HT9 STRAIN MODELING FOR FUEL PIN DEFORMATION

被引:0
|
作者
Hackett, Micah [1 ]
Latta, Ryan [1 ]
Miller, Sam [1 ]
Xu, Cheng [1 ]
机构
[1] TerraPower LLC, Bellevue, WA 98005 USA
关键词
TYPE-316; STAINLESS-STEEL; IRRADIATION CREEP; FUSION HEATS; STRESS; TEMPERATURE; 9CR-1MO; ALLOYS;
D O I
暂无
中图分类号
TH [机械、仪表工业];
学科分类号
0802 ;
摘要
The Terra Power Traveling Wave Reactor (TWR) is a sodium-cooled fast reactor design that utilizes a high-burnup metallic uranium fuel cycle. The fuel system depends on a cladding material with demonstrated swelling resistance to high doses as well as adequate thermal creep strength. HT9 steel is a leading cladding candidate for the first TWR, having demonstrated excellent swelling and strain performance to doses > 200 dpa. A strain model was developed as a design tool to predict fuel pin deformation as a function of irradiation dose, stress, and temperature. The sources of strain deformation will be described along with the uncertainties in utilizing existing data to build a mechanistic model. The strain model is then incorporated into a fuel performance code to provide new insight in deformation behavior of HT9 fuel pins.
引用
收藏
页数:6
相关论文
共 50 条
  • [11] Thermomechanical Processing for Improved Mechanical Properties of HT9 Steels
    Byun, Thak Sang
    Collins, David A.
    Lach, Timothy G.
    Choi, Jung Pyung
    Maloy, Stuart A.
    MATERIALS, 2024, 17 (15)
  • [12] Microchemical evolution of irradiated additive-manufactured HT9
    Xiu, Pengyuan
    Massey, Caleb P.
    Green, T. M. Kelsy
    Taller, Stephen
    Isheim, Dieter
    Sridharan, Niyanth
    Field, Kevin G.
    JOURNAL OF NUCLEAR MATERIALS, 2022, 559
  • [13] MICROSTRUCTURAL EVOLUTION OF SELF-ION IRRADIATED HT9
    Beckett, Elizabeth
    Hackett, Micah
    Jiao, Zhijie
    Sun, Kai
    Was, Gary S.
    PROCEEDINGS OF THE 21ST INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING - 2013, VOL 1, 2014,
  • [14] Transmission electron microscopy characterization of the fuel-cladding chemical interactions in HT9 cladded U-10Zr fuel
    Wang, Yachun
    Miller, Brandon D.
    Harp, Jason M.
    Salvato, Daniele
    Capriotti, Luca
    Yao, Tiankai
    JOURNAL OF NUCLEAR MATERIALS, 2022, 572
  • [15] Transmission electron microscopy characterization of the fuel-cladding chemical interactions in HT9 cladded U-10Zr fuel
    Wang, Yachun
    Miller, Brandon D.
    Harp, Jason M.
    Salvato, Daniele
    Capriotti, Luca
    Yao, Tiankai
    JOURNAL OF NUCLEAR MATERIALS, 2022, 572
  • [16] Irradiation dose and temperature dependence of fracture toughness in high dose HT9 steel from the fuel duct of FFTF
    Byun, Thak Sang
    Toloczko, Mychailo B.
    Saleh, Tarik A.
    Maloy, Stuart A.
    JOURNAL OF NUCLEAR MATERIALS, 2013, 432 (1-3) : 1 - 8
  • [17] Thermal creep analysis and correlation development for manufactured HT9 cladding
    Kim, Dong-Ha
    Lee, Cheol Min
    Kim, Jun-Hwan
    Kim, Sung -Ho
    Yeo, Sunghwan
    Lee, Yong -Kook
    JOURNAL OF NUCLEAR MATERIALS, 2024, 593
  • [18] Mechanism-based modeling of thermal and irradiation creep behavior: An application to ferritic/martensitic HT9 steel
    Wen, W.
    Kohnert, A.
    Kumar, M. Arul
    Capolungo, L.
    Tome, C. N.
    INTERNATIONAL JOURNAL OF PLASTICITY, 2020, 126
  • [19] HT9钢的高温蠕变性能研究
    刘桂良
    张毅勇
    张程煜
    赵蒙蒙
    王辉
    唐睿
    钢铁研究学报, 2019, 31 (01) : 72 - 81
  • [20] Corrosion of ferritic-martensitic steel HT9 in supercritical water
    Ren, X.
    Sridharan, K.
    Allen, T. R.
    JOURNAL OF NUCLEAR MATERIALS, 2006, 358 (2-3) : 227 - 234