Generating of Thermal Neutron Scattering Data File in CENACE

In order to simulate the calculation of related neutronics problems, an ACE format multiple-temperature continuous energy cross section library CENACE was developed by China Nuclear Data Center. Thermal scattering data of 18 materials from ENDF/B-Ⅶ.1 were processed into ACE format with NJOY99 program on the purpose of calculating problems related to thermal neutron. In order to verify the integrity and availability of this thermal neutron ACE file, drawing test was done for the data, and calculation result of thermal scattering cross section was also compared with the experimental value. The test result shows that all the ACE format data are accurate and reliable, and there is no abnormal and unreasonable phenomenon; for common reactor moderator materials, the new generated thermal neutron scattering data are in good agreement with experimental values except individual data need to be improved.