Benchmarking Of Evaluated Neutron Nuclear Data For ²³⁸U And Measurement Of Spallation Neutrons Induced By 296 MeV Deuterons Relevant To ADS

Accelerator driven sub-critical system?ADS?is inherently safe and can make the most of uranium or thorium to breed nuclear fuel.It can also transmute nuclear waste effectively.Thus the problems of nuclear energy on safety,nuclear fuel supply and nuclear waste process can be solved or alleviated.It has good prospects for development.ADS utilizes proton beam of high energy and intensity bombarding heavy metal spallation target as an external neutron source with high flux and broad spectrum to drive the subcritical core.Its operational characteristics are largely due to the distribution of source neutrons and the interactions between neutrons and various materials in the core.The design of ADS requires a large amount of precise spallation neutron data and neutron nuclear reaction data.According to the requirements for neutronics of ADS spallation target and the key material ²³⁸U,research work on experimental measurement of spallation neutron energy spectra as well as validation of²³⁸U neutron nuclear data were carried out,which has important sense to the neutronic design of ADS.Leakage neutron spectra from ²³⁸U slab samples in size of 10 cm×10 cm×2 cm,10 cm×10 cm×5 cm and 10 cm×10 cm×11 cm interacting with 14.8 MeV D-T neutrons were measured at 60°and 120°by Time-of-Flight method at the neutron integral experimental setup of Nuclear Data Key Laboratory,China Institute of Atomic Energy?CIAE?.Calculations of the experiments were carried out by MCNP Monte Carlo code with the main neutron evaluated nuclear data libraries?ENDF/B-VII.1,JENDL-4.0,CENDL-3.1,TENDL-2015,JEFF-3.2,ENDF/B-VIII.0 and JEFF-3.3?in the world to validate the neutron evaluated nuclear data for ²³⁸U.Meanwhile,neutron fluxes and reaction rates in the sample were calculated in detail to analyze the contributions of different reaction channels to the leakage neutron spectra.It can be seen that calculated results with JENDL-4.0 fitted the experimental data perfectly.The results calculated with CENDL-3.1 overestimated largely in 2.5-8 MeV,which was due to the overestimation of the fission neutron energy distribution in this region.Compared with ENDF/B-VII.1,²³⁸U neutron data given in ENDF/B-VIII.0 have improved a lot.The results calculated with ENDF/B-VIII.0 eliminated the overestimation of results calculated with ENDF/B-VII.1 in 8.5-15.5 MeV.JEFF-3.3 has improved to some extent with regard to JEFF-3.2.But there still exists large discrepancies between the calculated results and the experimental data.Combined samples of different materials were bombarded with D-T neutrons.And leakage neutron spectra at 60°were measured by Time-of-Flight method to study neutron transportation in different materials.Materials in the combined samples were tungsten,uranium,graphite and polyethylene,which were added up one by one along the neutron flight path.Some evaluated nuclear data files?tungsten:ENDF/B-VII.1,uranium:JENDL-4.0,graphite and polyethylene:JENDL-4.0 or ENDF/B-VII.0?were selected according to the benchmarking results of these materials.MCNP was used to calculate the leakage neutron spectra from the combined samples,direct contribution of the sub-samples to the leakage neutron spectra as well as the neutron yields with these nuclear data files.It was found that the calculated results fitted the experimental data quite well,which proved the correctness of their benchmarking results.Graphite and polyethylene have strong ability of scattering neutrons.When they were added to the combined samples,direct contribution to the leakage neutron spectra from the uranium sample declined significantly.Polyethylene is better than graphite for neutron scattering.In addition,in order to acquire more information during calculation,the integral experimental setup was simulated with GEANT4 Monte Carlo code.The feasibility of the calculation method was confirmed since the calculated results were basically the same as the results calculated with MCNP.At HIRFL-CSR nuclear data experimental terminal,a beryllium target in size of?3.5 cm×24 cm and a lead target in size of?5 cm×1 cm were bombarded with 296MeV deuterons.Spectra of neutrons emitted at various angles were measured with Time-of-Flight method.Calculations were performed by using GEANT4 and MCNPX Monte Carlo codes with various spallation reaction models.Beryllium can be used as reflector in reactors,neutron multiplication material in fusion facilities,beam window of accelerator and target material of neutron source.Lead is a very important material as spallation target and reactor coolant.Studying the spallation neutron characteristics of these materials is instructive for the design of the coupled system of spallation target and reactor core.In addition,neutron spectra at different angles for 256 MeV protons bombarding different materials were calculated with GEANT4 and FLUKA Monte Carlo codes to validate the spallation reaction models.It can be seen that results calculated with INCL model can fit the corresponding experimental data much better.