Neutron Cross Sections Research Articles

Thermal cross section of poly(2-hydroxyethyl methacrylate) hydrogels for neutron dose calculation

Phantom materials are used to design radiation safety and protection strategies by means of numerical dose calculation. In the case of thermal neutrons, the radiation transport in the phantom relies on mass attenuation coefficients and total cross sections which are dependent on the physical and chemical properties of the material. Specifically for medical applications, such as neutron capture therapy, the neutron-induced dose is mainly related to the neutron absorption by hydrogen and nitrogen in the human tissue and body. Here, we investigate the use of poly(2-hydroxyethyl methacrylate) as a phantom material by experimental measurement and modeling of its total neutron scattering cross section and mass attenuation coefficient. We show that by varying the hydration level from 10 to 40 w%, one can obtain a neutron attenuation coefficient similar to polymethyl methacrylate or more representative of the human body, respectively. By benchmarking our phenomenological model on the experimental data, we provide a new example of how to use the average functional group approximation to accurately model total neutron cross sections in the framework of personalized medicine approaches.

Production and evaluation of the polyester composites containing pyrite and niobium diboride for radiation protection

The presented study focused on investigating the radiation shielding properties of polyester/pyrite/niobium diboride (FeSNbB) composites. Various gamma-ray shielding parameters were evaluated with help of a HPGe detector system. Theoretical and simulation results are further evaluated by WinXCOM, GEANT4 and FLUKA codes, and the obtained results are compared with experimental values in the energy region 59.5 keV–1332.5 keV. The prepared samples are additionally evaluated using kerma relative to air values and Exposure build-up factors. Then, for the evaluation of neutron radiation shielding, effective removal cross sections for fast neutrons were determined for FeSNbB composites. Electron attenuation efficiency (AE%) for the produced composites was assessed across the energy spectrum of 4 MeV–18 MeV, utilizing experimental data from a Linac machine and theoretical predictions from Eclipse TPS. The results were found to be compatible with each other. FeSNbB50 has the best shielding properties among the studied composites.

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

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

A Consistent Comparison of Upper Subcritical Limit Methods

This study conducts a consistent and comprehensive comparison of several methods for estimating upper subcritical limits (USLs) for nuclear criticality safety analysis. To eliminate inconsistency caused by the discrepancy between the estimated covariance data and the true degree of uncertainty present in nuclear data, the experimental system eigenvalues k exp were assumed to equal the calculated eigenvalue for the systems using nominal ENDF/B-VII.1 cross sections, and the estimated calculated eigenvalue k calc was assumed to equal the calculated eigenvalue for the systems using nuclear data from one perturbed cross-section library. USLs are estimated for a variety of validation application cases using the Whisper, TSURFER, and USLSTATS tools and are compared to a reference 95/95 limit. The TSURFER approach produced the strongest agreement with the reference USLs; the Whisper produced USLs that were reliably conservative compared to the reference USL; and the USLSTATS approach experienced difficulty consistently producing accurate USL estimates. Last, this study observed a noteworthy discrepancy when using Cholesky decomposition to randomly sample neutron cross-section covariance data that are small in magnitude.

Exploring the Radiation Shielding Efficiency of High-Density Aluminosilicate Glasses and Low-Calcium SCMs

A lot of work is now going into making low-calcium supplementary cementitious materials (SCMs) and aluminosilicate glasses so that they can be used as radiation shielding materials. These materials demonstrate superior performance in several aspects as compared to conventional concrete. The present investigation focuses on the radiation shielding characteristics of the evaluated materials, specifically their capacity to reduce the intensity of gamma rays and neutrons. Regarded for their exceptional density and ability to include heavy metal oxides, aluminosilicate glasses have remarkable shielding characteristics, especially when designed at the molecular scale. An evaluation of the performance of these materials in comparison to traditional concrete is carried out using Phy-X/PSD software. The goal is to determine the most important shielding properties, such as the mass attenuation coefficient, the linear attenuation coefficient, and the half-value layer. Our study findings suggest that some aluminosilicate glasses, such as GM, consistently demonstrate exceptional photon and neutron attenuation efficiency. The observation that GM performs better than other materials in tests like effective atomic number, rapid neutron removal cross section, and energy absorption accumulation factor supports this claim. There is evidence that using low-calcium glass-crystal materials (SCMs) with aluminosilicate glasses not only improves radiation protection but also makes solutions work better when space or weight are limited. The present investigation validates that these materials exhibit superior performance compared to conventional concrete in challenging environments such as nuclear waste storage, where safety is of utmost significance.

Generating a system of group constants for neutron-physical calculations of fast reactors based on ROSFOND-2020.2 library files

The subject of constant support development is currently becoming increasingly important in connection with the national strategy of transition to a closed nuclear fuel cycle. At the same time, the importance of the tasks of refining calculation models and minimizing methodological, statistical, and constant errors is increasing. In this connection, the idea of creating a universal system of constants describing with equal accuracy both uranium loading and loading with mixed oxide uranium-plutonium fuel and the possibility to perform on its basis both precision Monte Carlo calculations (data format – ACE) and multigroup calculations (formats – ABBN and ANISN) was laid in the basis of this work. In this paper we analyze the freely available libraries of evaluated nuclear data in order to justify the choice of source files for the formation of a new system of group constants for neutron-physical calculations of fast reactor cores, describe the process of formation of a new library of reactor constants, and verify the obtained system on computational test models of BN-800 reactor and critical systems. The process of formation of the new system of group constants included selection of initial neutron data files, updating of data tables of basic neutron cross-sections, self-shielding factors and Doppler coefficients, as well as data on fission spectra for the main fuel nuclides. The methodological component of the error for test models of the BN-800 core was evaluated. Using the ABBN-RF22 system of group constants it was possible to estimate the correction in the 299 groups calculation, which amounted to 0.3%. Previously, when using the ABBN-93 library, there was no such possibility due to the lack of continuity between the files of evaluated neutron data and the group constants used.

Effect of low concentrations of WO3 on synthesis, structural, physical, optical, neutron attenuation, and charged particle absorption properties of B2O3+Na2O+BaO glassy composites

This study presents the physical, optical, and radiation shielding attributes of 75B2O3+5Na2O+(20-x)BaO+xWO3 glasses, with the aim of producing effective transparent and light shields. The well-known melting-and-quenching technique was used to prepare the transparent glasses for x = 0; 03; 10 mol%). The glasses were characterized by determining their density, molar volume, optical transmittance, bandgap, refractive index, dielectric constant, reflection loss, molar refraction charged radiation stopping powers, projected range, fast neutron removal cross section and thermal neutron total cross-section. All the estimated parameters showed variations with glass stoichiometry. The addition of WO3 triggered the BO4 → BO3 unit transformation which prompted the variations in the physical and optical parameters of the glasses. The optical transparency in the visible region and density of the glasses decreased with WO3 concentration while the molar volume increased. For electrons, the stopping powers were in the range, 1.436–11.76 cm2/g, 1.433–11.66 cm2/g, and 1.426–11.44 cm2/g, for BNB-W00, BNB-W03, and BNB-W10, respectively. For protons, alpha particles, and carbon ions, the mass stopping powers increased with WO3 content. Based on projected range data, particles with greater energy per nucleon penetrated the glasses deeper and the WO3- rich glasses had higher transmissivity for charged radiation. Compared to some existing neutron moderators, the investigated glasses displayed higher fast neutron cross-sections. The studied glasses can function well as barriers to moderate fast neutrons, attenuate thermal neutrons, and absorbed charged radiation.

Lead fluoride as a multifunctional modifier in bismuth gadolinium borate glasses: Optical, structural, and radiation attenuation characterization

This study systematically evaluates the role of PbF2 in altering the optical and radiation shielding properties of bismuth gadolinium borate glasses with compositions 30Bi2O3–Gd2O3-xPbF2-(70-x)B2O3, for x ranging from 0 to 40 mol. %. The introduction of PbF2 significantly modified the absorbance profile, exhibiting distinct absorption peaks at 3.45, 4.15, and 4.69 eV, shifting from a flat curve in the undoped sample to well-defined peaks with increasing PbF2 content. Crucially, a reduction in both the direct and indirect optical band gaps was observed as the PbF2 content increased, indicating changes in the electronic structure of the glasses. Photon and neutron shielding capabilities were analyzed, revealing that the linear attenuation coefficient (μ), half-value layer (T0.5), and effective atomic number (Zeff) were enhanced with PbF2 additions. Notably, the BGPB5 sample contained 40 mol. % PbF2 outperformed other compositions, offering the highest attenuation efficiency, which surpassed some conventional glasses and concretes. Simultaneously, an inversely proportional relationship was found between PbF2 concentration and the effective removal cross-section for fast neutrons (ΣR), demonstrating the BGPB1 sample's superiority as a neutron shield. The incorporation of PbF2 in the glass matrix not only improved gamma radiation shielding but also tuned the optical properties, marking these glasses as potential candidates for multifunctional radiation shielding applications.

Data and modeling sensitivity analysis for molten salt fast reactor benchmark – Static calculations

The Molten Salt Reactor (MSR) idea is increasingly being recognized in the nuclear field due to its potential safety, sustainability, and economic efficiency advantages. The Molten Salt Fast Reactor (MSFR) benchmark, introduced in 2019, highlighted variations in results tied to different neutron cross-section libraries. This study investigates the impact of utilizing the ENDF/B-VIII.0 and JEFF-3.3 cross-section libraries for MSFR benchmark assessment compared to the ENDF/B-VII.1 database. Monte Carlo based open source code OpenMC is used for the analyses. Rigorous sensitivity analyses assess the influence of individual components, including the cross-section database, resonance elastic scattering, and Thermal Scattering Law (TSL). Beyond the criticality assessments, parameters such as delayed neutron fraction, temperature coefficient of reactivity, and neutron spectrum are compared for different cross-section libraries. Our analyses reveal that incorporating new evaluations for 233U (n,γ) and fission cross-sections in ENDF/B-VIII.0 significantly alters criticality results, i.e., more than 1700 pcm difference is seen between libraries. Similarly, critical concentration using ENDF/B-VII.1 and JEFF-3.3 is over-predicted by approximately 3%. The variations in Thermal Scattering Law (TSL) files do not yield substantial differences in outcomes due to the fast spectrum of the reactor. In some cases, the treatment of resonance elastic scattering leads to reactivity differences greater than 50 pcm. The benchmark compares 233U-started and Minor Actinide (MA)-started core. From the reactor physics point of view, the MA-started core leads to a 29% higher (n, γ) reaction rate than the 233U-started core. A 3–4% smaller value of thermal reactivity coefficient is obtained using the ENDF/B-VIII.0 library compared to the ENDF/B-VII.1 value. Using the ENDF/B-VIII.0 for the MSFR benchmark signifies using newer and better data for the GEN-IV reactors neutron physics calculations.

Structural, thermal, and radiation shielding properties of B2O3-TeO2-Bi2O3-CdO-Tm2O3 glasses: The role of Tm2O3

This study presents a detailed investigation into the structural, thermal, and radiation shielding properties of a new glass composition, labeled as (50-x)B2O3 + 30TeO2 + 10Bi2O3 + 10CdO + x Tm2O3 (where x = 0, 1, 2, 3, 4, 5 mol %). Various radiation shielding parameters such as mass attenuation coefficient, linear attenuation coefficient, half-value layer, mean free path, effective atomic number, and absorbed equivalent dose rates for neutrons were determined through experimental measurements. The study also employed theoretical calculations to assess the exposure buildup factor and effective removal cross-section for neutrons. Additionally, the SRIM program was utilized to investigate mass stopping power and projected range for proton and alpha particles. The structural characteristics of the glasses were analyzed using XRD and FT-IR, while thermal properties were examined through Differential Thermal Analysis (DTA). FTIR analysis revealed distinctive bands corresponding to Bi—O, B—O—B, and Te—O bonds. Glass transition temperatures (Tg) increased with Tm2O3 content, ranging from 422 °C to 444 °C. The radiation shielding properties of bismuth boro-tellurite glasses showed that a dose of 1.2025 μSv/h emitted from a fast neutron source was absorbed by 35.1574 % in pure glass and 40.0391 % in glass doped with 5 mol% Tm2O3. Incorporating Tm2O3 into the glass matrix significantly improved the shielding and thermal properties, thus enhancing their potential for advanced high-energy radiation absorption. This investigation contributes to a deeper understanding of how Tm2O3 enhances the structural and thermal attributes of these glasses, positioning them as promising materials for radiation shielding applications.

A new hybrid gadolinium nanoparticles-loaded polymeric material for neutron detection in rare event searches

Experiments aimed at direct searches for WIMP dark matter require highly effective reduction of backgrounds and control of any residual radioactive contamination. In particular, neutrons interacting with atomic nuclei represent an important class of backgrounds due to the expected similarity of a WIMP-nucleon interaction, so that such experiments often feature a dedicated neutron detector surrounding the active target volume. In the context of the development of DarkSide-20k detector at INFN Gran Sasso National Laboratory (LNGS), several R&D projects were conceived and developed for the creation of a new hybrid material rich in both hydrogen and gadolinium nuclei to be employed as an essential element of the neutron detector. Thanks to its very high cross-section for neutron capture, gadolinium is one of the most widely used elements in neutron detectors, while the hydrogen-rich material is instrumental in efficiently moderating the neutrons. In this paper results from one of the R&Ds are presented. In this effort the new hybrid material was obtained as a poly(methyl methacrylate) (PMMA) matrix, loaded with gadolinium oxide in the form of nanoparticles. We describe its realization, including all phases of design, purification, construction, characterization, and determination of mechanical properties of the new material.

Effect of Angular Momentum–Dependent Transparency on the Neutron-Nucleus Cross Section

We study the effect of the angular momentum dependence of the optical potential transparency on the neutron-nucleus cross section. In this work, we derive a functional form of the neutron-nucleus cross section with respect to spherical optical potential using the collision matrix and modify it. We also apply an analytical model (Ramsauer) to the square optical potential and the optical model using the TALYS 1.96 code to get the angular momentum–independent transparency and demonstrate the effect of the angular momentum dependence via comparison of two models. In this work, we calculate the neutron cross sections for 40 ≤ A ≤ 90 nuclei, with energies 100 MeV ≤ En ≤ 200MeV theoretically, the for ℓ ≠ 0 case. We find that the angular momentum dependence of the transparency treats good calculations of the neutron-nucleus cross section; a finite number of angular momentums will contribute appreciably to the nuclear reaction. The present study on smooth optical potential leads to useful insight into the mechanisms of neutron-induced reactions, particularly for medium nuclei at high energies.

Shedding Light on the Origin of ^{204}Pb, the Heaviest s-Process-Only Isotope in the Solar System.

Asymptotic giant branch stars are responsible for the production of most of the heavy isotopes beyond Sr observed in the solar system. Among them, isotopes shielded from the r-process contribution by their stable isobars are defined as s-only nuclei. For a long time the abundance of ^{204}Pb, the heaviest s-only isotope, has been a topic of debate because state-of-the-art stellar models appeared to systematically underestimate its solar abundance. Besides the impact of uncertainties from stellar models and galactic chemical evolution simulations, this discrepancy was further obscured by rather divergent theoretical estimates for the neutron capture cross section of its radioactive precursor in the neutron-capture flow, ^{204}Tl (t_{1/2}=3.78 yr), and by the lack of experimental data on this reaction. We present the first ever neutron capture measurement on ^{204}Tl, conducted at the CERN neutron time-of-flight facility n_TOF, employing a sample of only 9mg of ^{204}Tl produced at the Institute Laue Langevin high flux reactor. By complementing our new results with semiempirical calculations we obtained, at the s-process temperatures of kT≈8 keV and kT≈30 keV, Maxwellian-averaged cross sections (MACS) of 580(168)mb and 260(90)mb, respectively. These figures are about 3% lower and 20% higher than the corresponding values widely used in astrophysical calculations, which were based only on theoretical calculations. By using the new ^{204}Tl MACS, the uncertainty arising from the ^{204}Tl(n,γ) cross section on the s-process abundance of ^{204}Pb has been reduced from ∼30% down to +8%/-6%, and the s-process calculations are in agreement with the latest solar system abundance of ^{204}Pb reported by K. Lodders in 2021.

Physical, chemical and radiation shielding properties of metakaolin-based geopolymers containing borosilicate waste glass

Upcycling waste borosilicate glass (BS) is a viable and sustainable way of preserving the natural resources required for the production of new materials and ensuring environmental sanitation and safety. This study was aimed at investigating how the addition of BS influences the physical, chemical, and mechanical properties and shielding performance against neutrons and charged particles on metakaolin-based geopolymer G. Four batches of G infused with 10%, 20%, and 30% wt BS samples and coded as G, G-10BS, G-20BS, and G-30BS, respectively, were fabricated based on the solid-state reaction method. The samples were characterized accordingly using experimental and theoretical procedures. X-ray diffraction analysis of the materials showed that the addition of BS suppressed crystal peaks in the geopolymers. The Vickers hardness values were calculated as 554, 503, 517, and 528 HV under a 0.1 kg load. The values of fast neutron macroscopic cross-section increased from 0.0601 cm−1 to 0.0706 cm−1 as BS content increased from 0 to 30 wt%. The coherent, incoherent, absorption, and total interaction probabilities of thermal neutrons increased in the geopolymer samples. For electrons, the stopping powers were in the range of 1.575–13.17 cm2/g, 1.571–13.15 cm2/g, 1.571–13.16 cm2/g, and 1.571–13.18 cm2/g for G, G-10BS, G-20BS, and G-30BS, respectively. The projected ranges of carbon ions in G, G-10BS, G-20BS, and G-30BS are within the limits 0.069–17.87 μm, 0.064–16.99 μm, 0.064–16.99 μm, and 0.057–15.24 μm. The trend of the projected ranges of electrons, protons, α-particles, and C ions with kinetic energies within 15 keV and 15 MeV was mostly in the sequence G > G-10BS > G-20BS > G-30BS. This order was consistent with the density and BS content of the geopolymer samples. The addition of BS to the metakaolin-based geopolymer improved the fast neutron moderating capacity and thermal neutron cross-sections and reduced the range of charged particle radiation in the geopolymers. The neutron shielding ability of G-30BS was found to be better than that of some existing shielding materials. The BS-infused concrete can be useful for the production of concrete used for radiation control.

Prompt gamma activation analysis for boron determination in the tens of milligram range at the HOTNES facility

Boron is an elective element for the Prompt Gamma Activation Analysis (PGAA), due to its exceptionally large neutron capture cross section. This technique, usually performed in nuclear reactors with neutron fluxes as high as 108 cm−2 s−1, can determine quantities of boron as low as tens of nanograms. Some applications, such as the industry of neutron shielding materials, would better benefit from a less sensitive but more portable and accessible boron PGAA, which could be established at construction or fabrication sites. For these purposes ENEA and INFN jointly setup a compact PGAA based on a 0.5 cm3 Cadmium–Zinc–Telluride gamma spectrometer and the HOTNES thermal neutron source. Relying on a series of borated resins with known composition and on comprehensive experimental and Monte Carlo evaluations, this technique features a detection limit in the order of few milligrams in terms of boron mass. As the facility consists simply on a lab-scale neutron source and a polyethylene block with well-established geometry, this simplified PGAA system is suited to be replicated or transported to construction or fabrication sites for QA/QC purposes on borated construction materials for the nuclear sector.

Development and Verification of Sampling Code NECP-SOUL for Evaluated Nuclear Data Files in ENDF-6 Format

A code called NECP-SOUL is developed to generate samples of evaluated nuclear data files in the ENDF-6 format. The motivation for the development is to obtain more accurate uncertainty quantification results based on the covariance data given in evaluated nuclear data files than the sampling method based on multigroup covariances. NECP-SOUL can handle all covariance data in ENDF-6 files, including average fission neutron multiplicities, resonance parameters, cross sections, angular distributions, energy distributions, and cross sections for the production of radioactive nuclides. In this work, samples of ENDF-6 files are input to the nuclear data processing code to generate A Compact ENDF (ACE)-formatted libraries, and these ACE-formatted libraries are used by the Monte Carlo code. Through statistical analysis, the uncertainty of the neutronics results can be obtained. NECP-SOUL is verified against another similar code, SANDY. In addition, the advantage of sampling ENDF-6 files is analyzed by comparing the uncertainty quantification results with those obtained by sampling multigroup covariance matrices. Moreover, the covariances of cross sections and angular distributions newly provided in ENDF/B-VIII.0 are used to analyze their effect on the uncertainty qualification results.

Lithium Neutron Cross Sections During the Manhattan Project and the Quest for the H-Bomb

Neutron cross sections of the stable lithium isotopes 6Li and 7Li were of interest in the 1940s and 1950s in part because of their reactions, which form tritium using moderated neutrons on 6Li and higher-energy neutrons on either isotope. Lithium remains of interest today for use as a blanket and shielding material in fusion reactors, where it can be used to breed tritium for a self-sustaining fuel cycle. During the Manhattan Project, the resonance in the 6Li(n,t) reaction was discovered and later became important for enhancing tritium production for nuclear technologies. The dominant natural isotope 7Li was and remains of interest because of the expense of enriching 6Li. It has been oft reported that the 1954 Castle Bravo nuclear test had a yield twice as large as expected because the nuclear explosive device designers had not properly accounted for the benefits from the 7Li isotope in the fuel; we note that this explanation is false.

In-situ investigation of helium effect on diverse precipitate/matrix interface areas in irradiated 5052 aluminum alloy

Due to its low neutron cross-section and high irradiation swelling resistance, 5052 aluminum alloy is extensively utilized in the core structure and irradiation shielding of research nuclear reactors. Indeed, irradiation-induced defects may accumulate preferentially in the precipitates of aluminum alloys during service, leading to prior impairment of these precipitates, thereby inducing the degradation of mechanical properties. The interaction between irradiation defects and precipitates under irradiation, however, remains inadequately understood. In this work, the microstructure of 5052 aluminum alloy irradiated in-situ with He+ at different temperatures and subsequent annealing was investigated by transmission electron microscopy. After 1.0 dpa irradiation at different temperatures, the evolution of He bubbles within the precipitate was greater than in the matrix, and manifested a pronounced growth threshold between 373 K and 473 K. During post-irradiation annealing, compared to the matrix, also the dynamic evolution of bubbles was observed on the inner side of the precipitate interface, where bubbles coalesced and rapidly dissolved. The dynamic behavior of He bubbles within different precipitates manifested a certain discrepancy, and sink strength of two typical Al-containing precipitates was estimated. The difference in bubble evolution can be attributed to the different sink efficiency and vacancy concentration gradient around the precipitate.

Actinium-225 photonuclear production in nuclear reactors using a mixed radium-226 and gadolinium-157 target

BackgroundActinium-225 is one of the most promising radionuclides for targeted alpha therapy. Its limited availability significantly restricts clinical trials and potential applications of 225Ac-based radiopharmaceuticals. MethodsIn this work, we examine the possibility of 225Ac production from the thermal neutron flux of a nuclear reactor. For this purpose, a target consisting of 1.4 mg of 226Ra(NO3)2 (T1/2 = 1600 years) and 115.5 mg of 90 % enriched, stable 157Gd2O3 was irradiated for 48 h in the Breazeale Nuclear Reactor with an average neutron flux of 1.7·1013 cm−2·s−1. Gadolinium-157 has one of the highest thermal neutron capture cross sections of 0.25 Mb, and its neutron capture results in emission of high-energy, prompt γ-photons. Emitted γ-photons interact with 226Ra to produce 225Ra according to the 226Ra(γ, n)225Ra reaction. Gadolinium debulking and separation of undesirable, co-produced 227Ac from 225Ra was achieved in one step by using 60 g of branched DGA resin. After 225Ac ingrowth from 225Ra (T1/2 = 14.8 d), 225Ac was extracted from the 226Ra and 225Ra fraction using 5 g of bDGA resin and then eluted using 5 mM HNO3. ResultsMeasured activity of 225Ac showed that 6(1) kBq or 0.16(3) μCi (1σ) of 225Ra was produced at the end of bombardment from 0.9 mg of 226Ra. ConclusionThe developed 225Ac separation is a waste-free process which can be used to obtain pure 225Ac in a nuclear reactor.

Effects of shape/phase transition regions on neutron capture cross sections

A recent study found a new, purely empirical correlation between two-neutron separation energies and neutron capture cross sections in keV neutron energy regimes. In shape/phase transition regimes, such as that near A= 150, S2n\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_{2n}$$\\end{document} values show an anomaly—a flattening of the normal near linear decrease with neutron number. This paper addresses two questions: (1) Using this new correlation, is this anomaly in S2n\\documentclass[12pt]{minimal} \\usepackage{amsmath} \\usepackage{wasysym} \\usepackage{amsfonts} \\usepackage{amssymb} \\usepackage{amsbsy} \\usepackage{mathrsfs} \\usepackage{upgreek} \\setlength{\\oddsidemargin}{-69pt} \\begin{document}$$_{2n}$$\\end{document} values sizeable enough to produce an observable effect in capture cross sections? and (2) Can the correlation be used to quantitatively reproduce the cross sections in the transition region? It is found that the answer to both questions is in the affirmative. Possible relations to the r-process are briefly discussed.