EPJ Nuclear Sci. Technol.
Volume 2, 2016
|Number of page(s)||6|
|Published online||01 February 2016|
Evaluations of Mo-alloy for light water reactor fuel cladding to enhance accident tolerance
Electric Power Research Institute (EPRI), Palo Alto, CA 94304, USA
2 GE Global Research Center, Schenectady, NY 12309, USA
* e-mail: email@example.com.
Received in final form: 26 November 2015
Accepted: 3 December 2015
Published online: 1 February 2016
Molybdenum based alloy is selected as a candidate to enhance tolerance of fuel to severe loss of coolant accidents due to its high melting temperature of ∼2600 °C and ability to maintain sufficient mechanical strength at temperatures exceeding 1200 °C. An outer layer of either a Zr-alloy or Al-containing stainless steel is designed to provide corrosion resistance under normal operation and oxidation resistance in steam exceeding 1000 °C for 24 hours under severe loss of coolant accidents. Due to its higher neutron absorption cross-sections, the Mo-alloy cladding is designed to be less than half the thickness of the current Zr-alloy cladding. A feasibility study has been undertaken to demonstrate (1) fabricability of long, thin wall Mo-alloy tubes, (2) formability of a protective outer coating, (3) weldability of Mo tube to endcaps, (4) corrosion resistance in autoclaves with simulated LWR coolant, (5) oxidation resistance to steam at 1000–1500 °C, and (6) sufficient axial and diametral strength and ductility. High purity Mo as well as Mo + La2O3 ODS alloy have been successfully fabricated into ∼2-meter long tubes for the feasibility study. Preliminary results are encouraging, and hence rodlets with Mo-alloy cladding containing fuel pellets have been under preparation for irradiation at the Advanced Test Reactor (ATR) in Idaho National Laboratory. Additional efforts are underway to enhance the Mo cladding mechanical properties via process optimization. Oxidation tests to temperatures up to 1500 °C, and burst and creep tests up to 1000 °C are also underway. In addition, some Mo disks in close contact with UO2 from a previous irradiation program (to >100 GWd/MTU) at the Halden Reactor have been subjected to post-irradiation examination to evaluate the chemical compatibility of Mo with irradiated UO2 and fission products. This paper will provide an update on results from the feasibility study and discuss the attributes of the coated Mo cladding design to meet the challenging requirements for improving fuel tolerance to severe loss of coolant accidents.
© B. Cheng et al., published by EDP Sciences, 2016
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