Cross section determination of^(27)Al(n,2n)^(26)Al reaction induced by 14-MeV neutrons uniting with D-T neutron activation and AMS techniques

Aluminum is the primary structural material in nuclear engineering,and its cross section induced by 14-MeV neutrons is of great significance.To address the issue of insufficient accuracy for the^(27)Al(n,2n)^(26)Al reaction cross section,the activation method and accelerator mass spectrometry(AMS)technique were used to determine the^(27)Al(n,2n)^(26)Al cross section,which could be used as a D-T plasma ion temperature monitor in fusion reactors.At the China Academy of Engineering Physics,neutron activation was performed using a K-400 neutron generator produced by the T(d,n)4He reaction.The^(26)Al∕^(27)Al isotope ratios were measured using the newly installed GYIG 1 MV AMS at the Institute of Geochemistry,Chinese Academy of Sciences.The neutron flux was monitored by measuring the activity of 92mNb produced by the 93Nb(n,2n)92mNb reaction.The measured results were compared with available data in the experimental nuclear reaction database,and the measured values showed a reasonable degree of consistency with partially available literature data.The newly acquired cross-sectional data at 12 neutron energy points through systematic measurements clarified the divergence,which has two different growth trends from the existing experimental values.The obtained results are also compared with the corresponding evaluated database,and the newly calculated excitation functions with TALYS−1.95 and EMPIRE−3.2 codes,the agreement with CENDL−3.2,TENDL-2021 and EMPIRE−3.2 results are generally acceptable.A substantial improvement in the knowledge of the^(27)Al(n,2n)^(26)Al reaction excitation function was obtained in the present work,which will lay the foundation for the diagnosis of the fusion ion temperature,testing of the nuclear physics model,evaluation of nuclear data,etc.

Measurement of the neutron-induced total cross sections of ^(nat)Pb from 0.3 eV to 20 MeV on the Back-n at CSNS

The neutron-induced total cross sections of natural lead have been measured in a wide energy range(0.3 eV-20 MeV)on the back-streaming white neutron beamline(Back-n)at the China Spallation Neutron Source.Neutron energy was determined by the neutron total cross-section spectrometer using the time-of-flight technique.A fast multi-cell fission chamber was used as the neutron detector,and a 10-mm-thick high-purity natural lead sample was employed for the neutron transmission measurements.The on-beam background was determined using Co,In,Ag,and Cd filters.The excitation function of ^(nat)Pb(n,tot)reaction below 20 MeV was calculated using the TALYS-1.96 nuclear-reaction modeling program.The present results were compared with previous results,the evaluated data available in the five major evaluated nuclear data libraries(i.e.,ENDF/B-VIII.0,JEFF-3.3,JENDL-5,CENDL-3.2,and BROND-3.1),and the theoretical calculation curve.Good agreement was found between the new results and those of previous experiments and with the theoretical curves in the corresponding region.This measurement obtained the neutron total cross section of natural lead with good accuracy over a wide energy range and added experimental data in the resonance energy range.This provides more reliable experimental data for nuclear engineering design and nuclear data evaluation of lead.

Measurement of the neutron total cross section of carbon at the Back-n white neutron beam of CSNS

To verify the performance of the neutron total cross-sectional spectrometer, the neutron total cross section of carbon is initially measured in the energy range of 1 eV to 20 MeV using the time-of-flight method. The measurement is performed at the Back-n white neutron source with a 76-m time-of-flight path using the China Spallation Neutron Source. A multilayer fast fission chamber with 235U and 238U is employed as the neutron detector. The diameter and thickness of the natural graphite sample are 70 mm and 40 mm, respectively. Signal waveforms are collected using a data acquisition system. Off-line data processing was used to obtain the neutron time-of-flight spectra and transmissions. The uncertainty of the counting statistics is generally approximately 3% for each bin in the energy range of 1–20 MeV. It is determined that the results for the neutron total cross section of carbon obtained using ^235U cells are in good agreement with the results obtained using 238U cells within limits of statistical uncertainty. Moreover, the measured total cross sections show good agreement with the broadening evaluated data.

Preparation and verification of mixed high-energy neutron crosssection library for ADS

An accelerator-driven subcritical system(ADS)is driven by an external spallation neutron source, which is generated from a heavy metal spallation target to maintain stable operation of the subcritical core, where the energy of the spallation neutrons can reach several hundred megaelectron volts. However, the upper neutron energy limit of nuclear cross-section databases, which are widely used in critical reactor physics calculations, is generally 20 MeV.This is not suitable for simulating the transport of highenergy spallation neutrons in the ADS. We combine the Japanese JENDL-4.0/HE high-energy evaluation database and the ADS-HE and ADS 2.0 libraries from the International Atomic Energy Agency and process all the data files for nuclides with energies greater than 20 MeV. We use the continuous pointwise cross-section program NJOY2016 to generate the ACE-formatted cross-section data library IMPC-ADS at multiple temperature points. Using the IMPC-ADS library, we calculate 10 critical benchmarks of the International Criticality Safety Benchmark Evaluation Project manual, the 14-MeV fixed-source problem of the Godiva sphere, and the neutron flux of the ADS subcritical core by MCNPX. To verify the correctness of the IMPCADS, the results were compared with those calculated using the ENDF/B-VII.0 library. The results showed thatthe IMPC-ADS is reliable in effective multiplication factor and neutron flux calculations, and it can be applied to physical analysis of the ADS subcritical reactor core.

Measurement of ^(232)Th(n,2n)^(231)Th reaction cross-sections at neutron energies of 14.1 MeV and 14.8 MeV using neutron activation method

In this study, the activation cross-sections were measured for ^(232)Th(n,2n)^(231)Th reactions at neutron energies of 14.1 and 14.8 MeV, which were produced by a neutron generator through a T(d,n)~4He reaction. Induced gamma-ray activities were measured using a low background gamma ray spectrometer equipped with a high resolution HPGe detector. In the cross-section calculations, corrections were made regarding the effects of gamma-ray attenuation, dead-time, fluctuation of the neutron flux, and low energy neutrons. The measured cross-sections were compared with the literature data, evaluation data(ENDF-B/VII.1, JENDL-4.0 and CENDL-3.1), and the results of the model calculation(TALYS1.6).

Measurements of fast-neutron capture cross sections for ^(159)Tb and ^(169)Tm

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Measurements of the ^(107)Ag neutron capture cross sections with pulse height weighting technique at the CSNS Back-n facility

Silver indium cadmium(Ag–In–Cd) control rod is widely used in pressurized water reactor nuclear power plants,and it is continuously consumed in a high neutron flux environment. The mass ratio of ^(107)Ag in the Ag–In–Cd control rod is 41.44%. To accurately calculate the consumption value of the control rod, a reliable neutron reaction cross section of the ^(107)Ag is required. Meanwhile,^(107)Ag is also an important weak r nucleus. Thus, the cross sections for neutron induced interactions with ^(107)Ag are very important both in nuclear energy and nuclear astrophysics. The(n, γ) cross section of ^(107)Ag has been measured in the energy range of 1–60 eV using a back streaming white neutron beam line at China spallation neutron source. The resonance parameters are extracted by an R-matrix code. All the cross section of ^(107)Ag and resonance parameters are given in this paper as datasets. The datasets are openly available at http://www.doi.org/10.11922/sciencedb.j00113.00010.

Measurement of fast-neutron capture cross sections for ^(75)As

The cross sections of the75 As(n,γ)76As reaction were measured in the neutron energy range from 0.50 to 1.50 MeV by using the activation technique. Neutrons were produced via the T(p,n)3He reaction and the cross sections of the 197Au(n,γ)198Au reaction were used to determine the absolute neutron flux. Present results are compared with existing measurements and evaluations.

Cross-section measurement of(n,2n) reactions for Nd isotopes induced by 14 MeV neutrons

Cross-sections of the(n,2n) reactions for neodymium(Nd) isotopes induced by 14 MeV neutrons were measured in this work by using the activation and relative methods. The measured cross-sections of the ^(150)Nd(n,2n)^(149)Nd,^(148)Nd(n,2n)^(147)Nd, and ^(142)Nd(n,2 n)^(141)Nd reactions were 1854 ± 81, 1789 ± 119, and 1559 ± 98 mb, respectively, at a neutron energy of 14.2 ± 0.2 MeV,and 1485 ± 74, 1726 ± 85, and 1670 ± 119 mb, respectively, at 14.9 ± 0.2 MeV. The results were compared with the experimental values from the reported literature, with the evaluated data from the ENDF/B-VII.1, CENDL-3.1, and JENDL-4.0 libraries, and with the curves calculated by the Talys-1.8 code.

Theoretical Analysis of Neutron Double-Differential Cross Sections of n ＋^9Be Reactions

By using the nuclear reaction model for light nuclei, the calculations of the double-differential cross sections of outgoing neutrons from n ＋^9Be reactions are performed. The total outgoing neutrons are only come from the （n, 2n）2a reaction channel. The （n, 2n）2a reaction channel is achieved through six different reaction approach, which are illustrated in this paper. The calculated results agree very well with the measured data at En = 7.1, 8.09, 8.17, 9.09, 9.97 and 10.26 MeV, because the updated level schemes related to the n ＋ ^9Be reactions have been employed in this calculations.

Calculation of production cross sections of γ-rays from thermal-neutron captures

The calculation methods of production cross sections of γ-rays for thermal-neutron captures are briefly presented. The check of intensity balance is made. The examples are given to illustrate its application.

The Role of Replacing CdO by Fe₂O₃on the Fast Neutron Removal Cross Sections in Cd-Boro Phosphate Glass Shield

This work deals with the application of [MERCSF-N] computer program in calculating the macroscopic effective removal cross-section of fast neutrons, ΣR (cm-1), for two different boro phosphate glass systems: (0.5 - x) CdO-x Fe2O3-0.4 P2O5-0.1 B2O3 and (0.5 - x) B2O3-x Fe2O3-0.1 CdO-0.4 P2O5 (with 0.05 ≤ x ≤ 0.5 by mole), to realize from the role of iron in the attenuation process and hence the usefulness of the glass containing iron as neutrons shielding material. The effect of replacing cadmium and boron oxides by iron oxide has been analyzed which proved that iron is more efficient than cadmium in attenuating and removing fast neutrons and that the presence of small amounts of B2O3 at least 0.1 mole fraction, with iron is needed to aid improving the removal cross-section of iron phosphate glasses. Experimental IR results are developed and used to trace the structural change and confirm the role of iron in the removal cross section.

The Systematics Study of (n, p) Reaction Cross-Sections at 14.7 MeV Neutron Energy

Based on the statistical model and taking into account the Q-value dependence and odd-even effects, we proposed a new empirical formula to reproduce the cross sections of the (n, p) reactions at 14.7 MeV neutron energy and at the target mass number 14 ≤ A ≤ 198 for even A and 29 ≤ A ≤ 205 for odd A. All calculated results from the proposed empirical formula were compared to the experimental data as well as the available semi-empirical formula obtained by other authors. A high level of agreement has been found between the collected experimental data and the most of semiempirical formulae obtained by others.

Measurement of the ^(232)Th(n,f)cross section in the 1-200 MeV range at the CSNS Back-n

The ^(232)Th(n,f)cross section is very important in basic nuclear physics and applications based on the Th/U fuel cycle.Using the time-of-flight method and a multi-cell fast-fission ionization chamber,a novel measurement of the^(232)Th(n,f)cross sec-tion relative to^(235)U in the 1–200 MeV range was performed at the China Spallation Neutron Source Back-n white neutron source(Back-n).The fission event-neutron energy spectra of^(232)Th and^(235)U fission cells were measured in the single-bunch mode.Corrected 232Th/235U fission cross-sectional ratios were obtained,and the measurement uncertainties were 2.5–3.7%for energies in the 2–20 MeV range and 3.6–6.2%for energies in the 20–200 MeV range.The^(232)Th(n,f)cross section was obtained by introducing the standard cross section of^(235)U(n,f).The results were compared with those of previous theoreti-cal calculations,measurements,and evaluations.The measured 232Th fission cross section agreed with the main evaluation results in terms of the experimental uncertainty,and 232Th fission resonances were observed in the 1–3 MeV range.The present results provide^(232)Th(n,f)cross-sectional data for the evaluation and design of Th/U cycle nuclear systems.

Conceptual design of a Cs2LiLaBr6 scintillator‑based neutron total cross section spectrometer on the back‑n beam line at CSNS

To reduce the experimental uncertainty in the 235 U resonance energy region and improve the detection efficiency for neutron total cross section measurements compared with those obtained with the neutron total cross section spectrometer(NTOX), a dedicated lithium-containing scintillation detector has been developed on the Back-n beam line at the China Spallation Neutron Source. The Fast Scintillator-based Neutron Total Cross Section(FAST) spectrometer has been designed based on a Cs2Li La Br6(CLLB) scintillator considering the γ-ray flash and neutron environment on the Back-n beam line. The response of the CLLB scintillator to neutrons and γ-rays was evaluated with different 6Li/7 Li abundance ratios using Geant4. The neutron-γdiscrimination performance of the CLLB has been simulated considering different scintillation parameters, physical designs,and light readout modes. A cubic 6Li-enriched( > 90%) CLLB scintillator, which has a thickness of 4-9 mm and side length of no less than 50 mm to cover the Φ 50 mm neutron beam at the spectrometer position, has been proposed coupling to a side readout SiPM array to construct the FAST spectrometer. The developed simulation techniques for neutron-γ discrimination performance could provide technical support for other neutron-induced reaction measurements on the Back-n beam line.

Measurement of Br(n,γ)cross sections up to stellar s-process temperatures at the CSNS Back‑n

The neutron capture cross sections(n,γ)of bromine were obtained using the time-of-flight technique at the Back-n facility of the China Spallation Neutron Source.Promptγ-rays originating from neutron-induced capture events were detected using four C_(6)D_(6) detectors.The pulse-height weighting technique and double-bunch unfolding method based on Bayesian theory were used in the data analysis.Background deductions,normalization,and corrections were carefully considered to obtain reliable measurement results.The multilevel R-matrix Bayesian code SAMMY was used to extract the resonance parameters in the resolved resonance region(RRR).The average cross sections in the unresolved resonance region(URR)were obtained from 10 to 400 keV.The experimental results were compared with data from several evaluated libraries and previous experi-ments in the RRR and URR.The TALYS code was used to describe the average cross sections in the URR.The astrophysical Maxwell average cross sections(MACSs)of ^(79,81)Br from kT=5 to 100 keV were calculated over a sufficiently wide range of neutron energies.At a thermal energy of kT=30 keV,the MACS value for ^(79)Br 682±68 mb was in good agreement with the KADoNiS v1.0 recommended value.By contrast,the value of 293±29 mb for ^(81)Br was substantially higher than that of the evaluated database and the KADoNiS v1.0 recommended value.

Measurement of Neutron Capture Cross Section for ^(180)Hf

The neutron capture cross section for 180Hf relative to the 197Au（n,γ）198Au reaction ismeasured at neutron energies of 0.52,1.10 and 1.60 MeV using the activation method.Theactivities of the products are measured with a high resolution HPGe detector gamma ray spec-trometer.The errors of the measured results are 5%～6%.Our results have a good agreementwith those of other authors within errors.Cross sections for the reaction are calculated withUNF code and recommended data are also given.

Activation cross sections for reactions induced by 14 MeV neutrons on natural copper

The cross sections for the^(63)Cu(n,a)^(60(m+g))Co,^(65)Cu(n,2 n)^(64)Cu, and ^(65)Cu(n,p)^(65)Ni reactions have been studied in the neutron energy range of 13.5–14.8 MeV using the activation technique. The neutron beams were produced via the^3 H(d,n)~4 He reaction. The neutron energies of different directions in the measurements were determined beforehand by the method of cross section ratios for the^(90)Zr(n,2 n)^(89 m+g)Zr and ^(93)Nb(n,2 n)^(92 m)Nb reactions. The results in the present work were discussed and compared with measurement results found in the literatures.

Water equivalence of some 3D dosimeters:a theoretical study based on the effective atomic number and effective fast neutron removal cross section

Effective atomic numbers for photon energy absorption(ZPEA_(eff)) and their corresponding electron numbers (NPEA_(eff)), and effective macroscopic removal cross sections of fast neutrons(RR) were calculated for 27 different types of three-dimensional dosimeters, four types of phantom materials, and water. The values of ZPEA_(eff) and NPEA_(eff) were obtained using the direct method for energies ranging from 10 keV to 20 MeV. Results are presented relative to water, for direct comparison over the range of examined energies. The effect of monomers that are used in polymer gel dosimeters on the water equivalence is discussed. The relation between Σ_R and hydrogen content was studied. Micelle gel dosimeters are highly promising because our results demonstrate perfect matching between the effective atomic number, electron density number, and fast neutron attenuation coefficient of water.

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.