Sprr-300 Reactor Neutron Fluence Rate Distribution

According to the situation that there is no complete computing data about SPRR-300 reactor in history, the thesis analyzes two calculational code ANISN and MCNP, and calculates the neutron flux in the reactor core, thermal column and the concrete shielding with the code ANISN coupled with MCNP code. It provides scientific basis for reactor radioprotection, application of reactor as a neutron source and dealing of radioactive waste when the reactor is out of commission.This thesis sets up the calculational model of SPRR-300 reactor core with the repeated-structures capability of MCNP to dealing with the complex geometric structure. It not only reflects the space geometric structure, but also simples the water reflector outsides of the core. Moreover, it takes the temperature effect to collision between a neutron and a target atom in account. The computing results on the basis of the model described above confirm well with the experimental results. It shows that not only the computing results but also the experimental results are credible and provides the exact technical data for sample irradiation in the core and research about reactor.As to the computation of the components outsides of reactor core, MCNP code is not very reasonable because it is a deep penetration problem. On the other hand, ANISN code is also not reasonable because the dealing the core uniformly distributed will bring serious error. If we combine ANISN with MCNP code by computing reactor core with MCNP code and regarding the results as the independent neutron source to compute the neutron flux in the thermal column and the concrete shielding, we can solve both assuming the core uniformly distributed and the difficulty of deep-penetration. The calculational results and the experimental are well conformed. It proves that this method above is reasonable and accurate results can be got. So it will be a good new means to compute radioactivity shielding later.