MECHANICAL STUDY OF 9CR-316SS-1MO FOR NUCLEAR POWER REACTOR FUEL CLADDING MATERIALS

Abstract

Cladding or cladding used for nuclear fuel elements requires special requirements such as good mechanical properties, radiation resistance, corrosion resistance, adequate physical and chemical properties, and low thermal neutron absorption. Materials that are widely used as structural materials for nuclear reactors are beryllium, magnesium, aluminum, zirconium and their alloys. Alloy 9Cr-316-1Mo as Nuclear fuel cladding material. 9Cr-316-1Mo material has been made. The material 9Cr-316-1Mo was chosen as an alternative cladding material because it has good mechanical and chemical properties with a microscopic cross-section of 2.6 barn. At first, the 9Cr-316-1Mo sample was cast withThe composition is 90% 316L Stainless Steel, 9% Chromium (Cr) and 1% Molybdenum (Mo) is added using an electric arc. After casting the alloy,the composition of 9Cr-316-1Mo was tested without heating and then given heat treatment (tempering) starting from 100 0C, 200 0C, 300 0C, 400 0C and 500 0C and allowed to stand for 60 minutes, then tested for hardness (rockwell) , Impact testing, micro structure testing and Chemical composition testing (XRF). The results of the Rockwell hardness test obtained the highest hardness value in the sample without tempering, namely 20.66 HRc and the lowest hardness value in the sample with tempering 500 0C, namely 16.33 HRc. Then the impact test results obtained the highest impact energy value in the sample with tempering 100 0C, which is 0.9996 Joule/mm2 and the lowest impact energy value in the sample with tempering 500 0C, which is 0.9962 Joule/mm2. The results of the microstructure test showed that the structure of 9Cr-316-1Mo still contained ferrite, pearlite and martensite. The testing process is still not fully completed and will be reported later.