurements of solar and cosmic ray neutron flux, measurements of neutron interaction cross sections, monitoring of neutrons at nuclear facilities, oil exploration, and searches for fissile weapons of mass destruction. Many neutron detectors are plagued by large backgrounds from x-rays and gamma rays, and most current neutron detectors lack single-event energy sensitivity or any information on neutron directionality. Even the best detectors are limited by cosmic ray neutron backgrounds. All applications would benefit from improved neutron detection sensitivity and improved measurements of neutron properties. Here we show data from a new type of detector that can be used to determine neutron flux, energy distribution, and direction of neutron motion. The detector is free of backgrounds from x-rays, gamma rays, beta particles, and relativistic singly charged particles. It is relatively insensitive to cosmic ray neutrons because of their distinctive angular and energy distributions. It is sensitive to thermal neutrons, fission spectrum neutrons, and high energy neutrons, with detection features distinctive for each energy range. It is capable of determining the location of a source of fission neutrons based on characteristics of elastic scattering of neutrons by helium nuclei. A portable detector could identify one gram of reactor grade plutonium, one meter away, with less than one minute of observation time.

Nuclear data evaluation of 55Mn by the EMPIRE code with emphasis on the capture cross-section

Manganese is one of the constituents of alloys for structural components of fission and fusion devices and a well-known neutron dosimeter; however, existing ENDF-B/VII.0 55Mn evaluation was produced by Shibata (1989). This work is an attempt to re-evaluate neutron-induced cross-sections of 55Mn using the latest release of the EMPIRE code. Sensitivity studies on the physical and fitting parameters are presented, with special emphasis on the capture and neutron inelastic cross-sections. A calculated nuclear data file in ENDF-6 format of the neutron interaction cross-sections is produced. It extends up to 150 MeV, which is of interest for fusion and accelerator driven system applications. This evaluation is compared with the ENDF/B-VII.0 evaluation and with a selection of experimental microscopic cross-sections. The evaluation is tested using integral data: the OKTAVIAN integral experiment on a manganese shell and an FNG experiment with manganese activation foils. Benchmark results provide needed feedback for the refinement of the physics parameters.

NEUTRON-INDUCED CAVITATION TENSION METASTABLE PRESSURE THRESHOLDS OF LIQUID MIXTURES

Tensioned metastable fluids provide a powerful means for low-cost, efficient detection of a wide range of nuclear particles with spectroscopic capabilities. Past work in this field has relied on one-component liquids. Pure liquids may provide very good detection capability in some aspects, such as low thresholds or large radiation interaction cross sections, but it is rare to find a liquid that is a perfect candidate on both counts. It was hypothesized that liquid mixtures could offer optimal benefits and present more options for advancement. However, not much is known about radiation-induced thermal-hydraulics involving destabilization of mixtures of tensioned metastable fluids. This paper presents results of experiments that assess key thermophysical properties of liquid mixtures governing fast neutron radiation-induced cavitation in liquid mixtures. Experiments were conducted by placing liquid mixtures of various proportions in tension metastable states using Purdue’s centrifugally-tensioned metastable fluid detector (CTMFD) apparatus. Liquids chosen for this study covered a good representation of both thermal and fast neutron interaction cross sections, a range of cavitation onset thresholds and a range of thermophysical properties. Experiments were devised to measure the effective liquid mixture viscosity and surface tension. Neutron-induced tension metastability thresholds were found to vary non-linearly with mixture concentration; these thresholds varied linearly with surface tension and inversely with mixture vapor pressure (on a semi-log scale), and no visible trend with mixture viscosity nor with latent heat of vaporization.

A background-free direction-sensitive neutron detector

We show data from a new type of detector that can be used to determine neutron flux, energy distribution, and direction of neutron motion for both fast and thermal neutrons. Many neutron detectors are plagued by large backgrounds from x-rays and gamma rays, and most current neutron detectors lack single-event energy sensitivity or any information on neutron directionality. Even the best detectors are limited by cosmic ray neutron backgrounds. All applications (neutron scattering and radiography, measurements of solar and cosmic ray neutron flux, measurements of neutron interaction cross sections, monitoring of neutrons at nuclear facilities, oil exploration, and searches for fissile weapons of mass destruction) will benefit from the improved neutron detection sensitivity and improved measurements of neutron properties made possible by this detector. The detector is free of backgrounds from x-rays, gamma rays, beta particles, relativistic singely charged particles and cosmic ray neutrons. It is sensitive to thermal neutrons, fission neutrons, and high energy neutrons, with detection features distinctive for each energy range. It is capable of determining the location of a source of fission neutrons based on characteristics of elastic scattering of neutrons by helium nuclei. The detector we have constructed could identify one gram of reactor grade plutonium, one meter away, with less than one minute of observation time.