Neutron Cross Sections Research Articles

The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport–Thermal Hydraulics Problems

The coarse-mesh finite difference (CMFD) method is commonly used to accelerate the iterative convergence of single-physics neutron transport problems. For multiphysics problems, the neutron cross sections depend on the temperature and density, both of which depend on the fission heat source; the resulting nonlinear feedback can significantly degrade the performance of CMFD and even cause instability. In this paper, we propose, for a class of one-dimensional (1-D) model multiphysics problems, a new nonlinearly implicit low-order (NILO) CMFD (NILO-CMFD) acceleration method to improve the performance of CMFD-based methods for solving loosely coupled multiphysics problems. Our numerical testing and Fourier analysis show that for the 1-D model problems, the new NILO-CMFD method achieves the same rapid convergence rate that CMFD achieves for single-physics problems.

On the simulation of neutron noise induced by vibrations of fuel pins in a fuel assembly

Vibrations of fuel assemblies are an important issue in the safe operation of nuclear reactors, because they can challenge the integrity of the fuel with potential for radioactive releases. Reactor neutron noise-based techniques for monitoring vibrations are valuable for core diagnostic since they are not intrusive and make use of ordinary neutron flux measurements from ex-core and in-core detectors. The application of these techniques involves the solution of inverse problems that require numerical simulations capable of estimating the reactor neutron noise, given a model of the vibrations. For this purpose, several novel reactor neutron noise solvers have been developed in the CORTEX project using either Monte Carlo or deterministic methods, such as the discrete ordinates method, the method of characteristics, and the diffusion approximation. In the current work, these solvers have been scrutinized by computing the neutron noise induced by vibrations of one or multiple fuel pins in a simplified UOX fuel assembly benchmark, via proper variations of macroscopic neutron cross sections. The comparison of these neutron noise solutions obtained from the different methods shows novel insights into the simulation of neutron noise induced by mechanical vibrations, such as the challenges posed by the Monte Carlo method, the impact of the angular discretization on the application of the discrete ordinates method, and the accuracy of the diffusion approximation assessed via the higher-order neutron transport methods.

Silicate glass system and significant role of Bi2O3 on radiation protection ability against gamma, neutron, and charged particle

The use of ionizing radiation and hence, shielding materials has witnessed tremendous diversity in the last three decades. The use of glasses as shields in radiation facilities is gaining attention due to many attractive characteristics of different glass systems. In this study, the gamma-ray, neutron (fast, thermal, and cold), and charged particles’ (electron, proton, alpha particles, and carbon ion) shielding ability of 60SiO2–15ZnO– (25-x)Li2O–xBi2O3 glasses (x = 0 (SZL-00Bi), 5 (SZL-05Bi), 10 (SZL-10Bi), 15 (SZL-15Bi) and 20 (SZL-20Bi) mol%) was examined through the estimation and analysis of interaction parameters. The calculated shielding parameters showed strong dependence on radiation energy and glass composition. The mass attenuation coefficient of the glasses fluctuates between 0.0236 and 0.2223 cm2g-1 for SZL-00Bi, 0.0308–1.7083 cm2g-1 for SZL-05Bi, 0.0343–2.542 cm2g-1 for SZL-10Bi, 0.0360–3.0756 cm2g-1 for SZL-15Bi, and 0.0372–3.4463 cm2g-1 for SZL-20Bi. Also, for photons of energy within 0.1–10 MeV, the estimated mean free path in the SZL-xyBi glasses varies from 1.64 cm to 15.47 cm for SZL-00Bi, 0.17 cm–9.32 cm for SZL-05Bi, 0.10 cm–7.25 cm for SZL-10Bi, 0.07 cm–5.99 cm for SZL-15Bi, and 0.06 cm–5.33 cm for SZL-20Bi. Analysis of other gamma shielding parameters indicated that Bi2O3 enhanced the photon shielding ability of the glasses. The stopping power value of SZL-xyBi glasses for electron, proton, alpha particles and heavy carbon ion varies as follow: SZL-20Bi˂SZL-15Bi˂SZL-10Bi˂SZL-05Bi˂SZL-00Bi. In addition, SZL-00Bi has the highest RCSDA and RP whereas, the SZL-20Bi has the lowest estimated particle range among the studied glasses. The total removal cross section (ΣR) for fast neutrons is 0.0975 cm−1 for SZL-00Bi, 0.0991 cm−1 for SZL-05Bi, 0.0983 cm−1 for SZL-10Bi, 0.1013 cm−1 for SZL-15Bi and 0.1011 cm−1 for SZL-20Bi. The best thermal neutron absorber among the glasses is SZL-05Bi. A comparison of the photon shielding prowess of the present glasses with conventional shields showed the supremacy of the SZL-xyBi glasses. The present glasses thus have great potential as radiation shields.

Total thermal neutron cross section measurements of yttrium hydride from 0.0005 - 3 eV

Yttrium hydride serves as a neutron moderator material that enables compact, high temperature nuclear reactors. However, in order to accurately design and simulate a nuclear system relying upon yttrium hydride, the fundamental nuclear data of yttrium hydride must be well understood. Thermal neutron scattering law (TSL) evaluations represent an important aspect of nuclear data as thermal scattering can drastically alter the neutron multiplication factor of a system. Therefore, to support evaluation and validation of thermal neutron scattering for yttrium hydride, researchers at Rensselaer Polytechnic Institute (RPI) performed total thermal neutron cross section measurements for YH1.68 and YH1.85 over the energy range of 0.0005 - 3 eV. These measurements represent the first total cross section measurements for yttrium hydride that encompass the entire thermal region. Comparisons were made against the ENDF-B/VIII.0, Zerkle & Holmes and Oak Ridge National Laboratory TSL evaluations, where generally good agreement was found.

Study of the 176Hf(n,2n)175Hf cross section at 18.9 MeV

Neutron nuclear reactions can provide significant information in the field of nuclear physics and applications. Hafnium (Hf) is one of the rare-earth isotopes with a relative large neutron total cross-section in the thermal neutron energy region and neutron induced reactions in reactor materials could lead to the production of long-lived isomeric states of Hf isotopes. Thus, the knowledge of neutron cross-sections on Hf isotopes is of great importance for basic research in Nuclear Physics as well as for applications concerning the interaction of neutrons with matter. In this study measurements of experimental cross section for the 176Hf(n,2n)175Hf and 180Hf(n,n΄γ)180mHf reactions were carried out, using the activation technique. The neutron beam energy at 18.9 MeV was produced via the 3H(d,n)4He reaction at the 5.5 MeV Tandem Van de Graaf accelerator laboratory of NCSR “Demokritos”. A thin metallic foil of natural Hf was used, while for the determination of the neutron flux at the target position a reference foil of Al was placed at the back of the Hf target. The irradiation was continuous for 28 hours leading to a total neutron fluence of 1010 n/cm2 and a BF3 detector was used for monitoring the neutron flux during the irradiation. After the end of irradiation, the activity of the Hf target and the Al foil were measured off-line by two HPGe detectors. Both detector efficiencies were obtained using a calibrated 152Eu source, placed in the same distance as the target and the reference foil. The 176Hf(n,2n)175Hf reaction has been corrected from the contribution of 177Hf(n,3n)175Hf and the 174Hf(n,γ)175Hf reactions and the 180Hf(n,n΄γ)180mHf reaction from the 179Hf(n,γ)180mHf. Theoretical calculations of the 174Hf(n,2n)175Hf, 176Hf(n,2n)175Hf and 180Hf(n,n΄γ)180mHf reaction cross-sections have also been performed using the nuclear statistical code “EMPIRE 3.2.3” and they have been compared with the data.

Doppler Broadening of Neutron Cross-Sections Using Kaniadakis Entropy

In the last seven years, Kaniadakis statistics, or -statistics, have been applied in reactor physics to obtain generalized nuclear data, which can encompass, for instance, situations that lie outside thermal equilibrium. In this sense, numerical and analytical solutions were developed for the Doppler broadening function using the -statistics. However, the accuracy and robustness of the developed solutions contemplating the κ distribution can only be appropriately verified if applied inside an official nuclear data processing code to calculate neutron cross-sections. Hence, the present work inserts an analytical solution for the deformed Doppler broadening cross-section inside the nuclear data processing code FRENDY, developed by the Japan Atomic Energy Agency. To do that, we applied a new computational method called the Faddeeva package, developed by MIT, to calculate error functions present in the analytical function. With this deformed solution inserted in the code, we were able to calculate, for the first time, deformed radiative capture cross-section data for four different nuclides. The usage of the Faddeeva package brought more accurate results when compared to other standard packages, reducing the percentage errors in the tail zone in relation to the numerical solution. The deformed cross-section data agreed with the expected behavior compared to the Maxwell–Boltzmann.

Effect of the B4C content on microstructure, microhardness, corrosion, and neutron shielding properties of Al–B4C composites

Aluminum–Boron Carbide (Al–B4C) metal matrix composites (MMCs) are frequently studied because of their high thermal conductivity and ability to absorb neutrons, as an alternative material for the manufacture of baskets in spent fuel casks in the nuclear industry. Keeping in mind, this study aimed to produce Al–B4C MMCs and examine their properties to shield against neutron radiation. The Al-xB4C (x = 5, 10, 15, 20, 25, 30, 40, 45, and 50 wt%) composite powders were prepared by a high-energy planetary ball milling. The prepared powders were compacted with uniaxial cold compaction at 500 MPa and sintered in a tube furnace at 600 °C for 1 h under an Argon atmosphere. The microstructure of the obtained MMCs was characterized by using X-ray diffraction (XRD) and scanning electron microscopy with energy-dispersive X-ray spectroscopy (SEM/EDX). The XRD result indicates that the synthesized Al–B4C MMCs have a predominant phase of Al and B4C. The SEM images show the distribution of B4C reinforcement inside the Al matrix. The relative density, measured by the Archimedes principle decreased from 99.95 to 91.55% as the B4C ratio increased. The microhardness value was increased in the beginning from 44 to 75.5 HV and then decreased to 36.5 HV as the B4C ratio increased. The polarization curves show that the corrosion current density and the corrosion rate were significantly increased from 52.25 to 375.36 μA cm−2 and 0.573–4.121 mm/year as the B4C ratio increased. The simulation results, computed by the MCNP6.2 program show that the thermal neutron cross-section and the fast neutron cross-section was increased from 4.5 to 40.1 cm−1 and 0.13 to 0.15 cm−1 as the B4C ratio increased. Further, the neutron equivalent dose rate was calculated experimentally with the 241Am–Be neutron source and the data was completely agreed with the computational results.

Recycling potential of mobile phone waste glasses for radiation shielding applications

The possible use of mobile phone waste cover glass (WCG) in the fabrication of ionizing radiation glass shield composites is considered in this study. Mixtures of WCG and Bi2O3 with weight proportion for WCG:Bi2O3 of 100:0 (WCG-Bi0), 95:05 (WCG-Bi5), 90:10 (WCG-Bi10), 85:15 (WCG-Bi15), and 80:20 (WCG-Bi20) were prepared using a melt and quench process. The density of the produced glasses was measured, and the Makishima-Mackenzie model was used for predicting the mechanical properties. Furthermore, the charged particles shielding ability of the waste samples was assessed using the ESTAR, PSTAR, ASTAR, and SRIM softwares. The fast and thermal neutron cross-sections were also calculated. The longitudinal, shear, bulk, and elastic modulus were found to vary between 70.86 and 63.53, 30.48–27.33, 48.00–43.03, and 75.47–67.66 MPa for WCG-Bi0 to WCG-Bi20, in that order. The stopping power values of the glass samples (WCG-Bi0 to WCG-Bi20) in regards to β-particles were 1.53–12.47, 1.54–12.83, 1.52–12.54, 1.50–12.26, and 1.48–11.97 MeV cm2/g with increasing Bi2O3 concentration. The shielding ability of the glass was enhanced by the addition of Bi2O3, thereby increasing the potential of using the WCG as an effective neutron and charged particle shield.

241Am Neutron Capture Cross Section in the keV region using Si and Fe-filtered neutron beams

ABSTRACT The neutron capture cross-section of 241Am was measured in the keV neutron range using the recently implemented neutron filtering system of the Accurate Neutron-Nucleus Reaction Measurement Instrument (ANNRI) beamline in the materials and life science (MLF) facility of the Japan Proton Accelerator Research Complex (J-PARC). Filter arrays consisting of 20 cm of nat Fe and nat Si were employed in separate measurements to provide filtered neutron beams with averaged neutron energies of 23.5 (Fe), 51.5 and 127.7 (Si) keV. The present 241Am results were obtained relative to the 197Au neutron capture yield by applying the total energy detection principle together with the pulse-height weighting technique. The 241Am neutron capture cross-section was determined as 2.72(29) b at 23.5 keV, 2.14(26) b at 51.5 keV and 1.32(10) b at 127.7 keV with total uncertainties in the range of 8 to 12%, much lower in comparison to the latest time-of-flight experimental data available.

Synthesis, thermal conductivity, and hydrogen compatibility of a high melt point solid solution uranium carbide, (U0.2Zr0.8)C

Uranium-zirconium carbides, (U,Zr)C, have been previously considered for nuclear thermal propulsion (NTP) reactors due to their high melting point, low neutron cross section, and good hydrogen compatibility. Fuel in NTP reactors would operate under extreme environments: high temperatures (∼3000 K) and under H2 exposure. Despite (U,Zr)C fuels showing promise for extreme environment operation, very little is known about their thermal conductivity as a function of temperature and their H2 compatibility at temperatures over 2500 K. In this work, phase pure UyZr1-yCz with y = 0.1, 0.2, 0.3 and z = 0.8, 0.85, 0.93 was synthesized and high density (U0.2Zr0.8)C samples were fabricated via spark plasma sintering. The thermal diffusivity, specific heat, and thermal expansion of (U0.2Zr0.8)C were measured, from which the thermal conductivity up to 1473 K was calculated for the first time. Furthermore, (U0.2Zr0.8)C samples with different geometries were exposed to H2 at 2600 K for 3 h using the compact fuel element environmental test facility at NASA Marshall Space Flight Center. Both samples performed remarkably well under hot hydrogen attack and no macroscopic cracking was identified. The combination of the relatively high thermal conductivity of these fuels, compared to UO2 for example, which increases with temperature, and their remarkable hydrogen compatibility make them excellent candidates for NTP reactors.

Enhancing the physical, optical and shielding properties for ternary Sb2O3–B2O3–K2O glasses

Ternary Sb2O3–B2O3–K2O glass system, with general composition of x Sb2O3–(70-x) B2O3–30 K2O (where x = 0, 10, 20, 30, 40, 50) were prepared using melt-quenching technique. Structures of these glasses were investigated using XRD analysis and FT-IR spectroscopy. The optical properties were investigated using the UV–Vis NIR JASCO (Model V-670) Double Beam Spectrophotometer. Different physical parameters, such as density (ρ), molar volume (VM), oxygen molar volume (VO) and oxygen packing density values (OPD) have been estimated. Also, the Gamma radiation shielding ability have been characterized for the investigated glasses using Phy-X/PSD software in the photon energy range (0.015–15 MeV). XRD analysis confirm the amorphous nature of the prepared glasses. The optical energy gap of SBK glasses is decreasing from 4 to 2.63 eV with increasing Sb2O3 content while refractive index is increasing from 2.17 to 2.5 because of increasing the non-bridging oxygens (NBOs) in the glass matrix. The molar refractivity (Rm), molar polarizability (αm) and the third-order nonlinear optical susceptibility values χ3 have been calculated, their values are found to increase with increasing Sb2O3 content. Glass density and molar volume values for the SBK glasses increase with increasing Sb2O3 content. Increasing the Sb2O3 concentration enhance the radiation shielding features of the prepared glasses against gamma rays and neutrons. Hence, the mass attenuation coefficient (MAC) increases from 8.6 to 35.4 cm2/gm while the half value layer (HVL) decreases from 0.036 to 0.0046 cm at 0.015 MeV as the Sb2O3 concentration increases from 0 to 50 mol%. The fast neutron cross section of the six investigated SBK glasses are (0.087, 0.092, 0.095, 0.091, 0.094 and 0.092 cm−1), respectively.

Experimental cross section of the 164Dy(n,γ)165Dy reaction at the neutron energy of 0.0372 eV using Neutron Diffraction Facility

Abstract The neutron capture cross section for the 164Dy(n,γ)165Dy reaction at 0.0372 eV energy was measured using the horizontal channel (Neutron Diffraction Facility) of Es-Salam nuclear research reactor in Birine-Algeria. The Neutron Activation Analysis (NAA) technique using 197Au(n,γ)198Au as reference reaction was adopted. The γ ray spectra for the Dy2O3 target powder and two thin gold foil were detected for the Germanium Hyper Pure coupled in the Genie 2000 data acquisition software. A resulting σ-value (2198.5 ± 94) b was measured for the thermal neutron cross section after correcting the thermal neutron self-shielding and γ attenuation effects. The measured value was then compared to the evaluated cross sections reported in two different libraries of neutron databases ENDF/B-VIII.0 and JEFF-3.3. The measured data at 0.0372 eV has been extrapolated to the energy of 0.0253 eV assuming 1/ν dependence in the thermal energy region and the result was compared with the fully measured values reported in the literature.

Neutronic analysis on TRU multi-recycling in PWR-based SMR core loaded with MOX (U-TRU) and FCM (TRU) fueled assemblies

An analysis on TRU multi-recycling in a light water-cooled SMR core loaded with MOX-FCM fuel assemblies employing MOX and FCM (Fully Ceramic Micro-encapsulated) fuel rods was performed for TRU transmutation. We used a two-step procedure composed of McCARD/MASTER where the assembly-wise two group homogenized neutron cross sections are generated using a Monte Carlo depletion code, while the core analysis was performed using a two-group nodal diffusion code. The reload core calculation based on three-batch refueling scheme was conducted using a TRU recycling scheme in which TRU in MOX fuel rods are multi-recycled but TRUs in FCM fuel ones are sent to the final repository after the discharge without recycling. An overall TRU mass balance analysis through the 5th to 20th cycle showed that 21.1% of external TRU feed can be consumed by utilizing the light water-cooled SMR core without considerable degradation of safety-related parameters.

A feasibility study of SMART reactor power performance optimizations-part 2: Reflector material selection

The neutron reflector is a general component in the nuclear reactor design. The original SMART reactor design used light water as a reflector surrounding the fission zone. However, this design has low uranium utilization rates in the outermost fuel assemblies, so poor fuel economy. In this study, six potential reflector materials, i.e., heavy water, graphite, beryllium metal and its oxide, steel, and tungsten carbide has been investigated as the neutron reflector for SMART. Firstly, the materials’ cross-sections of neutron scattering and (n, 2n) reactions have been cited from ENDF data libraries and analyzed. This analysis found for the neutrons with energy lower than 1 eV, the 9Be atom has six barn elastic scattering cross-sections, and the 9Be(n, 2n)8Be reaction can compensate for neutron leakage. Then, OpenMC is employed to simulate the effect of these reflector materials on power distribution and depletion. Compared with the original design, the beryllium oxide can improve the initial keff from 1.22906 to 1.27446, flat the radial direction power distribution. The analysis on both scales agrees that the beryllium oxide is an efficient neutron reflector choice with good material properties.

Prospects for Neutron Reactions on Excited States in High-Density Plasmas

With the reactions of high flux neutrons, such as in a DT plasma, there is a prospect of seeing new kinds of neutron-nucleus reactions for the first time. If neutrons excite a heavy nucleus, for example, there is possibility of a second neutron reacting on excited states of the residual nucleus before that nucleus has de-excited to its ground state. The possibility of such reactions on excited states has rarely been considered. The cross section for neutron induced fission on the isomeric state of 235U has been measured (D’Eer et al., Phys. Rev. C, 1988, 38: 1270–1276) and calculated (Younes et al., 2003, Maslov, 2007), and reactions on rotationally excited nuclear states has been calculated (Kawano et al., Phys. Rev. C, 2009, 80: 024611). In high flux plasmas, however, a much wider range of reactions is possible. We therefore need to consider excited states at much higher-energies than previously modeled, and then estimate whether second neutrons are likely to rescatter on those excited states. To determine the likelihood of such rescattering events, we first need to know the probable time series of nuclear decays of those excited states. The lifetimes of many low-lying states have been measured experimentally, but now we need to know the lifetimes of the many higher excited states that could be produced from incident 14 MeV neutrons. These are too numerous to be measured and also too numerous to be calculated individually, so statistical Hauser-Feshbach decay models are used. I show some lifetime calculations for 89Y, 169Tm, and 197Au targets, and predictions for the number of rescattering events in plausible plasma scenarios.

Gamma and Neutron Shielding Parameters of Polyester-based composites reinforced with boron and tin nanopowders

The usage of composites as the shielding materials are highly recommended since they could be used in order to attenuate the undesired radiation with unique properties and advantages in the areas where the radiation is prevalent. In this context, not only are their radiation shielding properties important but also their flexibility, durability and low cost. Due to the mentioned superior characteristics, the polyester based composites are among the most preferred materials. With the aim of creating unique and novel radiation shielding materials, this study investigates gamma and neutron shielding capabilities of the polyester composites reinforced with Boron and Tin nanopowders at different proportions (0–50%, 10–40%, 20–30%, 30-20% and 40-10%, 50-0%). The gamma shielding abilities of the prepared polyester composite materials were evaluated using an HPGe detector system, WinXCOM computer program and different simulation tools (FLUKA and GEANT4) at the energies varying from 59.5 to 1332.5 keV. The experimental, theoretical and simulation results showed remarkable agreement between each other, and the addition of Sn enhances the gamma attenuation performance of the chosen polyester composite materials. In addition to gamma analysis results, neutron shielding properties of the proposed composites are further determined. On this purpose, the transmitted neutron numbers through the samples (as functions of neutron energy and the sample thickness) and effective neutron removal cross sections were evaluated. The neutron shielding performance of the samples showed that the prepared composites could be alternative materials to the existing neutron shields in the literature.

Enhancement on radiation shielding performance of B2O3 + Li2O + ZnO + Na2O glass system

Three borate glasses with the chemical composition 50B2O3 + xLi2O + 20ZnO + (30 - x) Na2O (where x = 5, 10, & 20 mol. %) and coded as BLZN-A, BLZN-B, and BLZN-C were reported and investigated for their gamma-ray, charged particle (electrons, protons, alpha particles and carbon ions), and neutrons shielding viability. The gamma ray attenuation coefficients were computed by the FLUKA simulations and XCOM methods, while the stopping powers and range of the particles were evaluated based on SRIM software and NIST database. The obtained results showed that gamma-ray and charged particle shielding ability increases for the present glass samples in the order of BLZN-A < BLZN-B < BLZN-C for radiation energies of 0.01–10 MeV. The removal cross section corresponds to 0.106, 0.1087, and 0.1132 cm−1 for BLZN-A, BLZN-B, and BLZN-C, while the total cross section for thermal neutron is 69.91, 71.46, and 73.83 cm−1. The shielding performance of the glasses is higher than those of some conventional shields such as ordinary concrete and commercial glass shields. Thus, the present BLZN glass system have huge potential for practical applications in hospitals and nuclear facilities.

Accelerator mass spectrometry measurements of 93Zr for astrophysical and nuclear technology applications

Zirconium-93 is a long-lived radionuclide with a half-life of (1.61 ± 0.05) × 106 yr. Production cross sections for 93Zr by neutron capture on stable 92Zr in the keV and thermal energy ranges are important input parameters for astrophysical network calculations and nuclear industry, respectively. Despite their importance, existing experimental data suffer from large uncertainties. Here, the combination of neutron activation and accelerator mass spectrometry (AMS) is presented as an alternative method to online time-of-flight measurements for the determination of these cross sections. The main challenges for AMS of 93Zr are the interference from the stable isobar 93Nb and the production of suitable reference material. At the Heavy Ion Accelerator Facility (HIAF) the first challenge is tackled with the available high particle energies and by investigating different Zr compounds and extracted molecular beams. Using ZrF5− beams extracted from ZrF4 sample material, it is shown that the Nb background can be reduced by up to two orders of magnitude relative to the extraction of ZrO− beams from ZrO2. Using the 13+ charge state and particle energies of ∼190 MeV, 93Zr/Zr background levels in the 10−12 range are regularly achieved at HIAF. The ZrF5− output from ZrF4 samples may be highly variable but can be stabilised using an intimate mixture of ZrF4 and PbF2 at a ratio of 1:10 that has been dried down from a concentrated hydrofluoric acid solution. An option to produce a well-characterised 93Zr reference material by utilising the well-known 235U-fission yield of 93Zr is discussed. Once this remaining challenge is resolved, accurate measurements of the astrophysically and technologically relevant neutron-capture cross sections of 92Zr using AMS will be possible at HIAF.

Measurement of neutron-capture cross sections of Ge70,72 using the DANCE facility

Background: Approximately half of all atomic nuclei heavier than iron are synthesized by the slow neutron-capture process. The weak component of this process is not well understood and the reaction rates of each isotope in the s-process path affect nucleosynthesis abundances downstream.Purpose: To measure the neutron-capture cross sections of two weak s-process nuclei, $^{70,72}\mathrm{Ge}$, using the neutron time-of-flight technique. Measuring the capture cross sections for isotopes in this region of the chart of nuclides has proven challenging due to dominant scattering cross sections.Method: Samples consisted of pellets made of pressed enriched metallic powders. The $^{70,72}\mathrm{Ge}$ neutron-capture cross sections were measured as a function of neutron energy using the Detector for Advanced Neutron Capture Experiments at Los Alamos National Laboratory.Results: Neutron-capture cross sections were measured from 10 eV to 1 MeV. These are the first measurements for $^{70,72}\mathrm{Ge}$ between 300 keV and 1 MeV neutron energy. Maxwellian-averaged cross sections were calculated in the astrophysically relevant neutron energy range (5 keV $\ensuremath{\le}kT\ensuremath{\le}$ 100 keV). Their value at $kT=30$ keV was found to be $89\ifmmode\pm\else\textpm\fi{}11$ mb for $^{70}\mathrm{Ge}$ and $58\ifmmode\pm\else\textpm\fi{}5$ mb for $^{72}\mathrm{Ge}$. Both values are in agreement with recent time-of-flight measurements at n_TOF (neutron Time-Of-Flight facility at the European Organization for Nuclear Research).Conclusions: The average cross section results from this work for $^{70}\mathrm{Ge}$ show minor ($&lt;1\ensuremath{\sigma}$) disagreement with a recent measurement by the n_TOF collaboration at higher neutron energies. This corresponds to the neutron energy region that had previously never been measured ($&gt;300$ keV). Two reaction library databases underestimate the $^{72}\mathrm{Ge}$ average cross section below 30 keV according to n_TOF and DANCE. This is likely due to capture resonances that are missing from the theoretical cross sections in the databases that were identified in both time-of-flight measurements. Additionally, a rudimentary analysis of the impact of both cross section measurements on stellar nucleosynthesis abundances using the NETZ nucleosynthesis tool is presented.

In vivo evaluation of the effects of combined boron and gadolinium neutron capture therapy in mouse models

While boron neutron capture therapy (BNCT) depends primarily on the short flight range of the alpha particles emitted by the boron neutron capture reaction, gadolinium neutron capture therapy (GdNCT) mainly relies on gamma rays and Auger electrons released by the gadolinium neutron capture reaction. BNCT and GdNCT can be complementary in tumor therapy. Here, we studied the combined effects of BNCT and GdNCT when boron and gadolinium compounds were co-injected, followed by thermal neutron irradiation, and compared these effects with those of the single therapies. In cytotoxicity studies, some additive effects (32‒43%) were observed when CT26 cells were treated with both boron- and gadolinium-encapsulated PEGylated liposomes (B- and Gd-liposomes) compared to the single treatments. The tumor-suppressive effect was greater when BNCT was followed by GdNCT at an interval of 10 days rather than vice versa. However, tumor suppression with co-injection of B- and Gd-liposomes into tumor-bearing mice followed by neutron beam irradiation was comparable to that observed with Gd-liposome-only treatment but lower than B-liposome-only injection. No additive effect was observed with the combination of BNCT and GdNCT, which could be due to the shielding effect of gadolinium against thermal neutrons because of its overwhelmingly large thermal neutron cross section.