Behavior of irradiated zivcaloy-4 fuel cladding under simulated LOCA conditions

被引：11

作者：

Ozawa, M ^{[1
]}

Takahashi, T ^{[1
]}

Homma, T ^{[1
]}

Goto, K ^{[1
]}

机构：

[1] Nucl Dev Corp, Dept Res, Fuel & Mat Engn Branch, Tokai, Ibaraki 3191111, Japan

来源：

ZIRCONIUM IN THE NUCLEAR INDUSTRY: TWELFTH INTERNATIONAL SYMPOSIUM | 2000年 / 1354卷

关键词：

zirconium; zirconium alloys; cladding tubes; loss-of-coolant accident; zirconium-steam reaction; thermal shock; nuclear materials; nuclear applications; radiation effects;

D O I：

10.1520/STP14304S

中图分类号：

TL [原子能技术]; O571 [原子核物理学];

学科分类号：

0827 ; 082701 ;

摘要：

High-temperature oxidation and mechanical strength under a loss-of-coolant accident (LOCA) were investigated using irradiated and unirradiated fuel cladding. Cladding from fuel rods irradiated up to 49 GWD/MTU in a Japanese commercial PWR and unirradiated cladding, most of which was preoxidized and precharged with hydrogen to simulate high burnup fuel, were subjected to the tests. High-temperature oxidation tests showed that the oxidation weight gain for irradiated cladding was equal to, or slightly lower than, that for unirradiated material. Preoxidized cladding showed less oxidation weight gain, and no effect of hydrogen absorption on oxidation behavior was observed. The mechanical tests (uniaxial strength and ductility) after a thermal sequence simulating a LOCA showed comparable behavior with that of unirradiated cladding, due to the recovery of the irradiated microstructure. These test results suggest: (1) there is no adverse effect of irradiation on the high-temperature oxidation behavior: and (2) radiation damage in cladding is eliminated during a LOCA condition.

引用

页码：279 / 299

页数：21

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