IRRADIATION INDUCED-HARDENING AND EMBRITTLEMENT OF REACTOR PRESSURE VESSEL STEELS AS REVEALED BY ATOMIC LEVEL SIMULATIONS

A THERMAL HYDRAULIC MODEL FOR POWER UPGRADING OF WWR-S RESEARCH REACTOR

Thermal Hydraulic Evaluation of a proposed annular Fuel for VVER1000 Reactor

Mechanical and gamma ray attenuation properties of N316L steel treated by rf plasma as a nuclear material

Calculation of Corrosion Rate in Ex-reactor Considering the Reactor Effects on the (VVER-1000) fuel clad

High Borated Stainless Steel Alloys for Nuclear Reactor Domains

Characterization of Nuclear Properties of Fe- based Alloys for Application in Nuclear Domain

SHIELDING ASSESSMENT OF THE ET:RR-1 REACTOR UNDER POWER UPGRADING.

Stresses Induced by LOCA in Steam Generator Collector of VVER1000

Sub-channel thermal-hydraulic analysis for VVER- 1000 generation III PWR