2016 Annual Meeting

Presentation information

Oral Presentation

IV. Nuclear Fuel Cycle and Nuclear Materials » 402-1. Nuclear Materials and the Irradiation Behavior

[2H01-07] Zr alloys

Sun. Mar 27, 2016 9:30 AM - 11:15 AM Room H (Lecture Rooms B B204)

Chair: Kazunori Morishita (Kyoto Univ.)

11:00 AM - 11:15 AM

[2H07] Effect of interal pressure on oxidation of Zircaloy-2 cladding tube under steam

*Ikuo Ioka1, Hitoshi Kato1, Hiroaki Ogawa1, Masahiko Osaka1 (1.Japan Atomic Energy Agency)

Keywords:Zircaloy-2, Internal pressure, Steam oxidation, Spent fuel pool

When the function of cooling system for a spent fuel pool loses, the spent fuel pin put under vapor environment. The helium gas at the time of making and the FP gas are enclosed in the spent fuel pin. The spent fuel pin with internal pressure is oxidized during severe accident. The effect of internal pressure on oxidation behavior of Zircaloy-2 was investigated. The oxidation test was carried out at 600 °C in steam atmosphere. The length of Zircaloy-2 specimen is 500 mm, and heated zoon is 200 mm. The amount of hydrogen gas was measured during the test, and the oxide film was analyzed. The amount of hydrogen gas decreased to about 70 ppm after increasing rapidly in early stages of the test. The thickness of oxide film decreased with increasing the internal pressure. No remarkable difference in cross section of the oxide film was observed in this test condition.