INFORMATICAInformatica0868-49520868-4952VUINF1221110.3233/INF-2001-12211Research articleThe Solution of Two-Dimensional Neutron Diffusion Equation with Delayed NeutronsSapagovasMifodijusm.sapagovas@ktl.mii.ltVileiniškisVirginijusVytautas Magnus University, Vileikos 8, 3035 Kaunas, LithuaniaInstitute of Mathematics and Informatics, Akademijos 4, 2600 Vilnius, LithuaniaVytautas Magnus University, Vileikos 8, 3035 Kaunas, LithuaniaLithuanian Energy Institute, Breslaujos 3, 3035 Kaunas, Lithuania0101200112233734201112000

The distribution of neutron population in nuclear reactor is described by using transport equations. One of possible approximations of neutron transport equation is given by the neutron diffusion equation. The paper presents numerical solution method of one group neutron diffusion equation with one group of delayed neutrons.

neutron diffusion equationthe method of summary approximationdouble sweep method