Research on Neutron Distribution in Reactor and Design of Radiation Shielding
|School||Harbin Engineering University|
|Course||Radiation protection and environmental protection|
|Keywords||MCNP code Hexagonal core Core physics calculations Shielding calculation|
Due to the rapidly increasing demand of energy , nuclear power world wide attention to the sharp increase in the number of nuclear power plants , reactor type innovation . Reactor design based on our long-term consideration of autonomy , the existing mature type of reactor and reactor physics calculations and analysis, and assessment of the safety of the reactor . In this paper, Monte Carlo ( MC ) calculation code MCNP core physical characteristics , including the critical state of the neutron multiplication factor Keff , neutron spectrum , neutron and photon flux density distribution of the power density distribution of these parameters . Due to the relatively large size of the reactor , the neutron and photon flux density in the core and reactor components by many orders of magnitude difference , the results are not satisfactory critical source power density nodal method to calculate this deep penetration shielding calculation , basket , pressure vessels , the average neutron flux density . The nodal method using a common source, and the need to give the correct source description . The results show that the core neutron and photon flux density on the radial and axial distribution are in good agreement with the actual , more gentle pressure vessels and the flux density in a heap outside the detection area , the shielding effect is ideal . First , calculate the power density distribution , and the calculation result as the second step shielding calculation data base provides a reasonable solution for large-scale reactor shielding .