Austenitic stainless steel 316 has very high mechanical properties and corrosion resistance. This type of steel is widely used both in the nuclear and non-nuclear industries. In the nuclear industry, SS316 is used as a cladding material for uranium fuel due to its good corrosion and mechanical properties, and also low neutron absorption cross-section. In the Center for Nuclear Fuel Technology (PTBBN BATAN), it is used as material for the container of nuclear waste that is to be stored on Temporary Storage Installation of Spent Fuel (KHIPSB3 BATAN). SS316 is used as material for can to contain high-activity solid waste from the testing activity in Radiometallurgy Installation (IRM BATAN). The lid of the container is sealed with the GTAW welding process in order to tightly contain the solid waste. The main problem with the heat treatment and welding process of austenitic stainless steel is the occurrence of sensitization in a temperature range of 500-800°C. Therefore fully electrochemical analysis of SS316 stainless steel in various mediums has been conducted. SS 316 specimen was heat-treated to simulate the heat generated by the welding process. Sensitization analysis was conducted with qualitative and quantitative methods by EIS and EPR, and pitting corrosion resistance was by cyclic polarization method. The solution used for EIS and cyclic polarization test was 0 − 3.5% concentration of NaCl, and for EPR test was a mixture of H2SO4 and KSCN. Material characterization before and after corrosion testing was microstructure examination. The result of the corrosion test showed that heat treatment on the temperature range of 500-800°C caused sensitization. The corrosion test curve result showed that a heat treatment temperature of 650°C for 1 hour had the highest activation current. The lowest Rp value for SS316 specimen post heat treatment in 675°C was 69.410 ohm. The welded SS316 specimen had a higher corrosion current than that of the unwelded specimen. The microstructure of the welded specimen showed that there was intergranular corrosion particularly in the HAZ region. The effect of NaCl concentration on the cyclic polarization test showed that the higher the NaCl concentration, the more easily the pitting corrosion. The indication of pitting corrosion occurrence was evaluated by considering the Epit and Erp values. The lower Epit value meant that pitting corrosion was more easily to occur.

STUDY OF FISSION GAS BUBBLES AND INTERACTION LAYER ON IRRADIATED U3Si2-Al DENSITY OF 4.8 gU/cm3. Uranium-silicide compound fuel dispersed in aluminium matrix (U3Si2-Al) have been used in a large number of research reactors around the world because of its excellent behavior under irradiation. This fuel also provides high uranium density with typical fuel loading up to 4.8 gU/cm3 to compensate for the reduced fissile amount in LEU. To improve the density of current U3Si2-Al (2.96 gU/cm3) used in Indonesian GA Siwabessy Multipurpose Research Reactor, U3Si2-Al dispersion fuel plate with density of 4.8 gU/cm3 (U235 ∼19.75%) had been irradiated in RSG GAS for 175 days at 15 MW power to burnup level of approximately 40%. The characterization was performed using SEM-EDS and optical microscope to study microstructure of the irradiatted fuel, largely the fission gas bubbles and the interaction layer between U3Si2 fuel and Al matrix. The average diameter of the bubbles with diameter from 0.06 to 0.55 µm was 0.21 µm. The interaction layer was identified as U(Al,Si)2,3 with thickness of approximately 1.5 µm. The relatively small fission gas bubbles and the interaction layer didn’t cause swelling on the fuel and the overall performance of the fuel plate was very good.Keyword:  LEU, uranium-silicide, post-irradiation examination, interaction layer, fission gas bubbles.

OXIDE LAYER CHARACTERIZATION OF AlMg2 CLADDING OF IRRADIATED U3Si2/Al FUEL WITH 4,8 gU/cm3 DENSITY. To investigate the performance of AlMg2 cladding in the U3Si2/Al dispersion fuel, oxide layer characterization of AlMg2 cladding of the irradiated U3Si2/Al fuel with 4.8 gU/cm3 density was conducted. The oxide layer  on the surface of AlMg2 cladding is one of the changes that occur on the cladding after the U3Si2/Al fuel plate has been irradiated in the RSG-GAS reactor to a burn-up of ∼40%. The characterization and observation of the oxide layer was conducted using SEM (Scanning Electron Microscope) and Energy-dispersive X-ray spectroscopy (EDS). Samples with a size of 3x3 mm were taken from the middle of the fuel plate (middle position). After cutting, metallographic preparation includes mounting, grinding, polishing, and ultrasonic cleaning. SEM preparation was carried out by sputter coating using Au layer. The oxide layer on the AlMg2 cladding has a thickness of 10.3 µm with a uniformly distributed cracks along the oxide layer.Keyword: LEU, uranium-silicide, post-irradiation examination, AlMg2 cladding, oxide layer.

PENGARUH PENAMBAHAN LARUTAN H3BO3 DAN LiOH TERHADAP PERILAKU KOROSI MATERIAL KELONGSONG ZIRCALOY-2 DALAM MEDIA AIR BEBAS MINERAL. Pendingin primer pada reaktor tipe PHWR dikendalikan secara kimia dengan penambahan asam borat dan LiOH untuk mengantisipasi korosi pada kelongsong bahan bakar yang menggunakan material zircaloy-2 (Zr-2.) Penelitian ini bertujuan untuk menganalisis pengaruh penambahan larutan H3BO3 dan LiOH terhadap jenis dan laju korosi material kelongsong bahan bakar Zr-2 dalam media air bebas mineral. Penambahan bahan H3BO3 dan LiOH pada pendingin reaktor PHWR adalah untuk pengendalian secara kimia. Uji korosi dilakukan pada temperatur kamar untuk menghilangkan pengaruh temperatur tinggi dan tekanan pada proses elektrokimia. Pada penelitian ini dilakukan pengamatan laju korosi material Zr-2 di dalam media air bebas mineral dengan penambahan H3BO3 dan LiOH dengan variasi konsentrasi. Hasil pengamatan, laju korosi Zr-2 menggunakan metode Tafel pada dalam media air bebas mineral dengan penambahan H3BO3 pada konsentrasi 10 ppm, 100 ppm, 500 ppm, 1000 ppm dan 2000 ppm diperoleh laju korosi masing masing 17,29 x 10-3 mpy, 18,51 x 10-3 mpy, 20,82 x 10-3 mpy, 22,71 x 10-3 mpy dan 23,29 x 10-3 mpy. Setelah air bebas mineral, kemudian ditambahkan H3BO3 sebesar 2000 ppm dan LiOH dengan konsentrasi 1 ppm, 2 ppm, 3 ppm, 4 ppm. Hasil analisis menunjukkan bahwa dengan penambahan H3BO3 dan LiOH diperoleh laju korosi semakin menurun secara berurutan yaitu 22,71 x 10-3 mpy, 21,88 x 10-3 mpy, 21.41 x 10-3 mpy, 21,39 x 10-3 mpy, dan penambahan LiOH 5 ppm menyebabkan laju korosi meningkat menjadi 21,45 x 10-3 mpy. Hasil penelitian ini dapat disimpulkan bahwa penambahan LiOH dan H3BO3 berpengaruh terhadap laju korosi material zircaloy-2 dalam media air bebas mineral. Semakin tinggi konsentrasi H3BO3 yang ditambahkan menyebabkan laju korosi semakin meningkat, sedangkan dengan penambahan LiOH menyebabkan laju korosi semakin menurun hingga konsentrasi 4 ppm Namun dengan penambahan LiOH pada konsentrasi 5 ppm menyebabkan laju korosi meningkat sehingga dapat disimpulkan bahwa konsentrasi optimum penambahan LiOH adalah 4 ppm.Kata kunci: zircaloy-2, laju korosi, H3BO3, LiOH, kelongsong