Neutron Cross Sections Research Articles

Measurements of neutron capture cross-sections for nuclides of interest in decommissioning: 45Sc, 63Cu, 64Zn, 109Ag, and 113In

ABSTRACT As nuclear facilities are dismantled in decommissioning, various and large amounts of waste are generated. Even more inconveniently, they are radioactive waste due to neutron activation. Thus, the neutron capture cross-sections of nuclides targeted for decommissioning are required to evaluate the radioactivity produced. The present study selected 45Sc, 63Cu, 64Zn, 109Ag and 113In among targeted nuclides, and performed the measurements of thermal-neutron capture cross-sections by an activation method at the Kyoto University Research Reactor. The thermal-neutron capture cross-sections were derived on the basis of Westcott’s convention. The present results were obtained as follows: 27.18 ± 0.28 barn for 45Sc(n,γ)46Sc, 4.34 ± 0.06 barn for 63Cu(n,γ)64Cu, 0.719 ± 0.011 barn for 64Zn(n,γ)65Zn, 4.05 ± 0.05 barn for 109Ag(n,γ)110mAg and 8.53 ± 0.27 barn for 113In(n,γ)114Inm1+m2 reactions. The results for 45Sc and 64Zn support evaluated cross-section data within the limits of uncertainties, while those for the other nuclides were slightly different from evaluated ones. The obtained results are useful not only for the evaluation of isotope production, but also for the consideration of those nuclides as flux monitors, as an alternative to Au and Co.

Cross-section induced uncertainties in neutron source mission rate calculations

Analysis presented in the paper is focused on the characterization of uncertainties induced by cross-section data, which contributing to the overall uncertainty of the correction factors. Advances in computational methods and computational power shifted the calculation of correction factors among the standard steps of the manganese bath method what decrease an interest about this part of the method. Anyway, with the development of computational tools, the cross-section data were also improved, and new evaluations of nuclear data libraries include more information about the cross-section data covariances. Therefore, the propagation of uncertainties induced by neutron cross-sections can be carried out within standard transport calculation of the manganese sulphate bath model. In this paper, the super-sequence SAMPLER module that implements stochastic techniques is used to assess the uncertainty in computed results. Reaction rates on all nuclides of the solution are computed in 400 cases with uncertain parameters and the results are evaluated by an auxiliary tool. Consideration of nuclear data uncertainty in calculations is a general trend that requires the attention of nuclear researchers and should draw attention in metrology. Additional 1.5 % (one-sigma) uncertainty is introduced to the overall uncertainty through correction factor.

Corrosion and fretting-corrosion behavior of Zr-Nb alloy under aqueous LiOH solution

Zirconium alloys are the key materials for nuclear reactor’s core components due to low neutron cross-section and good mechanical and corrosion properties. During operation in nuclear reactor, some core components (exposed to the corrosive environment of coolant water having LiOH) may undergo corrosion and fretting-corrosion. This work presents a novel experimental approach for investigating the Zr-2.5 Nb alloy's corrosion and in-situ fretting-corrosion response. The corrosion analysis was performed using Electrochemical Impedance Spectroscopy (EIS) and potentiodynamic polarization, to study the effect of LiOH concentration in water and exposure duration. It is observed that the corrosion rate increases with increasing the concentration of LiOH in water and decreases with increasing the exposure duration. The fretting-corrosion tests were performed between SS-410 and Zr-2.5 Nb in aqueous LiOH solution. The results reveal that the contact resistance and coefficient of friction (COF) decrease with increasing concentration of LiOH in the solution. The present work highlights the scientific understanding of corrosion and fretting-corrosion phenomena in zirconium alloy and elucidates key mechanisms and factors governing the degradation behavior which can help in preventing the failure of the components.

A high-intensity, low-energy heavy ion source for a neutron target proof-of-principle experiment at LANSCE

The capability to directly study neutron capture reactions on radionuclides with half-lives on the of order minutes would allow key cross section measurements in nuclear astrophysics and energy applications. However, such experiments with stationary targets are currently impossible because of signal detection and target sample fielding issues. To overcome these challenges, a neutron target facility is being developed to permit neutron capture experiments on unstable isotopes in inverse kinematics at the Los Alamos Neutron Science Center (LANSCE). This next-generation facility will consist of a heavily moderated, high-intensity spallation neutron target coupled with a radioactive ion beam storage ring. A proof-of-principle experiment is underway to demonstrate this neutron target concept at LANSCE in the near term with stable ions and without a storage ring. Here, mA-level beams of 10-50 keV heavy ions exhibiting large resonant neutron capture cross sections will be generated by a new ion source, transported through a large-volume neutron moderator surrounding an adjacent spallation target driven by the LANSCE accelerator, and collected downstream of the moderator for subsequent decay-counting measurements. The neutron density within the moderator will be obtained from these decay yields and compared with Monte Carlo N-Particle (MCNP) simulation results. The science application, operational requirements, and performance objectives for this heavy ion source will be presented.

Evaluation of neutron cross sections of 48Ti based on the Unified-Monte-Carlo-B method* *Supported by the National Key Research and Development (R&D) Program (2022YFA1602403) and Continuous Support Basic Scientific Research Project BJ010261223282

A cross section evaluation of neutron induced reactions on 48Ti is undertaken using the Unified Monte Carlo-B (UMC-B) approach. The evaluation concentrates on estimating the covariance and the use of the UMC-B allows avoiding the deficiencies of linear regression brought by the traditional least squares method. Eight main neutron and charged particle emission reactions from n+48Ti in the fast neutron energy region below 20 MeV are studied in this work. The posterior probability density function (PDF) of each neutron cross section is obtained in a UMC-B Bayesian approach by convoluting the model PDFs sampled based on model parameters and the likelihood functions for the experimental data. Nineteen model parameters including level density, pair corrections, optical model and Kalbach matrix element parameter are stochastically sampled with the assumption of normal distributions to estimate the model uncertainty. The Cholesky factorization approach is applied to consider potential parameter correlations. Finally, the posterior covariance matrices are generated using the UMC-B generated weights. The new evaluated results are compared with the CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, TENDL-2021 and JENDL-5 evaluations and differences are discussed.

Investigation of the Cd-Ratio difference method to measure the cross section of 186W(n,γ)187W at the Jordan Research and Training reactor

The neutron capture cross section of 186W(n,γ)187W was measured at different energies, including the thermal capture cross section (0.025 eV) and resonance integral. Neutron capture cross section values were determined using the activation technique and by applying the Cadmium-ratio difference method. Two irradiation locations at the Jordan Research and Training Reactor (JRTR) were used (NAA1 and NAA2). The thermal capture cross section values were determined to be 39.56 ± 0.69 ± 2.58b and 32.17 ± 3.72 ± 2.00b at NAA1 and NAA2, respectively. While resonance integral values were found to be 113.93 ± 8.20 ± 16.87b and 69.35 ± 26.63 ± 29.69b at NAA1 and NAA2, respectively. The present results of the resonance integral are not in good agreement with the most of the previous experimental and evaluated data as well as the theoretical calculations.

Pushing the high count rate limits of scintillation detectors for challenging neutron-capture experiments

One of the critical aspects for the accurate determination of neutron capture cross sections when combining time-of-flight and total energy detector techniques is the characterization and control of systematic uncertainties associated to the measuring devices. In this work we explore the most conspicuous effects associated to harsh count rate conditions: dead-time and pile-up effects. Both effects, when not properly treated, can lead to large systematic uncertainties and bias in the determination of neutron cross sections. In the majority of neutron capture measurements carried out at the CERN n_TOF facility, the detectors of choice are the C6D6 liquid-based either in form of large-volume cells or recently commissioned sTED detector array, consisting of much smaller-volume modules. To account for the aforementioned effects, we introduce a Monte Carlo model for these detectors mimicking harsh count rate conditions similar to those happening at the CERN n_TOF 20 m flight path vertical measuring station. The model parameters are extracted by comparison with the experimental data taken at the same facility during 2022 experimental campaign. We propose a novel methodology to consider both, dead-time and pile-up effects simultaneously for these fast detectors and check the applicability to experimental data from 197Au(n, γ), including the saturated 4.9 eV resonance which is an important component of normalization for neutron cross section measurements.

Breakthrough electroneutron multi-response miniature dosimetry/spectrometry in medical accelerator

Breakthrough multi-response miniature dosimetry/spectrometry of electroneutrons (EN) was made on surface and in-depths of whole-body polyethylene phantom under 10 cm × 10 cm electron beam of 20 MV Varian Clinac 2100C electron medical accelerator commonly applied for prostate treatment. While dosimetry/spectrometry of photoneutrons (PN) has been well characterized for decades, those of ENs lagged behind due to very low EN reaction cross section and lack of sensitive neutron dosimeters/spectrometers meeting neutron dosimetry requirements. Recently, Sohrabi “miniature neutron dosimeter/spectrometer” and “Stripe polycarbonate dosimeter” have broken this barrier and determined seven EN ambient dose equivalent (ENDE) (µSv.Gy–1) responses from electron beam and from albedo ENs including beam thermal (21 ± 2.63), albedo thermal (43 ± 3.70), total thermal (64 ± 6.33), total epithermal (32 ± 3.90), total fast (112.00), total thermal + epithermal (l96 ± 10), and total thermal + epithermal + fast (208 ± 10.23) ENs. Having seven ENDE responses of this study and seven PNDE responses of previous study with the same accelerator obtained at identical conditions by the same principle author provided the opportunity to compare the two sets of responses. The PNDE (µSv.Gy–1) responses have comparatively higher values and 22.60 times at isocenter which provide for the first time breakthrough ENDE responses not yet reported in any studies before worldwide.

Verification of shielding calculation capability of cosRMC with SINBAD fusion benchmarks

In the design of fusion reactors, extensive shielding calculations are necessary to verify the feasibility and safety of the design. The issue of deep penetration is a common and significant problem in fusion reactor shielding calculations. When the thickness of the shielding layer exceeds a certain value, the statistical error in Monte Carlo method calculations significantly increases. Therefore, it is imperative to adopt variance reduction techniques to enhance the computational accuracy for deep penetration problems. cosRMC is a reactor Monte Carlo code, and its accuracy in calculating neutron-deep-penetration shielding problems should be verified. For this purpose, three deep penetration benchmarks from Shielding Integral Benchmark Archive Database (SINBAD) were simulated by cosRMC and neutron cross section data library ENDF/B-VII.0. These benchmarks describe the penetration behaviors of neutrons in models with different geometric shapes and materials. This paper shows detailed cosRMC models based on benchmark experimental models, used to calculate neutron flux spectra at different penetration depths, which are recorded by detectors in the related benchmark experiments. In this way, a direct comparison between cosRMC results and experimental measurements is provided. In order to improve the accuracy of cosRMC in calculating deep penetration problems, the variance reduction technique of cell importance was used. The neutron flux spectrum calculated by cosRMC was compared with the neutron flux spectrum results calculated by MCNP5 and the experimental values of the benchmarks. The comparison results show that the calculations of cosRMC are in compliance with the MCNP5 reference values and experimental values, and the relative deviations between calculations of cosRMC and MCNP5 is within three time of standard errors. Therefore, this paper has verified that cosRMC can applies to neutron deep penetration problems of fusion reactors.

Pure CaF2 crystal for fast neutron detection

Pure CaF2 crystals have been widely used for optical functional crystals due to their good optical properties, mechanical properties, and chemical stability. This study discovered the crystal's pulse shape discrimination (PSD), proving it a promising scintillator for fast neutron detection via the 40Ca neutron capture reaction. Luminescence emission spectra were studied by X-ray luminescence measurements. The maximum emission peak was observed between 250 and 300 nm, corresponding to self-trap exciton emission. The crystal response to gamma, alpha, and neutron were studied under 137Cs, 241Am, and 252Cf radiation sources, respectively. The difference in gamma and alpha decay times enabled the PSD capability of pure CaF2. Large neutron reaction cross-sections, PSD capability, and good chemical stability make pure CaF2 crystal candidates for fast neutron beam monitoring detectors. Moreover, the crystal can be used for 40Ca(n,p)40K and 40Ca(n,α)37Ag neutron capture cross-section study.

Computing the gamma-ray, charged particles and fast neutron-shielding performances of selected alloys

Gamma-ray, charged particles and neutron-shielding properties of ternary alloy materials, (97.3 – x) Pb – x Cd – 2.7Ag (x = 10, 18 and 30) have been theoretically investigated. First, the effectiveness of gamma-ray shielding for the alloys was studied by using the Phy-X/PSD programme. Mass and linear attenuation coefficients (MAC and LAC), half-value layer (HVL), tenth-value layer (TVL), mean free path (MFP), effective atomic numbers (Zeff), effective electron densities (Neff) and effective conductivity (Ceff) were calculated for photon energy range 0.015–15 MeV. Second, the mass stopping powers (MSPs; dE/ρdx; MeVcm2/g) and ranges for electron, proton and alpha particle interactions were calculated at energy 0.01–20 MeV. Finally, the removal cross-section of fast neutron (FNRCS) for the alloys was calculated by the partial density method. Moreover, the neutrons and gamma-rays shielding effectiveness of ternary alloys were investigated using Monte Carlo (MCNP4b). The MCNP4b calculations were performed for neutrons in the energies 4 and 4.5 MeV. While the shielding parameters of gamma rays were done in the energy range of 0.015–15 MeV. The obtained results of radiation shielding showed that the P1 sample owns a good shielding effectiveness against neutrons and gamma rays compared to previously published studies. The comparison of the computed values has revealed a reasonable agreement between Phy-X/PSD and MCNP4b. This study gives essential data for potential applications of these materials in nuclear reactor core design and other industries for the evaluation of effective ionising radiation shielding materials.

Radiation shielding for inertial electrostatic confinement fusion system utilizing concrete and water

In this investigation, the impact of water thickness and various types of concrete materials, each characterized by different densities and elemental compositions, is examined for their role in shielding a radiation generator based on Inertial Electrostatic Confinement Fusion (IECF) device. The study focuses on assessing several vital parameters, including the effective removal cross-section (∑Rt) for fast neutrons, total mass attenuation coefficients (μmt), linear attenuation coefficients (μl), half-value layer (HVL), and mean free path (MFP) for X-rays across different concrete types. Different water thicknesses around the IECF chamber, ranging from 0.5 to 10.0 cm, are investigated, and five concrete types are evaluated: Ilmenite-magnetite Concrete (IMC), Ordinary Concrete-1 (OC1), Barite-containing Concrete (BC), Ordinary Concrete-2 (OC2), and Serpentine-containing Concrete (SC). The results indicate that, among these materials, SC requires the least thickness to attenuate 2.45 MeV generated from IECF to 1/100th of its initial intensity across varying water thicknesses (6.5 cm in case of 5 cm water thickness). The values of μmt, HVL, and MFP are also calculated for different water thicknesses and X-ray energies (ranging from 0.2 to 3.0 MeV). These calculations highlight BC as the material requiring the least thickness to attenuate X-rays to 1/100th of their initial intensity. Moreover, neutron dose rate measurements are conducted on a commercial IECF system shielded with 50 cm of water and operated at a neutron intensity of 105 ns−1, which was ∼12 nSv/h on average, approximately 0.002% of the initial intensity. This underscores the efficacy of water shielding in attenuating the outcomes of IECF.

Novel integrated structure and function of Mg–Gd neutron shielding materials

Abstract As the lightest metal structural materials, magnesium (Mg) alloys offer extensive application potential. Gadolinium (Gd), as the primary alloying element in Mg alloys and recognized for its notable thermal neutron capture cross-section, is considered one of the most efficient neutron absorbers. Thus, the Mg–Gd alloy is highly expected to emerge as a material with remarkable neutron absorption capacity. Hence, in this study, the thermal neutron-shielding capabilities of Mg–Gd alloys were comprehensively examined by fabricating four as-cast Mg–xGd alloys with varying compositions (x = 0, 5, 10, and 15 wt%). The obtained results were further corroborated by sophisticated modeling and calculations using SuperMC. The results revealed a direct correlation between the thermal neutron absorption capacity of the Mg–Gd alloys and the increase in Gd content, with a noteworthy neutron attenuation factor of 22.33. Moreover, in an Au ion irradiation experiment conducted at 200°C, the Mg–15Gd alloy exhibited exceptional radiation resistance, with a displacement per atom (dpa) of 10. The matrix and second-phase regions were devoid of any cavity formation. Instead, a finite number of dislocation rings were observed, forming both leaf-like and granular Gd-rich nanoscale precipitates. This study underscores the versatility of Mg–Gd alloys as efficient neutron shielding materials and structural materials tailored for applications demanding radiation resistance in diverse environments.

Neutron and gamma-ray shielding effectiveness of novel polyaniline composites

This study uses experimental, theoretical, and Monte Carlo simulation techniques to investigate the neutron and gamma-ray shielding performances of polyaniline-based composites filled with different amounts of formic acid, boron nitride, and boric acid. The radiation shielding performances of composites were evaluated at neutron energies of 0.025 eV, 1 keV, and 10 MeV, and gamma-ray energy of 662 keV. The study found that the composite material made of polyaniline and boron nitride with a higher percentage of boron has better thermal neutron shielding properties than other polyaniline composites. The research also showed that a polyaniline-boric acid composite with a thickness of 1.6 mm can reduce thermal neutrons by 64.5%. As a result, the composite containing equal amounts of polyaniline and boric acid was identified as the most promising material for shielding. The macroscopic cross-sections of neutrons and mass attenuation coefficients of gamma-rays were found to have expanded uncertainties up to 13.47% and 7.11%, respectively. The factors contributing to the uncertainty of the experimental, theoretical, and simulation results were identified, evaluated, and shown how to include uncertainty components in the uncertainty budget.

Neutronic performance of 208Pb-Bi eutectic-cooled fast reactor with uranium nitride fuel (enriched 15N)

Fast reactors (FRs) require more uranium enrichment than LWRs and fuel reprocessing. The solution proposed to address these issues is to use the modified CANDLE burnup scheme on the reactor core. The use of uranium nitride (UN) as fuel was proposed because the UN fuel has several advantages, such as high melting point, high heavy metal density, and large thermal conductivity. However, using UN as a fuel can produce the radioactive isotope 14C via the reaction 14N (n, p)14C. Therefore, in this study, a neutronic analysis was performed in a modified CANDLE fast reactor employing UN and UN (enriched 15N) as fuel using SRAC. The modified CANDLE fast reactor fuelled with UN (99% 15N) has a higher keff than the UN-fuelled reactor with similar parameters. Similarly, the atomic density of fissile nuclides and power density are higher because the 15N isotope has a smaller neutron capture cross-section than the 14N isotope.

Theoretical analysis of the double-differential cross-sections of neutron, proton, deuteron, $$^3$$He, and $$\alpha$$ for the $$\text {p}+^6$$Li reaction

Theoretical analysis of the double-differential cross-sections of neutron, proton, deuteron, $$^3$$He, and $$\alpha$$ for the $$\text {p}+^6$$Li reaction

Radiation shielding effectiveness, structural, and mechanical properties of HDPE/B4C composites reinforced with Fe2O3-Al2O3-Al-Fe fillers

The fast neutron and gamma-ray attenuation properties of pure high-density polyethylene and its composites were investigated, including 90 HDPE/10 B4C (wt%) and 60 HDPE/10 B4C/30X (wt%) (Where X represents iron oxide, aluminum oxide, iron, and aluminum). The examined samples were analyzed using a scanning electron microscope, and their mechanical properties were assessed. The composite of 60 HDPE/10 B4C/30 α-Fe2O3 (wt%) shows better mechanical properties than those of other composites under investigation. 239Pu–Be was used as a source of fast neutrons with a neutron yield of 1.7 × 106 n s−1 and was detected by the Stilbene scintillator. Neutron removal cross-section with dependent parameters, the mean free path and half-value layer of the prepared composites, were calculated. Furthermore, the gamma-ray transmission through the produced composites has been examined. The composite containing 30% iron oxide exhibited superior shielding characteristics for neutrons and gamma rays compared to the other analyzed samples. The computed values of the shielding characteristics indicate that the produced composites are highly effective for shielding fast neutrons and gamma rays in radiation facilities.

A Neutron Beam Shaping Assembly for Boron Neutron Capture Therapy of Superficial Tumors

The Vacuum Insulated Tandem accelerator have been developed in Budker Institute of Nuclear Physics. Neutrons are generated in 7Li(p,n)7Be reaction. Neutron beam shaping assembly is used for therapeutic beam forming. It consists of the moderator, reflector and filter. Magnesium fluoride is considered optimal material for neutron slowing down because of noticeable cross section of inelastic neutron scattering. Previously, we showed that it is optimal to use proton beam at energy 2.3 MeV for neutron generation.As a result of a critical analysis of our earlier decisions on the methods used to form a therapeutic neutron beam and decisions of other research groups, as well as successful experiments on the irradiation of laboratory pets and cell cultures carried out at our experimental facility, we noticed that with the recent trend towards a decrease in proton energy the process of inelastic scattering in MgF2 is no longer decisive in neutron moderation, and it was decided to consider materials based on plexiglass as a moderator material.In this work we considered Poly-Biz as moderator material and get the neutron beam the same quality as with MgF2 moderator and proton energy 2.3 MeV but at lower proton energy and current that can cause treatment time reducing and allows more reliable neutron generation.For a long time, the development of BNСT was hindered by the lack of charged particle accelerators that provide stable production of a stationary 2.5 MeV 10 mA proton beam. The use of BSA with Poly-Biz allows the use of these accelerators at lower energy, which improves the reliability of their operation, and reduces the therapy time by more than 2 times, which is also important for therapy. Also, the use of such an BSA simplifies the requirements for a charged particle accelerator and this opens up both the possibility of optimizing the developed accelerators and the use of other accelerators that have not yet reached the required parameters.

Gadolinium Concentration Measurement with an Atomic Absorption Spectrophotometer

Abstract Because gadolinium (Gd) has the highest thermal neutron capture cross section, resulting in an 8 MeV gamma cascade upon capture, it has been proposed for dissolution in water Cherenkov detectors to achieve efficient neutron tagging capabilities. Whereas metallic Gd is insoluble in water, several compounds are very easy to dissolve. Gadolinium sulfate, Gd2(SO4)3, has been thoroughly tested and proposed as the best candidate. Accurate measurement of its concentration, free of doubt from impurities in water, is crucial. An atomic absorption spectrophotometer (AAS) is a device that suits this purpose and is widely used to measure the concentration of many elements. In this study, we describe three different approaches to measure Gd sulfate concentrations in water using an AAS: doping samples with potassium and lanthanum, and employing tantalum and tungsten platforms.

Elastic Cross Section for the Reactions 235U(n, n) and 238U(n, n) at Low Energies Using Woods-Saxon and Coulomb Potentials

The elastic cross section for neutrons by Uranium atoms is calculated for the reactions 235U(n, n) and 238U(n, n) in the regime of low energies using Coulomb andWoods-Saxon potentials. The parameters used in the calculations were chosen by comparison with experimental data previously published by reducing chi-square. Using the same parameters in the equations the ratio 238U/235U is calculated. The results are shown next to the experimental data to illustrate the agreement of the model in the range of energies from 1.5 keV to 1.0 MeV .