Neutron Activation Cross Sections Research Articles

Characterization of a HPGe detector response for activation cross section measurements: regression method versus Monte Carlo method

ABSTRACT The full energy peak efficiencies and covariances of a high purity germanium (HPGe) detector studied by the regression method and Monte Carlo method were compared. The impact of the obtained results on the neutron activation cross-sections measured relative to monitor cross-sections was studied. In the regression method, the efficiencies measured for a set of calibration point sources were analysed by the least-squares analysis. In the Monte Carlo method, the efficiencies for the calibration point sources were calculated by MCNP. The covariances of the efficiencies determined by the regression method were calculated analytically. Perturbation analysis was performed to estimate the covariances of the efficiencies calculated by the Monte Carlo method. Positive correlations higher than 0.8 were found in the uncertainties of the MCNP data for point-like sources. In the case of the regression method, the correlation coefficient contains both positive and negative terms. The efficiencies and their covariances for finite sample geometry were also estimated by both methods to take into account sample effects such as geometrical effect, and gamma-ray self-absorption, and found considerable differences in the cross-sections and their uncertainties for reaction products quantified with low energy gammas. The efficiencies and covariances were clearly affected by the properties of the sample.

Materials Genomics Search for Possible Helium-Absorbing Nano-Phases in Fusion Structural Materials.

Civilian fusion demands structural materials that can withstand the harsh environments imposed inside fusion plasma reactors. The structural materials often transmute under 14.1MeV fast neutrons, producing helium (He), which embrittles the grain boundary (GB) network. Here, it is shown that neutron-friendly and mechanically strong nano-phases with atomic-scale free volume can have low He-embedding energy and >10at.% He-absorbing capacity, and can be especially advantageous for soaking up He on top of resisting radiation damage and creep, provided they have thermodynamic compatibility with the matrix phase, satisfactory equilibrium wetting angle, as well as a high enough melting point. The preliminary experimental demonstration proves that is a good ab initio predictor of He shielding potency in nano-heterophase materials, and thus, is used as a key feature for computational screening. In this context, a list of viable compounds expected to be good He-absorbing nano-phases is presented, taking into account , the neutron absorption and activation cross-sections, the elastic moduli, melting temperature, the thermodynamic compatibility, and the equilbrium wetting angle of the nano-phases with the Fe matrix as an example.

Determination of the half-life of 184gRe and the isomeric ratio of 184m, gRe induced by D-T neutrons* *Supported by the National Natural Science Foundation of China (11975113)

The isomeric ratio of 184m, gRe and the half-life of 184gRe were measured in the 185Re (n, 2n) 184Re reaction at 14.8 MeV, and the uncertainty was discussed in detail. The measurements were performed using the activation method implemented for a rhenium sample using the K-400 neutron generator at the Chinese Academy of Engineering Physics (CAEP). Isomeric state and ground state nuclei of 184Re were identified by their γ-ray spectra. To eliminate the effect of the γ-ray emitted from the isomer on the counting of the ground state characteristic peaks, the isomeric ratio of 184m, gRe was calculated to be 0.29 ± 0.11 according to the neutron activation cross-section formula. This result is consistent with previous data within the uncertainty and can be used to determine parameters that characterize the dependence of the level density on the excitation energy and angular momentum. Through exponential function fitting and a detailed discussion of the uncertainty evaluation, the half-life of 184gRe was determined as 35.43 ± 0.16 d, which is consistent with the currently recommended value; however, the uncertainty assessment of the latter was barely documented. In addition, this study indicates that the half-life of the ground state can be obtained by eliminating the contamination of γ-rays emitted from the isomer, which provides the possibility of determining the half-lives of nuclides containing isomers.

Neutron activation cross section data library

To satisfy the requirements of nuclear reaction cross sections in nuclear engineering applications and nuclear physics studies, the Neutron Activation Cross Section Data Library has been established. 818 target nuclei including unstable target or isomeric target nuclei are considered in this library. The induced neutron energy range region is between 10−5 eV and 20 MeV. The standard ENDF-6 format is adopted, including general information, reaction cross sections, multiplicities, and so on. The recommended reaction cross sections were obtained using UNF code system and FDRR nuclear model codes or systematic analysis based on available experimental data. The full evaluated dataset containing the evaluated activation cross sections is openly available at http://doi.org/10.57760/sciencedb.j00113.00024.

Erratum: Measurement of (n, α) and (n, 2n) reaction cross sections at a neutron energy of 14.92 ± 0.02 MeV for potassium and copper with uncertainty propagation, [A. Gandhi, Aman Sharma, Rebecca Pachuau et al., Chin. Phys. C 46, 014002 (2022)]

Experimentally measured neutron activation cross sections are presented for the Cu(n, α) Cu, K(n, α) Cl, and Cu(n,2n) Cu reactions with detailed uncertainty propagation. The neutron cross sections were measured at an incident energy of 14.92 0.02 MeV, and the neutrons were based on the t(d, n)α fusion reaction. The Al(n, α) Na reaction was used as a reference reaction for the normalization of the neutron flux. The pre-calibrated lead-shielded HPGe detector was used to detect the residues' γ-ray spectra. The data from the measured cross sections are compared to the previously measured cross sections from the EXFOR database, theoretically calculated cross sections using the TALYS and EMPIRE codes, and evaluated nuclear data.

Measurement of (n,α) and (n,2n) reaction cross sections at a neutron energy of 14.92 ± 0.02 MeV for potassium and copper with uncertainty propagation * *For the flawless operation of the PURNIMA neutron generator during the experiment, we gladly acknowledged the assistance of scientific officers (SO's) of Purnima laboratory and Analytical Chemistry Division, BARC, Mumbai, India. For the financial assistance

Experimentally measured neutron activation cross sections are presented for the (n,α) , (n,α) , and (n,2n) reactions with detailed uncertainty propagation. The neutron cross sections were measured at an incident energy of 14.92 ± 0.02 MeV, and the neutrons were based on the t(d,n)α fusion reaction. The (n,α) reaction was used as a reference reaction for the normalization of the neutron flux. The pre-calibrated lead-shielded HPGe detector was used to detect the residues' γ-ray spectra. The data from the measured cross sections are compared to the previously measured cross sections from the EXFOR database, theoretically calculated cross sections using the TALYS and EMPIRE codes, and evaluated nuclear data.

A modified unscented transformation method for uncertainty propagation in neutron induced cross section measurement

This paper investigates the application of the modified unscented transformation (MUT) method for uncertainty propagation in neutron induced cross section measurement in Nuclear Science. The method is the improvement on unscented transformation (UT) technique by incorporating more parameters (not random) to increase the number of sigma points.We examined the application of MUT method for covariance analysis of cross section measurement in comparison to the first order sensitivity analysis method and Monte Carlo methods. The MUT method is simple to use and easier to implement in comparison to the sandwich formula. The Monte Carlo simulations had been carried out to illustrate the good performance of the MUT method in covariance analysis in neutron activation cross section measurement.

Bismuth activation with quasi-Maxwellian neutrons at kT∼30keV

Bismuth capture of neutrons is the termination point of the $s$-process cycle of nucleosynthesis in stellar environments. A new measurement is reported here for neutron activation of bismuth with an intense quasi-Maxwellian neutron source at $kT\ensuremath{\sim}30\phantom{\rule{0.16em}{0ex}}\mathrm{keV}$. The measurement was performed at the SARAF phase I accelerator facility by bombarding a 1.5-mA proton beam on the liquid-lithium larget. The cross section of the $^{209}\mathrm{Bi}(n,\ensuremath{\gamma})$ capture reaction leading to the $^{210}\mathrm{Bi}$ ground state was determined by combining $\ensuremath{\beta}$ measurements from the $^{210\mathrm{g}}\mathrm{Bi}$ decay and $\ensuremath{\alpha}$ and $\ensuremath{\gamma}$ from the subsequent $^{210}\mathrm{Po}$ decay, along with detailed Monte Carlo simulations of the $^{7}\mathrm{Li}(p,n)$ reaction kinematics and the activation experimental setup. Deduced Maxwellian averaged cross sections (MACS) for $^{209}\mathrm{Bi}(n,\ensuremath{\gamma})^{210\mathrm{g}}\mathrm{Bi}$ at $kT=30\phantom{\rule{0.16em}{0ex}}\mathrm{keV}$ using the ENDF, JEFF, and JENDL databases for the corrections and extrapolations yielded a value of $1.84\ifmmode\pm\else\textpm\fi{}0.09\phantom{\rule{0.16em}{0ex}}\mathrm{mb}$. A comparison is made with previous measurements, including time-of-flight (TOF) measurements of the total bismuth capture cross section. Plans for obtaining the MACS for capture to the bismuth-210 metastable state in the reaction $^{209}\mathrm{Bi}(n,\ensuremath{\gamma})^{210\mathrm{m}}\mathrm{Bi}$ are discussed, along with estimates based on our results in comparison with TOF measurements. The bismuth neutron activation cross section is also of importance for design of GenIV reactor coolant and subcritical accelerator driven systems, especially in light of the 3 million year half-life of the $^{210\mathrm{m}}\mathrm{Bi}$ isomer.

Improving Activation Cross Sections for Fusion Applications

High-quality nuclear data are required for reliable design calculations of fusion reactors as well as analyses related to safety, licensing, waste management, and decommissioning issues. The current work is focused on improving neutron activation cross-section data for nuclear reactions needed for fusion applications but also satisfying specific astrophysical needs. The cross-section data were evaluated on the basis of nuclear model calculation using the TALYS-1.8 code with varying nuclear model parameters. The evaluated data were compared to recent experimental data taken from EXFOR and other evaluated nuclear data libraries. Validation of the current evaluations has been performed using both differential and integral data sets. The updated cross-section data are going to be adopted into the new version of the TALYS Evaluated Nuclear Data Library.

ANITA-IEAF activation code package – updating of the decay and cross section data libraries and validation on the experimental data from the Karlsruhe Isochronous Cyclotron

ANITA-IEAF is an activation package (code and libraries) developed in the past in ENEA-Bologna in order to assess the activation of materials exposed to neutrons with energies greater than 20 MeV. An updated version of the ANITA-IEAF activation code package has been developed. It is suitable to be applied to the study of the irradiation effects on materials in facilities like the International Fusion Materials Irradiation Facility (IFMIF) and the DEMO Oriented Neutron Source (DONES), in which a considerable amount of neutrons with energies above 20 MeV is produced. The present paper summarizes the main characteristics of the updated version of ANITA-IEAF, able to use decay and cross section data based on more recent evaluated nuclear data libraries, i.e. the JEFF-3.1.1 Radioactive Decay Data Library and the EAF-2010 neutron activation cross section library. In this paper the validation effort related to the comparison between the code predictions and the activity measurements obtained from the Karlsruhe Isochronous Cyclotron is presented. In this integral experiment samples of two different steels, SS-316 and F82H, pure vanadium and a vanadium alloy, structural materials of interest in fusion technology, were activated in a neutron spectrum similar to the IFMIF neutron field.

Study of concrete activation with IFMIF-like neutron irradiation: Status of EAF and TENDL neutron activation cross-sections

The aim of this paper is to check the performance of last versions of EAF and TENDL libraries (EAF2007, EAF2010, and TENDL2014) in the prediction of concrete activation under the neutron irradiation environment expected in IFMIF, an accelerator-based neutron source conceived for fusion materials testing. For this purpose Activity and dose rate responses of three types of concrete (ITER-Bioshield kind, barite and magnetite concretes) have been studied. For these quantities, dominant nuclides and production pathways have been determined and, then, a qualitative analysis of the relevant activation cross-sections involved has been performed by comparing data from mentioned libraries with experimental data from EXFOR database. Concrete activation studies have been carried out with IFMIF-like neutron irradiation conditions using the ACAB code and EAF and TENDL libraries. The cooling times assessed are related to safety and maintenance operations, specifically 1 hour, 1 day and 12 days. Final conclusions are focused on the recommendations for the activation library to be used among those analyzed and cross-section data to be improved.

The cross section functions for neutron induced reactions with Rhenium in the energy range 13.0–19.5 MeV

The technique for measuring neutron activation cross-sections using wide energy neutron beams (NAXSUN) was recently developed at JRC-Geel . This method is based on the detection of the gamma activity induced by the activation of the samples in different but overlapping neutron fields and following an unfolding procedure. In the present work, measurements of the cross-section functions by the NAXSUN technique for the (n,a), (n,2n), (n,p) and (n,3n) reactions on rhenium isotopes 185 Re and 187 Re were performed. The results are the first experimental data for the mentioned reaction cross-sections in the energy range 13.0–19.5 MeV. The obtained data are of interest for possible applications of Re in nuclear technology and medicine.

Measurement of Neutron Activation Cross Sections on Mo isotopes in the Energy Range from 7 MeV to 15 MeV

An experimental study of the 92 Mo(n,p) 92 Nb m , 92 Mo(n,α) 89 Zr, 95 Mo(n,p) 95 Nb m , 95 Mo(n,p) 95 Nb, 96 Mo(n,p) 96 Nb, 97 Mo(n,p) 97 Nb, 98 Mo(n,p) 98 Nb m , 98 Mo(n,α) 95 Zr, 100 Mo(n,α) 97 Zr, and 92 Mo(n,2n) 99 Mo activation reaction cross sections were carried out in the 7-15 MeV energy range at the CV28 compact cyclotron at Physikalisch-Technische Bundesanstalt, Braunschweig. The PTB TOF spectrometer with a D(d,n) source is well suited for this difficult energy range were significant correction for non-monoenergetic neutrons have to be applied. Gamma-ray spectrometry was applied for the measurements of the activity of the reaction products.

Analysis of Threshold Neutron Excitation Function Measurements Using Wide Energy Neutron Beams

In this study we have improved the technique for measuring the neutron activation cross-section using wide energy neutron beams (NAXSUN). We propose a method for the determination of the default function for the unfolding procedure, which is an important and critical part for extracting reaction cross-sections from this type of measurements. The new method was tested on the measurement of the excitation function from the threshold energy up to 5.6 MeV for the 113In(n,n’)113mIn and 115In(n,n’)115mIn reactions.

Measurements of the neutron activation cross sections for Bi and Co at 386 MeV

Neutron activation cross sections for Bi and Co at 386 MeV were measured by activation method. A quasi-monoenergetic neutron beam was produced using the (7)Li(p,n) reaction. The energy spectrum of these neutrons has a high-energy peak (386 MeV) and a low-energy tail. Two neutron beams, 0° and 25° from the proton beam axis, were used for sample irradiation, enabling a correction for the contribution of the low-energy neutrons. The neutron-induced activation cross sections were estimated by subtracting the reaction rates of irradiated samples for 25° irradiation from those of 0° irradiation. The measured cross sections were compared with the findings of other studies, evaluated in relation to nuclear data files and the calculated data by Particle and Heavy Ion Transport code System code.

Determination of the default curve for the unfolding procedure in the measurement of threshold neutron excitation functions

In this study we have improved the technique for measuring the neutron activation cross-section using wide energy neutron beams (NAXSUN). We propose a method for the determination of the default function for the unfolding procedure, which is an important and critical part for extracting reaction cross-sections from this type of measurements. The new method was tested on the measurement of the excitation function from the threshold energy up to 5.6MeV for the 113In(n,n')113mIn and 115In(n,n')115mIn reactions.

Production of cationic 198Au3+ and nonionic 198Au0 for radionuclide therapy applications via the natAu(n,γ)198Au reaction

198Au (βmax = 0.96 MeV (98.6 %), γmax = 0.412 MeV (95.5 %) and T1/2 = 2.7 days) is a radionuclide with very appealing characteristics. 198Au has been widely used to treat the uterus, bladder, cervix, prostate, melanoma, breast, skin and other cancers. In the present study, cationic 198Au+3 and nonionic 198Au0 are prepared following thermal neutron irradiation of commercially available natural gold compounds in Tehran Research Reactor via the natAu(n,γ)198Au reaction. The prospects in the production of pure 198Au0 and 198Au+3 for radionuclide therapy are discussed and effect of reduction on the activity of radioactive gold is evaluated. Au0 particles were synthesized via NaBH4 reduction of aqueous solutions of hydrogen tetrachloroaurate trihydrate. Then two quartz tubes were charged with cationic 198Au3+ and nonionic 198Au0. After irradiation by thermal neutrons, the samples were analyzed for a period of 1 month by liquid scintillation counter and high purity germanium detector. As a result, natAu3+ reduction process had no significant effect on the activity of the 198Au sample. In conclusions, natural gold thermal neutron activation cross section is reasonably high for medical application.

RADIOLOGICAL CHARACTERISTICS OF DECOMMISSIONING WASTE FROM A CANDU REACTOR

The radiological characteristics for waste classification were assessed for neutron-activated decommissioning wastes from a CANDU reactor. The MCNP/ORIGEN2 code system was used for the source term analysis. The neutron flux and activation cross-section library for each structural component generated by MCNP simulation were used in the radionuclide buildup calculation in ORIGEN2. The specific activities of the relevant radionuclides in the activated metal waste were compared with the specified limits of the specific activities listed in the Korean standard and 10 CFR 61. The time-average full-core model of Wolsong Unit 1 was used as the neutron source for activation of in-core and ex-core structural components. The approximated levels of the neutron flux and cross-section, irradiated fuel composition, and a geometry simplification revealing good reliability in a previous study were used in the source term calculation as well. The results revealed the radioactivity, decay heat, hazard index, mass, and solid volume for the activated decommissioning waste to be 1.04 10 16 Bq, 2.09 10³ W, 5.31 10 14 ㎥-water, 4.69 10? ㎏, and 7.38 10¹㎥, respectively. According to both Korean and US standards, the activated waste of the pressure tubes, calandria tubes, reactivity devices, and reactivity device supporters was greater than Class C, which should be disposed of in a deep geological disposal repository, whereas the side structural components were classified as low- and intermediate-level waste, which can be disposed of in a land disposal repository. Finally, this study confirmed that, regardless of the cooling time of the waste, 15% of the decommissioning waste cannot be disposed of in a land disposal repository. It is expected that the source terms and waste classification evaluated through this study can be widely used to establish a decommissioning/disposal strategy and fuel cycle analysis for CANDU reactors.

Measurements of the neutron activation cross sections for Bi at 287 and 370 MeV

Abstract We present cross section measurements for neutron-induced activation of Bi, at 287 and 370 MeV. These values were derived from the activation method using a quasi-monoenergetic neutron field based on the 7Li(p,n) reaction. In separate experiments, samples were irradiated with neutrons derived from 7Li(p,n) reaction at either 0º or 30º for proton beam axis. This approach allows the subtraction of the low energy neutron components. The measured cross sections are compared with the findings of other studies, and evaluated in relation to nuclear data files.

The Measurement of Neutron Activation Cross Section of 59Co Below 36 MeV

The Measurement of Neutron Activation Cross Section of 59Co Below 36 MeV