journal article·2023-08

Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

Takafumi Narukawa, Keietsu Kondo, Yuki Fujimura, Kazuo Kakiuchi, Yutaka Udagawa, Yoshiyuki Nemoto

Publisher page

Abstract

To evaluate the behavior of an oxide-dispersion-strengthened FeCrAl (FeCrAl-ODS) cladding tube under loss-of-coolant accident (LOCA) conditions of light-water reactors (LWRs), the following two laboratory-scale LOCA-simulated tests were performed: the burst and integral thermal shock tests. Four burst and three integral thermal shock tests were performed on unirradiated, stress-relieved FeCrAl-ODS cladding tube specimens, simulating ballooning and rupture, oxidation, and quenching, which were postulated during a LOCA. The burst temperature of the FeCrAl-ODS cladding tube was 200– higher than that of the Zircaloy cladding tube, and the FeCrAl-ODS cladding tube`s maximum circumferential strain was smaller than or equal to that of the Zircaloy-4 cladding tube. These results indicate that the FeCrAl-ODS cladding tube has higher strength at high temperatures than the conventional Zircaloy cladding tube. The FeCrAl-ODS cladding tube did not fracture after being subjected to an axial restraint load of ∼ , which is more than 10 times higher than the axial restraint load estimated for existing LWRs, during quenching, following isothermal oxidation at for . The FeCrAl-ODS cladding tube was hardly oxidized during this isothermal oxidation condition. However, it melted after a short oxidation at and fractured after abnormal oxidation at for . Based on these results, the FeCrAl-ODS cladding tube should not fracture in the time range expected during LOCAs below , where no melting or abnormal oxidation occurs.
Source: expertdata · Quality: candidate unverified · Source record ↗

Research context

OpenAlex topics: Nuclear Materials and Properties · Nuclear reactor physics and engineering · Nuclear Engineering Thermal-Hydraulics

SciVX discovery tags are terminology-based. OpenAlex topics are source-provided classifications.

References (39)

Show source-provided references
1. Event sequence of the Fukushima Daiichi accident
2. Accident tolerant fuels for LWRs: a perspective10.1016/j.jnucmat.2013.12.005 ↗
3. Overview of accident-tolerant fuel R&D program in Japan
4. Development of accident tolerant FeCrAl-ODS fuel cladding for BWRs in Japan10.1016/j.jnucmat.2021.153276 ↗
5. Fuel behavior under loss-of-coolant accident situations
6. Meyer · 2014
7. Nagase · 2020
8. Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident
9. Recent progress in development of accident tolerant FeCrAl-ODS fuel claddings for BWRs in Japan
10. Baker · 1962
11. Acceptance criteria for emergency core cooling systems for light-water-reactors, The Nuclear Safety Commission of Japan, Japan, 1981 (in Japanese).
12. OECD/NEA, Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions: State-of-the-art Report, Oganisation for economic co-operation and development nuclear energy agency, Paris (France) 2009.
13. Oxidation mechanism and kinetics of nuclear-grade FeCrAl alloys in the temperature range of 500–1500°C in steam10.1016/j.jnucmat.2022.153696 ↗
14. Stuckert · 2022
15. 2.07 - zirconium alloys: properties and characteristics
16. Microstructure and texture evolution and ring-tensile properties of recrystallized FeCrAl ODS cladding tubes10.1016/j.msea.2019.138636 ↗
17. Failure-bearing capability of oxidized Zircaloy-4 cladding under simulated loss-of-coolant condition10.1080/18811248.1983.9733491 ↗
18. Effect of pre-hydriding on thermal shock resistance of Zircaloy-4 cladding under simulated loss-of-coolant accident conditions10.1080/18811248.2004.9715539 ↗
19. Behavior of pre-hydrided Zircaloy-4 cladding under simulated LOCA conditions10.1080/18811248.2005.9726381 ↗
20. Experimental and statistical study on fracture boundary of non-irradiated Zircaloy-4 cladding tube under LOCA conditions10.1016/j.jnucmat.2017.11.049 ↗
21. Oxidation of Zircaloy-4 under high temperature steam atmosphere and its effect on ductility of cladding10.1080/18811248.1978.9735555 ↗
22. Uetsuka · 1984
23. Oxidation of Zircaloy-4 under limited steam supply from 1,000 to 1.400°C10.1080/18811248.1986.9735077 ↗
24. Oxidation kinetics and related phenomena of Zircaloy-4 fuel cladding exposed to high temperature steam and hydrogen-steam mixtures under PWR accident conditions10.1016/0029-5493(87)90286-X ↗
25. Advanced treatment of Zircaloy cladding high-temperature oxidation in severe accident code calculations: part I. Experimental database and basic modeling10.1016/j.nucengdes.2004.02.013 ↗
26. Nissley · 2005
27. The effect of azimuthal temperature distribution on the ballooning and rupture behavior of Zircaloy-4 cladding tube under transient-heating conditions10.1080/00223131.2016.1158671 ↗
28. Chung · 1978
29. Powers · 1980
30. Practical development of accident tolerant FeCrAl-ODS fuel claddings for BWRs in Japan
31. Pint · 2018
32. On the role of microstructure on the burst properties of oxide dispersion strengthened FeCrAl alloys during simulated loss-of-coolant accident conditions
33. The effect of oxidation and crystal phase condition on the ballooning and rupture behavior of Zircaloy-4 cladding tube-under transient-heating conditions10.1080/00223131.2015.1025884 ↗
34. OECD/NEA · 2016
35. Small Break LOCA Sensitivity Analyses for USAPWR: Mitsubishi Heavy Industries, LTD., Report no. MUAP-07025-NP(R3), 2011.
36. Large Break LOCA Code Applicability Report for US-APWR: Mitsubishi Heavy Industries, LTD., Report no. MUAP-07011-NP(R3), 2013.
37. LOCA simulation test of the cladding for high-burnup fuel
38. Thermal-shock behavior of PWR high-burnup fuel cladding under simulated LOCA conditions
39. Progress on Japanese development of accident tolerant Fe-CrAl-ODS fuel claddings for BWRs