Measurement and simulation of the leakage neutron spectra from Fe spheres bombarded with 14 MeV neutrons

Iron is commonly used as a structural and shielding material in nuclear devices. The accuracy of its nuclear data is critical for the design of nuclear devices. The evaluation data of ^(56)Fe isotopes in the latest version of the CENDL-3.2 library from China was significantly updated. This new data must be tested before it can be used. To test the reliability of this data and assess the shielding effect, a shielding benchmark experiment was conducted with natural Fe spherical samples using a pulsed deuterium–tritium neutron source at the China Institute of Atomic Energy(CIAE). The leakage neutron spectra from the natural spherical iron samples with different thicknesses(4.5, 7.5, and 12 cm) were measured between 0.8 and 16 MeV after interacting with 14 MeV neutrons using the time-of-flight method. The simulation results were obtained by Monte Carlo simulations by employing the Fe data from the CENDL-3.2, ENDF/B-VIII.0, and JEDNL-5.0 libraries. The measured and simulated leakage neutron spectra and penetration rates were compared, demonstrating that the CENDL-3.2 library performs sufficiently overall. The simulation results of the other two libraries were underestimated for scattering at the continuum energy level.

Experimental Studies on the Self-Shielding Effect in Fissile Fuel Breeding Measurement in Thorium Oxide Pellets Irradiated with 14 MeV Neutrons

The 14 MeV neutrons produced in the D-T fusion reactions have the potential of breeding Uranium-233 fissile fuel from fertile material Thorium-232. In order to estimate the amount of U-233 produced, experiments are carried out by irradiating thorium dioxide pellets with neutrons produced from a 14 MeV neutron generator. The objective of the present work is to measure the reaction rates of 232Th ＋ in --＊ 233Th --＊ 2a^Pa --＊ 2a3U in different pellet thicknesses to study the self-shielding effects and adopt a procedure for correction. An appropriate assembly consisting of high-density polyethylene is designed and fabricated to slow down the high-energy neutrons, in which Thorium pellets are irradiated. The amount of fissile fuel （~3~U） produced is estimated by measuring the 312 keV gammas emitted by Protactinium-233 （half-life of 27 days）. A calibrated High Purity Germanium （HPGe） detector is used to measure the gamma ray spectrum. The amount of 233U produced by Th232 （n, ~） is calculated using MCNP code. The self-shielding effect is evaluated by calculating the reaction rates for different foil thickness. MCNP calculation results are compared with the experimental values and appropriate correction factors are estimated for self-shielding of neutrons and absorption of gamma rays.

Cross section measurements on zirconium isotopes for ~14 MeV neutrons and their theoretical calculations of excitation functions

The cross sections for the 94Zr(n,d*)93m+gY,96Zr(n,γ)97Z,96Zr(n,2n)95Zr,90Zr(n,α)87mSr,94Zr(n,α)91Sr,90Zr(n,p)90mY,92Zr(n,p)92Y,and 94Zr(n,p)94Y reactions have been measured in the neutron energy range of 13.5-14.8 MeV by means of the activation technique.The neutrons were produced via the D-T reaction.A high-purity germanium detector with high energy resolution was used to measure the inducedγactivities.In combination with the nuclear reaction theoretical models,the excitation curves of the above-mentioned eight nuclear reactions within the incident neutron energy range from the threshold to 20 MeV were obtained by adopting the nuclear theoretical model program system Talys-1.9.The resulting experimental cross sections were analyzed and compared with the experimental data from published studies.Calculations were performed using Talys-1.9 and are in agreement with our experimental results,previous experimental values,as well as results of the theoretical excitation curves at the corresponding energies.The theoretical excitation curves generally match the experimental data well.

Cross-section measurements of(n,2n)and(n,p)reactions on ^(124,126,128,130,131,132)Xe in the 14 MeV region and theoretical calculations of their excitation functions

The reaction cross-sections of ^(124)Xe(n,2n)^(123)Xe,^(126)Xe(n,2n)^(125)Xe,^(128)Xe(n,2n)^(127)Xe,^(130)Xe(n,2n)^(129)mXe,^(132)Xe(n,2n)^(131)mXe,^(130)Xe(n,p)^(130)I,^(131)Xe(n,p)^(131)I,and ^(132)Xe(n,p)^(132)I were measured at the 13.5,13.8,14.1,14.4,and 14.8 MeV neutron energies.The monoenergetic neutrons were generated via the ^(3)H(d,n)^(4)He reaction at the China Academy of Engineering Physics using the K-400 Neutron Generator with a solid ^(3)H-Ti target.A high-purity germanium detector was employed to measure the activities of the product.The reactions ^(93)Nb(n,2n)^(92m)Nb and ^(27)Al(n,α)^(24)Na were adopted for neutron flux calibration.The cross sections of the(n,2n)and(n,p)reactions of the xenon isotopes were obtained within the 13-15 MeV neutron energy range.These cross-sections were then compared with the IAEA-exchange format(EXFOR)database-derived experimental data,together with the evaluation results of the CENDL-3,ENDF/B-Ⅷ.0,JENDL-4.0,RUSFOND,and JEFF-3.3 data libraries,as well as the theoretical excitation function obtained using the TALYS-1.95 code.The cross-sections of the reactions(except for the ^(124)Xe(n,2n)^(123)Xe and ^(132)Xe(n,p)^(132)I)at 13.5,13.8,and 14.1 MeV are reported for the first time in this study.The obtained results are beneficial in providing better cross-section constraints for the reactions in the 13-15 MeV region,thus improving the quality of the corresponding database.Meanwhile,these data can also be used for the verification of relevant nuclear reaction model parameters.

Monte Carlo simulation of GEM-based for 14 MeV neutron detector

A new concept of neutron detector based on Gas Electron Multiplier(GEM) technology is presented in this paper,in which a novel multi-layer high density polyethylene(HDPE) as neutron-to-proton converter is proposed and studied with Geant4 toolkit for fast 14 MeV neutron.Our preliminary results show that the detection efficiency of the detector with 400 converter units is higher than 2.3% and reconstruction accuracy of the incident neutron position is higher than 2.6%.

Cross-section measurements for(n,2n)reactions on stannum isotopes in the neutron energy range of 13.5 to 14.6 MeV

Cross-sections for （n, 2n） reactions have been measured on stannum isotopes at the neutron energies of 13.5 to 14.6 MeV using the activation technique. Data are reported for the following reactions： ^112Sn（n, 2n）^111Sn, ^118Sn（n, 2n）^117Sn and ^124Sn（n, 2n）^123mSn. The neutron fluences were determined using the monitor reaction ^93Nb（n, 2n）^92mNb or ^27Al（n, α）^24Na. The results of present work were compared with data published previously.

Measurement and analysis of the ^(238)U(n, 2n) reaction rate in depleted uranium/polyethylene shells

In order to check the conceptual design of the subcritical blanket in a fusion-fission hybrid reactor, a depleted uranium/polyethylene simulation device with alternate shells has been established. The measurement of the 23SU（n, 2n） reaction rate was carried out using an activation technique, by measuring the 208 keV T rays emitted from 237U. The self-absorption of depleted uranium foils with different thicknesses was experimentally corrected. The distribution of the 23SU（n, 2n） reaction rate at 90° to the incident D＋ beam was obtained, with uncertainty between 5.3% and 6.0%. The experiment was analyzed using MCNP5 code with the ENDF/BVI library, and the calculated results are all about 5% higher than the measured results.