Neutron Energy Region Research Articles

Neutron radiative capture cross section for sodium with covariance analysis

The neutron radiative capture cross sections measurement has been carried out for the $$^{23}$$ Na nucleus in the neutron energy region from 0.6 to 3.2 MeV using the neutron activation technique followed by off-line $$\gamma $$ -ray spectrometry. The measurement was made relative to the $$^{115}$$ In(n,n $$\prime $$ $$\gamma $$ ) $$^{115}\hbox {In}^{m}$$ reference monitor reaction cross section. The neutrons were produced via the $$^{7}$$ Li(p,n) $$^{7}$$ Be reaction. Detailed uncertainty propagation has been performed using the covariance analysis, and the measured cross sections are being reported with their uncertainties, covariance, and correlation matrix. The necessary corrections have been made for the low background neutron energy contribution, $$\gamma $$ -ray true coincidence summing, and self-attenuation process. The obtained neutron spectrum averaged cross sections of $$^{23}$$ Na(n, $$\gamma $$ ) $$^{24}$$ Na are discussed and compared with the existing cross sections data retrieved from the EXFOR database. EMPIRE-3.2 and TALYS-1.9 calculations were performed in order to determine the radiative capture cross section in this energy region. The present results are also compared with the evaluated nuclear data from ENDF/B-VIII.0, TENDL-2019, IRDFF-1.05, JENDL-4.0, and JEFF-3.3. The obtained cross section results are in good agreement with existing experimental data, evaluated libraries, and reaction models for the highest energy points (2.11 and 3.13 MeV), while the lowest-energy point at 0.61 MeV underestimates them.

Investigations on neutron attenuation properties of poly-boron materials using Am-Be and 252Cf sources neutron spectra

The neutron attenuation characteristics of polyboron materials was studied using Am-Be (α,n) and 252Cf (Spontaneous Fission) sources. Neutron Attenuation Factor (NAF) is obtained for poly boron materials with addition of different concentrations of boron by neutron transmission experiments. Neutron transport calculations are carried by simulating the transmission of neutron from Am-Be and 252Cf sources through polyboron materials using Monte Carlo based simulations. The experimental results are fairly in good agreement with transport calculations. The results showed that Neutron Attenuation Factor (NAF) mainly depend on the incident neutron energy and boron number density present in HDPE shielding material. The effect of boron in HDPE on neutron shielding is only marginal (2%) because of low value of neutron cross section at fast energy region. However, in the low neutron energy region, it varies with wt % of boron and 20% boron as additive in HPDE is an optimum value for neutron attenuation. The details of experimental observations and theoretical investigations are discussed in this work.

Development of real-time neutron detectors with different sensitivities to thermal, epithermal, and fast neutrons in BNCT

Boron neutron capture therapy (BNCT) is a type of radiation therapy which can selectively kill tumor cells. The irradiation field employed in BNCT is a high intensity with a wide energy range, and undesired neutrons can be irradiated in a patient's body during the treatment as a result. These neutrons are categorized into thermal, epithermal, and fast neutrons on the basis of their energy ranges. Their individual doses inside the body should be measured in real-time because their relative biological effectiveness differs in each neutron energy region. Therefore, in this study, we developed a system comprising real-time neutron detectors with separate sensitivities for each neutron type with an Eu:LiCaAlF6 scintillator combined with a polyethylene moderator. We tested the performance of the detectors and performed dose evaluation using a water phantom representing the human body and compared the results with those obtained via a commonly used activation method. Thermal, epithermal, and fast neutron doses were successfully measured in real-time at the surface of the entire body, and we were able to estimate the fast neutron dose inside the body.

Continuous-energy Monte Carlo calculation analysis of light-water moderation critical experiments with aqueous solution samples of B, Rh, Cs, Nd, Sm, Eu, Gd, and Er

ABSTRACT In order to validate the neutron absorption cross sections in the JENDL-4.0 nuclear data library, critical analysis was performed using a continuous-energy Monte Carlo calculation code for the light-water moderation UO2 cores loaded with the aqueous solution samples of eight elements (B, Rh, Cs, Nd, Sm, Eu, Gd, and Er). The concentrations of the elements in the solution samples were changed in the experiments, and the criticality water levels were measured. The calculated keff values of the critical cores for the different concentrations of each element were plotted against the reactivity worth of the solution samples calculated based on the differences in the critical water levels. They were fitted by an approximate linear curve to obtain the slope which was regarded as a negative value of a fractional bias of the absorption cross section in the thermal and epi-thermal neutron energy regions. The obtained fractional biases indicated that the absorption cross sections of Sm, Eu, and Gd agreed with the actual values with the uncertainties of about 3%, and those of B, Rh, and Er overestimated them by 6 3%, 5 3%, and 6 4%, respectively.

Cross-section measurements for 58,60,61Ni(n, α)55,57,58Fe reactions in the 4.50 – 5.50 MeV neutron energy region * * Supported by the National Natural Science Foundation of China (11775006) and China Nuclear Data Center and the Science and Technology on Nuclear Data Laboratory

The cross sections at 5 energy points of the 58Ni(n, α)55Fe reaction were measured in the 4.50 MeV ≤ En ≤ 5.50 MeV region while those for the 60Ni(n, α)57Fe and 61Ni(n, α)58Fe reactions were measured at En = 5.00 and 5.50 MeV using the 4.5 MV Van de Graaff accelerator at Peking University. A gridded twin ionization chamber (GIC) was used as the detector, and enriched 58Ni, 60Ni, and 61Ni foil samples were prepared and mounted at the sample changer of the GIC. Three highly enriched 238U3O8 samples inside the GIC were used to determine the relative and absolute neutron fluxes. The neutron energy spectra were obtained through unfolding the pulse height spectra measured by the EJ-309 liquid scintillator. The interference from the low-energy neutrons and impurities in the samples has been corrected. The present data of the 60Ni(n, α)57Fe reaction are the first measurement results below 6.0 MeV, and those of the 61Ni(n, α)58Fe reactions are the first measurement results in the MeV region. The present results have been compared with existing measurements, evaluations, and TALYS-1.9 calculations.

Application of a silicon detector array in [formula omitted] reaction cross-section measurements at the CSNS Back-n white neutron source

The Back-n white neutron source at the China Spallation Neutron Source (CSNS) was put into operation in 2018, providing a good platform for the cross-section measurement of neutron induced light charged particle (lcp) emission (n,lcp) reactions. Currently, in order to detect and identify light charged particles produced in (n,lcp) reactions at Back-n, a silicon detector array has been developed as a part of Light-charged Particle Detector Array. Composed of 15 silicon detectors, the array can detect the emitted particles ranging from 19.2°to 160.8°. The energy resolution of each silicon detector unit is better than 1% for 5.499 MeV α particles. Adopting this silicon detector array, several experimental measurements have been completed. Differential cross sections and angle-integrated cross sections in the neutron energy region ranging from the eV to the MeV range are extracted. Compared with previous measurements and evaluations, new results are obtained in the present measurements. These experiments indicate the feasibility of the silicon detector array in (n,lcp) reaction cross-section measurements.

Double-bunch unfolding methods for the Back-n white neutron source at CSNS

The Back-n white neutron source at China Spallation Neutron Source (CSNS) is designed specifically for nuclear data measurements and multidisciplinary neutron applications. The time of flight (TOF) method is deployed to obtain a high resolution of neutron energy. In the normal operation mode of CSNS there are two proton bunches with a time interval of 410 ns in each pulse which has a repetition frequency of 25 Hz. Due to the superposition of the event distributions corresponding to two bunches, the resolution of the TOF measurement at Back-n will be degraded by the double-bunch characteristics if the measured event distribution is used directly without unfolding, especially in the higher neutron energy region. To nearly recover the event distribution corresponding to a single proton bunch, the unfolding methods have been developed to obtain better time resolution and energy resolution. Two kinds of unfolding algorithms based on the iterative Bayesian method and the iterative event re-distribution method, respectively, have been developed and tested with simulated data and experimental data. The developed methods have shown nice performance in unfolding the double-bunch event distributions with high accuracy and fast convergence. With the deploying of the unfolding methods, the data of the first batch experiments conducted at Back-n in the normal double-bunch mode were analyzed and nearly the same time resolution as in the case of the single-bunch mode has been obtained.

Measurement of the U-238/U-235 fission cross section ratio at CSNS – Back-n WNS

Based on the China Spallation Neutron Source (CSNS) – Back-streaming white neutron source (Back-n WNS), the measurement of the U-238/U-235 fission cross section ratio in 1–20 MeV neutron energy region has been carried out by a multi-cell fast fission ionization chamber using the time-of-flight method. The measured resonance peaks of the 235U(n, f) reaction are in good agreement with those of the ENDF/B-8.0 database. The measured six groups of the U-238/U-235 fission cross section ratios in 1–20 MeV neutron energy region agree with those of the ENDF/B-8.0 database in trend, and the average discrepancies between experiment and database are from 1.9% to 2.6%. The relative experimental uncertainties are 2.3%–3.6% (1.4–20 MeV), and the measured data agree with those of database within experimental uncertainties at most energy points. The measured U-238/U-235 fission cross section ratio could provide information and data support in relevant nuclear data evaluation.

Measurements of neutron-induced light-charged particle emission reactions

In the past two decades cooperating with Frank Laboratory of Neutron Physics (FLNP), Joint Institute for Nuclear Research (JINR) measurements of (n, α) reaction cross sections for 6Li, 10B, 25Mg, 39K, 40Ca, 54,56,57 Fe, 58Ni, 63Cu, 64,67 Zn, 95Mo, 143Nd and 147,149 Sm nuclei were performed in the MeV neutron energy region based on the 4.5 MV Van de Graaff accelerator at Peking University. In recent years, our measurements were extended in three aspects. Firstly, measurements were expanded from two-body reactions to three-body reactions such as 10B (n, t2 α). Secondly, the neutron energy region was extended from below 8 MeV to 8 - 11 MeV by using the HI-13 tandem accelerator of China Institute of Atomic Energy (CIAE), with which cross sections of 54,56 Fe(n, α)53,51Cr reactions were measured. Thirdly, based on the newly-built China Spallation Neutron Source (CSNS) Back-n WNS (White Neutron Source), differential and angle-integrated cross sections for 6Li(n, t) and 10B(n, α) reactions were measured in the neutron energy region from 1 eV to 3 MeV.

Study of the (n,2n) reaction cross section on 174,176Hf isotopes

Cross section measurements of the 174Hf(n,2n)173Hf and 176Hf(n,2n)175Hf reactions have been performed at the VdG Tandem accelerator of NCSR “Demokritos” in Athens, in the neutron energy region from 8.8 to 11.5 MeV, using the activation technique. Statistical model calculations based on the Hauser - Feshbach theory have also been implemented by using the code EMPIRE-II with different sets of input parameters. The predictions were compared to the data of the present work as well as data from literature.

IRDFF-II: A New Neutron Metrology Library

High quality nuclear data is the most fundamental underpinning for all neutron metrology applications. This paper describes the release of version II of the International Reactor Dosimetry and Fusion File (IRDFF-II) that contains a consistent set of nuclear data for fission and fusion neutron metrology applications up to 60 MeV neutron energy. The library is intended to support: a) applications in research reactors; b) safety and regulatory applications in the nuclear power generation in commercial fission reactors; and c) material damage studies in support of the research and development of advanced fusion concepts. The paper describes the contents of the library, documents the thorough verification process used in its preparation, and provides an extensive set of validation data gathered from a wide range of neutron benchmark fields. The new IRDFF-II library includes 119 metrology reactions, four cover material reactions to support self-shielding corrections, five metrology metrics used by the dosimetry community, and cumulative fission products yields for seven fission products in three different neutron energy regions. In support of characterizing the measurement of the residual nuclei from the dosimetry reactions and the fission product decay modes, the present document lists the recommended decay data, particle emission energies and probabilities for 68 activation products. It also includes neutron spectral characterization data for 29 neutron benchmark fields for the validation of the library contents. Additional six reference fields were assessed (four from plutonium critical assemblies, two measured fields for thermal-neutron induced fission on 233U and 239Pu targets) but not used for validation due to systematic discrepancies in C/E reaction rate values or lack of reaction-rate experimental data. Another ten analytical functions are included that can be useful for calculating average cross sections, average energy, thermal spectrum average cross sections and resonance integrals. The IRDFF-II library and comprehensive documentation is available online at www-nds.iaea.org/IRDFF/. Evaluated cross sections can be compared with experimental data and other evaluations at www-nds.iaea.org/exfor/endf.htm. The new library is expected to become the international reference in neutron metrology for multiple applications.

Neutron capture and total cross-section measurements of 155Gd and 157Gd at ANNRI in J-PARC

Neutron capture and total cross sections of 155Gd and 157Gd were measured with the neutron time-of-flight method in Accurate Neutron-Nucleus Reaction measurement Instrument at Material and Life Science Experimental Facility of the Japan Proton Accelerator Research Complex using the NaI (Tl) spectrometer and Li-glass detectors. Preliminary cross sections were obtained in the neutron energy region from 4 meV to 100 meV. The derived cross sections agree with evaluated values in JENDL 4.0 and the experimental results by Mastromarco but were not consistent with those by Leinweber.

Fast Neutron Capture Reaction Data Measurement of Minor Actinides for Development of Nuclear Transmutation Systems

A research project entitled "study on accuracy improvement of fast-neutron capture reaction data of long-lived MAs for development of nuclear transmutation systems" is ongoing as a joint collaboration, including Tokyo Tech, Japan Atomic Energy Agency and Kyoto University in Japan. This project focuses on neutron capture reaction of minor actinides237Np,241Am and243Am, in the fast neutron energy region. The final goal of this project is to improve the neutron capture cross sections utilizing a high-intensity neutron beam from a spallation neutron source of the Japan Proton Research Complex. The overview of the project and the current progress are presented.

In-beam γ-rays of back-streaming white neutron source at China Spallation Neutron Source

The back-streaming neutron beam line (Back-n) was built in the beginning of 2018, which is part of the China Spallation Neutron Source (CSNS). The Back-n is the first white neutron beam line in China, and its main application is for nuclear data measurement. For most of neutron-induced nuclear reaction measurements based on white neutron facilities, the beam of gamma rays accompanied with neutron beam is one of the most important experimental backgrounds. The back streaming neutron beam is transported directly from the spallation target to the experimental station without any moderator or shielding, the flux of the in-beam gamma rays in the experimental station is much larger than those of these facilities with neutron moderator and shielding. Therefore, it is necessary to consider the influence of in-beam gamma rays on the experimental results. Studies of the in-beam gamma rays are carried out at the back-n. Monte-Carlo simulation is employed to obtain the energy distribution and the time structure of the in-beam gamma rays. According to the simulation results, when the neutron flight time is longer than 1.0 μs the energy distribution of the in-beam gamma rays does not vary with flight time. Therefore, the time structure of these gamma rays can be measured without the correction of the detection efficiency. In this work, the time structure of the in-beam gamma rays in the low neutron energy region is measured by both direct and indirect methods. In the direct measurement, a <sup>6</sup>Li loaded ZnS(Ag) scintillator is located on the neutron beam line and the time of flight method is used to determine the time structure of neutrons and gamma rays. The gamma rays are separated from neutrons with pulse-shape discrimination. The black filter method is used to verify the particle discrimination results. In the indirect measurement, the C<sub>6</sub>D<sub>6</sub> scintillation detectors are used to measure the gamma rays scattered off a Pb sample on the way of the neutron beam. The time structure of the in-beam gamma rays is derived from that of the scattered gamma rays. The experimental results are in good agreement with the simulations with the time-of-flight between 12 μs and 2.0 ms. Besides, according to the simulation results, the intensity of the in-beam gamma rays is 1.21 × 10<sup>6</sup> s<sup>–1</sup>·cm<sup>–2</sup> in the center of the experimental station 2 of Back-n, which is 76.5 m away from the spallation target of CSNS.

Influence of parasitic neutrons to the 176Hf(n,2n)175Hf reaction cross section

Cross sections for 176Hf(n,2n)175Hf reaction have been measured at the VdG Tandem accelerator of NCSR \Demokritos", in the neutron energy region from 8.80 to 11.02 MeV, using the activation technique. In order to account for the contamination of the 176Hf(n,2n)175Hf by the 174Hf(n,γ)175Hf reaction activated by the presence of parasitic neutrons, an experimental method has been developed based on the influence of the parasitic neutrons in the case of the (n,2n) and (n,γ) reactions on 197Au. The results were found to be consistent with the energy spectrum of the neutron beam which has been studied by means of the multiple foil activation analysis technique as well as by a liquid scintillator BC501A detector and subsequent deconvolution of its recoil energy spectra.

Measurements of differential and angle-integrated cross sections for the 10B(n, α)7Li reaction in the neutron energy range from 1.0 eV to 2.5 MeV * * Supported by financially the National Key R&D Program of China (2016YFA0401604) and the National Natural Science Foundation of China (11775006)

Differential and angle-integrated cross sections for the 10B(n, α)7Li, 10B(n, α0) 7Li and 10B(n, α1) 7Li* reactions have been measured at CSNS Back-n white neutron source. Two enriched (90%) 10B samples 5.0 cm in diameter and ~85.0 μg/cm2 in thickness each with an aluminum backing were prepared, and back-to-back mounted at the sample holder. The charged particles were detected using the silicon-detector array of the Light-charged Particle Detector Array (LPDA) system. The neutron energy En was determined by TOF (time-of-flight) method, and the valid α events were extracted from the En-Amplitude two-dimensional spectrum. With 15 silicon detectors, the differential cross sections of α-particles were measured from 19.2° to 160.8°. Fitted with the Legendre polynomial series, the (n, α) cross sections were obtained through integration. The absolute cross sections were normalized using the standard cross sections of the 10B(n, α)7Li reaction in the 0.3 – 0.5 MeV neutron energy region. The measurement neutron energy range for the 10B(n, α)7Li reaction is 1.0 eV≤En < 2.5 MeV (67 energy points), and that for the 10B(n, α0) 7Li and 10B(n, α1) 7Li* reactions is 1.0 eV ≤ En < 1.0 MeV (59 energy points). The present results have been analyzed by the resonance reaction mechanism and the level structure of the 11B compound system, and compared with existing measurements and evaluations.

Neutron capture cross-section measurement and resolved resonance analysis of 237Np

ABSTRACTThe neutron capture cross-section of 237Np has been measured in the neutron energy region from 10 meV to 500 eV. A neutron time-of-flight method was employed using the NaI(Tl) spectrometer in the ANNRI beamline at the Japanese Proton Accelerator Research Complex (J-PARC). The experimental capture yield was derived using the pulse-height weighting technique and an energy-dependent cross-section was obtained relative to the incident neutron spectrum derived from a 10B(n, )7Li reaction yield. The absolute cross-section was determined by normalizing the results to JENDL-4.0 cross-section data at the first resonance of 237Np. The thermal cross-section was measured to be 177.6 ± 3.8 b. The resolved resonance region was analyzed with the REFIT code. An average radiation width = 40.3 ± 0.5 meV, a mean level spacing D0 = 0.60 ± 0.02 eV and a neutron strength function 104S0 = 1.02 ± 0.12 were obtained.

Measurement of the 235U(n, f) cross section relative to the 6Li(n, t) and 10B(n, alpha) standards from thermal to 170 keV neutron energy range at n_TOF

.The 235U(n, f ) cross section was measured at n_TOF relative to 6Li(n, t) and 10B(n, alpha) , with high resolution ( L=183.49(2) m) and in a wide energy range (25meV-170keV) with 1.5% systematic uncertainty, making use of a stack of six samples and six silicon detectors placed in the neutron beam. This allowed us to make a direct comparison of the yields of the 235U(n, f ) and of the two reference reactions under the same experimental conditions, and taking into account the forward/backward emission asymmetry. A hint of an anomaly in the 10-30keV neutron energy range had been previously observed in other experiments, indicating a cross section systematically lower by several percent relative to major evaluations. The present results indicate that the cross section in the 9-18keV neutron energy range is indeed overestimated by almost 5% in the recently released evaluated data files ENDF/B-VIII.0 and JEFF3.3, as a consequence of a 7% overestimate in a single GMA node in the IAEA reference file. Furthermore, these new high-resolution data confirm the existence of resonance-like structures in the keV neutron energy region. The results here reported may lead to a reduction of the uncertainty in the 1-100keV neutron energy region. Finally, from the present data, a value of 249.7pm 1.4(mathrm{stat}) pm 0.94(mathrm{syst}) b·eV has been extracted for the cross section integral between 7.8 and 11eV, confirming the value of 247.5pm 3 b·eV recently established as a standard.

Study on background shielding for a compact photoneutron source

A compact photoneutron source (PNS), based on an electron LINAC was designed and constructed to provide required experimental nuclear data for Thorium Molten Salt Reactor (TMSR). All devices including the LINAC, the neutron production target and the detector systems were arranged in a shared hall, leading to a high background of neutron and γ-ray. The necessary shielding design was made to reduce the backgrounds. A series of preliminary measurements of total cross section in thermal neutron energy region for some key nuclides of TMSR, such as 232Th and 7Li, have been performed at this PNS. It was found, however, that additional shielding needs to be constructed to further reduce the neutron and γ-ray backgrounds. MCNP is a major tool for the background shielding design since it has the calculation capability for a coupled neutron-photon-electron transport problem. A two-step subsection simulation based on MCNP involving with variance reduction techniques (Sub-Var method) was performed to provide required flux and spectrum calculation for the new background shielding design. Based on the simulated results, a new local shielding was designed and constructed, which can reduce the thermal neutron background by three orders of magnitude and the γ-ray background by two orders of magnitude. The neutron flux and energy spectrum were measured under the new shielding condition and compared with the simulated ones, showing a good agreement with each other in the thermal energy region. Therefore, a significant improved shielding for the PNS is achieved, which is expected to provide a required assurance for the future measurements of neutron cross section at the PNS.

Fe56,54(n,α)Cr53,51 cross sections in the MeV region

$^{56,54}\mathrm{Fe}(n,\ensuremath{\alpha})^{53,51}\mathrm{Cr}$ cross sections were measured in the neutron energy region of 5--11 MeV using three neutron sources based on accelerators, a twin gridded ionization chamber, and highly enriched $^{56}\mathrm{Fe}$ and $^{54}\mathrm{Fe}$ foil samples. The $^{238}\mathrm{U}(n,f)$ reaction was used to monitor the neutron fluence, and the interferences from the low-energy neutrons were corrected according to the neutron energy spectra obtained through unfolding the pulse height spectra measured by a liquid scintillator. Both the measured cross sections show a ``shoulder'' structure in the 8--11 MeV region which may be caused by the level structure of the residual nuclei. The present paper aims to improve constraint of the $^{56,54}\mathrm{Fe}(n,\ensuremath{\alpha})^{53,51}\mathrm{Cr}$ excitation functions in the MeV region where their neutron energy dependences are significant, while related measurements are nonexistent or scarce and the deviations among existing measurements and evaluations are significant. To obtain more precise excitation functions of the $^{56,54}\mathrm{Fe}(n,\ensuremath{\alpha})^{53,51}\mathrm{Cr}$ reactions, further measurements and improved evaluations are required.