title: Pressure vessel calculations for VVER-440 reactors.
abstract: For the determination of the fast neutron load of the reactor pressure vessel a mixed calculational procedure was developed. The procedure was applied to the Unit II of Paks NPP, Hungary. The neutron source on the outer surfaces of the reactor was determined by a core design code, and the neutron transport calculations outside the core were performed by the Monte Carlo code MCNP. The reaction rate in the activation detectors at surveillance positions and at the cavity were calculated and compared with measurements. In most cases, fairly good agreement was found.
authors.affiliation: KFKI Atomic Energy Research Institute, 114, P.O. Box 49, H-1525 Budapest, Hungary. hordosy@sunserv.kfki.hu
authors.forename: G
authors.initials: G
authors.lastname: Hordósy
authors.forename: Gy
authors.initials: G
authors.lastname: Hegyi
authors.forename: A
authors.initials: A
authors.lastname: Keresztúri
authors.forename: Cs
authors.initials: C
authors.lastname: Maráczy
authors.forename: E
authors.initials: E
authors.lastname: Temesvári
authors.forename: P
authors.initials: P
authors.lastname: Vértes
authors.forename: E
authors.initials: E
authors.lastname: Zsolnay
mesh_terms: D000465:Algorithms; D003198:Computer Simulation; D004868:Equipment Failure; D019544:Equipment Failure Analysis / Q000379:methods; D005160:Facility Design and Construction / Q000379:methods; D006814:Hungary; D015233:Models, Statistical; D009010:Monte Carlo Method; D009502:Neutrons; D009688:Nuclear Reactors; D011312:Pressure; D011829:Radiation Dosage; D011834:Radiation Monitoring / Q000379:methods*; D011835:Radiation Protection / Q000295:instrumentation* / Q000379:methods*; D018570:Risk Assessment / Q000379:methods*; D012984:Software