失冷事故条件下fecr - ods包层管的性能

IF 2.8 2区 工程技术 Q3 MATERIALS SCIENCE, MULTIDISCIPLINARY
Takafumi Narukawa , Keietsu Kondo , Yuki Fujimura , Kazuo Kakiuchi , Yutaka Udagawa , Yoshiyuki Nemoto
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引用次数: 3

摘要

为了评估在轻水堆(LWRs)冷却剂损失(LOCA)条件下氧化物弥散强化FeCrAl (FeCrAl- ods)包层管的行为,进行了以下两个实验室规模的LOCA模拟试验:爆炸和整体热冲击试验。在未辐照、应力消除的FeCrAl-ODS包层管样品上进行了四次爆炸和三次整体热冲击试验,模拟了LOCA过程中假设的气球膨胀和破裂、氧化和淬火。熔覆管的破裂温度比锆合金熔覆管高200℃,熔覆管的最大周向应变小于或等于锆合金熔覆管。结果表明,与传统锆合金包层管相比,FeCrAl-ODS包层管具有更高的高温强度。在淬火过程中,等温氧化后,FeCrAl-ODS包层管在承受轴向约束载荷~(比现有轻水堆估计的轴向约束载荷高10倍以上)后没有断裂。在等温氧化条件下,FeCrAl-ODS包层管几乎没有被氧化。但在短时间氧化后熔化,在短时间氧化后断裂。根据这些结果,feral - ods包层管应该不会在LOCAs期间的预期时间范围内断裂,在LOCAs期间不会发生熔化或异常氧化。
本文章由计算机程序翻译,如有差异,请以英文原文为准。
Behavior of FeCrAl-ODS cladding tube under loss-of-coolant accident conditions

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.

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来源期刊
Journal of Nuclear Materials
Journal of Nuclear Materials 工程技术-材料科学:综合
CiteScore
5.70
自引率
25.80%
发文量
601
审稿时长
63 days
期刊介绍: The Journal of Nuclear Materials publishes high quality papers in materials research for nuclear applications, primarily fission reactors, fusion reactors, and similar environments including radiation areas of charged particle accelerators. Both original research and critical review papers covering experimental, theoretical, and computational aspects of either fundamental or applied nature are welcome. The breadth of the field is such that a wide range of processes and properties in the field of materials science and engineering is of interest to the readership, spanning atom-scale processes, microstructures, thermodynamics, mechanical properties, physical properties, and corrosion, for example. Topics covered by JNM Fission reactor materials, including fuels, cladding, core structures, pressure vessels, coolant interactions with materials, moderator and control components, fission product behavior. Materials aspects of the entire fuel cycle. Materials aspects of the actinides and their compounds. Performance of nuclear waste materials; materials aspects of the immobilization of wastes. Fusion reactor materials, including first walls, blankets, insulators and magnets. Neutron and charged particle radiation effects in materials, including defects, transmutations, microstructures, phase changes and macroscopic properties. Interaction of plasmas, ion beams, electron beams and electromagnetic radiation with materials relevant to nuclear systems.
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