Search

Radiative neutron capture cross-section and the resonance parameter are very important nuclear data in the field of the R&D of accelerator driven systems, the transmutation of nuclear waste and nuclear astrophysics. In the recent years, radiative neutron capture crosssections of important nuclides were mostly measured based on the white neutron source facilities, such as the CERN n_TOF, GELINA, LANSCE and so on. The back streaming white neutron beam line (Backn) of China Spallation Neutron Source (CSNS) is the first white neutron facility in China, which can provide pulsed neutron beam with 25 Hz frequency in the energy region from thermal neutron to hundreds MeV. The main application of Backn is to perform nuclear data measurement, such as crosssection, reaction yield, nuclear structure, and so on. The C6D6 liquid scintillation detector is widely used for radiative neutron capture crosssection measurements on timeofflight facilities, because of its low neutron sensitivity and good time response. A C6D6 detection system with four detectors was installed at Backn to perform radiative neutron capture crosssection measurement, especially in the resonance energy region. In this work, the measurement of 169Tm(n,γ)170Tm reaction was performed with the C6D6 detection system at the Backn. The stable isotope of thulium, 169Tm, is one of the fission product poisons, which makes the accurate neutron radiative capture crosssection of 169Tm be of major significance for fission and fusion reactor design. Besides, the radioactivity induced by the reactions of 169Tm(n,γ)170Tm and 170Tm(n,γ) 171Tm makes the 169Tm to be an ideal spectrumsensitive activation detector for the neutron intensity diagnosis, in which the precise neutron capture crosssection is of critical importance. In the present measurement, the pulse height weighting technique was used to make detection efficiency be proportional to the excitation energy of the compound nucleus. The black filter method was used to determine the experimental background due to scattered neutron and inbeam gamma ray. The saturated resonance method was used to normalize the capture yield. The Rmatrix code SAMMY was used to fit the capture yield of 169Tm to obtain the resonance parameters, such as resonance energy, neutron width, radiative capture width, and so on. Then the radiative capture crosssection was calculated with the measured resonance parameters and ReichMoore approximation. For most energy region between 1 eV and 100 eV, the present measurement result is consistent with the recommended value of ENDF/BⅧ.0, but there were still some discrepancies between the measured crosssection and the evaluated data. According to the current results, the measurement and data analysis methods used in this work are suitable for the measurement of radiative capture crosssection and resonance parameter. However, in order to perform high precision measurement of radiative capture crosssection at Backn, the neutron flux and the energy resolution function still need to be determined more accurately.

In the fusion reactor, the first wall, blankets and superconducting magnets are directly exposed to 14.5 MeV neutrons produced by fusion reaction. Some of the characteristics of Nb metal such as high melting point, good thermal conductivity and low neutron capture cross section make Nb often used in these parts of reactor. Therefore, the quality of the evaluated nuclear data of Nb is required more accurate to ensure the safety and economy in the process of the fusion reactor working. Benchmark experiments are an important way to test and evaluate the reliability of data. A benchmark experiment on Nb was carried out with 145 MeV DT neutron source in China Institute of Atomic Energy. Slab Nb with thickness of 5 cm, 10 cm and 15 cm was selected as samples. For each thickness of Nb, the leakage neutrons from 08 MeV to 16 MeV were measured using the timeofflight method in the 60° and 120° directions. The Monte Carlo neutron transport code MCNP4C was used to simulate the leakage neutron spectra under the same experimental conditions, the Nb data was retrieved from CENDL-3.1, ENDF/B-Ⅷ.0 and JENDL-4.0 evaluation data libraries. The whole time of flight spectrum was divided into the parts contributed by four reaction channels including (n, el), (n, inl)D, (n, inl)C, and (n, 2n). The simulation integral value of each part was compared with the experimental integral value, and the C/E was obtained as the standard to verify the reliability and accuracy of the relevant data. As the results of the benchmark experiment on Nb, it can be found that: Except a little overpredict at the simulation result of elastic scattering at 120°, the Nb nuclear data of the JENDL40 data library is in good agreement with experiment results. One of the problems of CENDL31 library are that it gives too large discrete inelastic scattering cross section. The other problem may be that the secondary neutron energy spectrum given by continuous level of inelastic scattering and (n, 2n) reaction is soft. The simulation results of the ENDF/B-Ⅷ.0 data library make a huge difference. It can be seen from the comparison results that the cross section of the energy spectrum in the discrete inelastic scattering part is high and the energy spectrum given in the continuum inelastic scattering part is also problematic. The discrepancies of elastic scattering peak between 60° and 120° are considered as caused by the wrong angular distribution, especially the underestimate at 60°. So the ENDF/BⅧ.0 library has a lot to improve. In a conclusion, these three libraries are all need to be improved to get more accurate data.

Bismuth is widely used in advanced nuclear reactors such as lead-bismuth eutectic (LBE) fast reactors, space reactors and so on. Its neutron nuclear data, especially the inelastic scattering cross sections have significant impact on safety and economics of these nuclear facilities. A facility for measurement of neutron reaction cross sections using prompt γ ray method based on the HI13 tandem accelerator was established at China Institute of Atomic Energy (CIAE). Neutrons were produced using a D2 gas target bombarded with deuterons at 65, 80 and 95 MeV. The neutron target was positioned in the center of a shielding in size of 2 m×2 m×2 m composed of concrete, iron, lead and borated polyethylene. Neutrons were extracted at 0° with respect to the deuteron beam through a collimator made of copper, iron, polyethylene and lead. Neutron inelastic scattering cross sections at 90, 105 and 120 MeV were measured experimentally for 209Bi with a 209Bi sample in size of 50 mm×4 mm. A natural titanium sample in size of 50 mm×1 mm was used as the reference sample. By measuring production cross sections of the 983.5 keV γ rays produced when neutrons scattered inelastically off 48Ti nuclei, a normalization factor can be determined. 4 Clover detectors were used to measure produced γ rays. They were placed at 30°, 70°, 110° and 150° with respect to the deuteron beam. Lead shielding in thickness of approximately 2 cm was used for detector head to shield scattered γ rays. Timeofflight method was used to determine neutron energy, thus neutrons produced by DD breakup could be discriminated with neutrons produced by DD fusion. The energy and absolute efficiency were calibrated using 152Eu, 60Co, 22Na and 133Ba standard γ sources with known radioactive activities. A 508 cm×508 cm liquid scintillation detector was positioned at the end of neutron beam line to monitor the neutron fluxes passing through the samples. Two specific measurement angles at 110° and 150° were selected to obtain angleintegrated γ production cross sections by performing a simple linear summation of the partial cross sections measured at these angles. Talys 195 code was used to calculate ratios of γ production cross sections and total inelastic scattering cross sections. The experiment results were compared with other experiment data, data retrieved from ENDF/BⅧ0, JEFF33, JENDL40, ROSFOND2010 and CENDL31 evaluated nuclear data libraries as well as calculation by Talys 195 code with the default parameters. The results show that the tendency of these 3 measured energy points is similar to these results. For cross sections measured at 90 and 105 MeV, the results are closer to data calculated with Talys 195 code. For cross section measured at 120 MeV, it fits the ROSFOND2010 evaluated data better.

In order to validate the evaluated nuclear data, leakage spectra in the range of 0.8 to 15 MeV from samples were measured by time-of-flight (TOF) technique using a D-T neutron source. An experimental system for benchmark validation of nuclear data with slab samples has been set up at China Institute of Atomic Energy (CIAE). In this study, test samples are iron slabs, of which the thickness are 5cm, 10cm and 15cm, and the measured angles were chosen to be about 60° and 120°. By comparing measured leakage spectrum with calculated ones by MCNP-4C code, using the data from the CENDL-3.1, ENDF/B-VIII.0, JENDL-4.0 and JEFF-3.3 nuclear data files, and the comparison was made by the spectrum shape and by the C/E values in different energy regions.

Benchmarking of evaluated neutron nuclear data libraries was performed for ∼14.8 MeV neutrons on the several targets, such as gallium, graphite, silicon carbide, uranium and tungsten samples. The experiments were performed at China Institute of Atomic Energy (CIAE). The neutron leakage spectra from the samples were measured at 60◦ and 120◦ by a TOF technique with a BC501A scintillation detector. The measured spectra are rather well reproduced by MCNP-4C simulations with the CENDL3.1, JENDL-4.0 and the new release ENDF/B-VIII.0, JEFF-3.3 evaluated nuclear data libraries and so on. There have some difference between experiments and simulations for the elastic and inelastic contributions in the partial energy range. And the discrepancies of the neutron leakage spectra in the MCNP simulations originate simply from the differences in the spectra distributions of the neutron reaction channels in the evaluated nuclear data libraries.

Radiative neutron capture cross sections are very important in the field of basic physics research and nuclear device R&D. The Back-n white neutron beam line of China Spallation Neutron Source (CSNS) is the first spallation neutron beam line in China. On the purpose for radiative neutron capture cross section measurement, a C6D6 detector system was built in the Back-n experimental station. The pulse height weighting technique (PHWT) was used to make the system’s detective efficiency independent of the cascade path and the energy of cascade gamma rays. The neutron energy spectrum was measured for the energy between 1eV and 80keV with a 6Li loaded ZnS scintillation detector. Besides, a testing experiment with 197Au and 169Tm samples was carried out to examine this system. According to the preliminary results, this C6D6 detector system can be used to perform neutron capture cross section measurement.

China Institute of Atomic Energy (CIAE) is one of the most important bases for nuclear data measurement in China. In this paper, the nuclear data measurement activities at CIAE in recent years are presented. The secondary neutron emission DX and DDX measurement, the integral experiment for nuclear data benchmarking, the excitation function measurement, the neutron induced fission yields measurement and the gamma production yields measurement carried out in recent years are introduced. Furthermore, the progress of some new facilities and proposed plans (e.g. A Gamma ray Total Absorption Facility (GTAF) for (n,γ) reaction cross section measurement, A 3He detector array for (n,2n) reaction cross section measurement, the back streaming white neutron beam for nuclear data measurement at China Spallation Neutron Source (CSNS)) are also presented.

The half-life of 95mTc and the cross-sections of the 96Ru (n, x) 95mTc reaction induced by D–T neutrons were measured through the neutron activation technique in combination with off-line γ-ray spectrometry. The neutron beam was generated from the T (d, n) 4He reaction using the K-400 neutron generator at the Chinese Academy of Engineering Physics (CAEP). Through exponential function fitting and a detailed discussion of the uncertainty evaluation, the measured half-life of 95mTc was 61.88 ± 0.22 days, which uncertainty is reduced greatly compared with the currently recommended value. Based on the determination of the 95mTc half-life, the cross-sections of 96Ru (n, x) 95mTc reaction at the 13.85 ± 0.2, 14.30 ± 0.2 and 14.72 ± 0.2 MeV neutron energies were measured relative to the 93Nb (n, 2n) 92mNb monitor reaction. Considering the correlations between different attributes, detailed uncertainty propagation was performed by the covariance analysis and the cross-sections were reported with their uncertainties and correlation matrix. Then, experimentally determined cross-sections were analyzed by comparing with the literature data available in the EXFOR database and theoretically calculated values using the TALYS-1.95 and EMPIRE-3.2.3 codes. The accuracy of current experimental results with the thorough uncertainties and covariance information is greatly improved, which is critical for verifying the reliability of the theoretical model and improving the quality of the nuclear database. [ABSTRACT FROM AUTHOR]

The secondary neutron emission differential and double-differential cross sections (DX and DDXs) of n + Be have been measured at the neutron energy of 21.94MeV using the multi-detector fast neutron time-of-flight (TOF) spectrometer. The data was derived by comparing the measured TOF spectra with detailed Monte Carlo simulation, and corrected with n-p scattering cross section. Meanwhile, theoretical calculations based on the Hauser-Feshbach and exciton model have been performed to compare with experimental data. Measured differential cross sections were also compared with other measurements. It was found that the experimental results were in agreement with other measurements and theoretical calculations, while discrepancies were also present in the whole energy region and at some angles. [ABSTRACT FROM AUTHOR]