高级检索

留言板

尊敬的读者、作者、审稿人, 关于本刊的投稿、审稿、编辑和出版的任何问题, 您可以本页添加留言。我们将尽快给您答复。谢谢您的支持!

姓名
邮箱
手机号码
标题
留言内容
验证码

耐事故燃料研发进展及技术发展趋势

李紫祎 王晓敏 王凯 张瑞谦 尹春雨 陈寰 石浩江 裴静远 陆永洪

李紫祎, 王晓敏, 王凯, 张瑞谦, 尹春雨, 陈寰, 石浩江, 裴静远, 陆永洪. 耐事故燃料研发进展及技术发展趋势[J]. 核动力工程, 2024, 45(5): 155-164. doi: 10.13832/j.jnpe.2024.05.0155
引用本文: 李紫祎, 王晓敏, 王凯, 张瑞谦, 尹春雨, 陈寰, 石浩江, 裴静远, 陆永洪. 耐事故燃料研发进展及技术发展趋势[J]. 核动力工程, 2024, 45(5): 155-164. doi: 10.13832/j.jnpe.2024.05.0155
Li Ziyi, Wang Xiaomin, Wang Kai, Zhang Ruiqian, Yin Chunyu, Chen Huan, Shi Haojiang, Pei Jingyuan, Lu Yonghong. Research Progress and Technological Development Trend of Accident Tolerant Fuel[J]. Nuclear Power Engineering, 2024, 45(5): 155-164. doi: 10.13832/j.jnpe.2024.05.0155
Citation: Li Ziyi, Wang Xiaomin, Wang Kai, Zhang Ruiqian, Yin Chunyu, Chen Huan, Shi Haojiang, Pei Jingyuan, Lu Yonghong. Research Progress and Technological Development Trend of Accident Tolerant Fuel[J]. Nuclear Power Engineering, 2024, 45(5): 155-164. doi: 10.13832/j.jnpe.2024.05.0155

耐事故燃料研发进展及技术发展趋势

doi: 10.13832/j.jnpe.2024.05.0155
详细信息
    作者简介:

    李紫祎(1997—),女,助理研究员,现从事Cr涂层锆合金包壳性能评价研究,E-mail: zoey0209@126.com

    通讯作者:

    王晓敏,E-mail: npicwxm@163.com

  • 中图分类号: TL352

Research Progress and Technological Development Trend of Accident Tolerant Fuel

  • 摘要: 耐事故燃料(ATF)研发已成为后福岛时代国际燃料界一个新的研究方向,其内容涉及先进包壳材料、新型燃料的研发。经过十余年的全面系统研究,以美国、法国为代表的国际核燃料界在近期解决技术方案上取得了重要进展,对中远期的技术方向也更加聚焦。本文主要综述了国内外在ATF包壳材料(包括Cr涂层、FeCrAl合金与SiC复合材料)、燃料方面[包括增强型UO2、高铀密度燃料和陶瓷基包覆颗粒弥散(CDM)燃料]研究取得的重要进展、面临的挑战及后续技术发展趋势。

     

  • 图  1  ATF主要技术研究方向

    CDM燃料—陶瓷基包覆颗粒弥散燃料

    Figure  1.  The Main Research Directions of ATF

    图  2  法马通公司制备的ATR辐照用SiC小棒与西屋公司制备的双层SiC包壳管

    Figure  2.  ATR SiC Irradiation Rods Made by Framatome and SiC Double-layered Cladding Tube Made by Westinghouse

    图  3  CDM燃料结构示意图

    Figure  3.  Structure Diagram of CDM Fuel

    表  1  Cr涂层包壳材料堆外性能试验与评价

    Table  1.   Out-of-pile Performance Test and Evaluation of Cr-coated Cladding

    技术指标 NPIC 国外
    长度 全尺寸制备能力 全尺寸制备能力
    物相结构 涂层为物相单一的Cr金属 涂层为物相单一的Cr金属
    结合质量 不同变形量拉伸压扁下涂层目视无裂纹剥落 不同变形量拉伸压扁下涂层目视无裂纹剥落
    厚度 10~20 μm 5~80 μm
    耐高温氧化性能 比锆合金低1个数量级 比锆合金低1个数量级
    耐腐蚀性能 比锆合金低约2个数量级 比锆合金低约2个数量级
    抗热冲击性能 1200℃涂层完好 暂无报道
    耐摩擦磨损性能 优于锆合金 优于锆合金
    力学拉伸性能 室温抗拉强度约500 MPa 室温抗拉强度约500 MPa
    抗内压爆破性能 44.4 MPa(400℃) 35 MPa(940℃,80 μm)
    耐蠕变性能 径向应变0.69%(400℃, 130 MPa) 径向应变0.57%~0.7% (385℃, 107.5 MPa)
    下载: 导出CSV

    表  2  FeCrAl包壳材料堆外性能试验与评价

    Table  2.   Out-of-pile Performance Test and Evaluation of FeCrAl Cladding

    技术指标 NPIC 国外
    包壳管尺寸 长 4 m、直径9.5 mm、壁厚0.37~0.40 mm的管材 长4 m、直径9.5 mm、壁厚<0.40 mm的管材
    拉伸性能(室温) 屈服强度约720 MPa 屈服强度约600 MPa
    抗拉强度约880 MPa 抗拉强度约800 MPa
    耐高温水蒸气氧化性能 平衡常数(Kp)值量级为10−11~10−13,比锆合金低4个数量级 Kp值量级为10−11~10−13,比锆合金低4个数量级
    耐均匀腐蚀性能 300 d腐蚀后的增重为5~7 mg/dm2,比锆合金低1个数量级 300 d腐蚀后的增重为1~5 mg/dm2,比锆合金低1个数量级
    蠕变性能 蠕变指数16.76 蠕变指数10~20
    辐照考验 开展了板材、管材离子辐照、中子辐照(研究堆) 开展了板材离子辐照、中子辐照(商用沸水堆)
    下载: 导出CSV

    表  3  SiC复合包壳堆外性能试验与评价

    Table  3.   Out-of-pile Performance Test and Evaluation of SiC Composite Cladding

    技术指标 NPIC 国外
    包壳管长度/m 1.5 4
    密度/(g·cm−3) >2.8 >2.7
    开孔孔隙率/% 约10 约10
    壁厚/mm 0.8与1 0.7~2.1
    轴拉强度/MPa 240~270 230~270
    环拉
    强度/MPa
    240 200~340
    气密性/
    (Pa·m3·s−1)
    约10−10 约10−3(带端塞)
    约10−10(热循环后)
    腐蚀失重/
    (mg·dm−2·d−1)
    0.23(含氧水环境) 0.02(电站水环境)
    热导率/
    (W·m−1·K−1)
    11 8.5~13
    热膨胀系数/
    (10−6·K−1)
    3.8~4.6(室温~1300℃) 2.5~6.5(300~1400℃)
    下载: 导出CSV

    表  4  几种燃料热物理性能对比

    Table  4.   Physical Performance Comparison of Nuclear Fuels

    物理性能 UO2 U3Si2 UN
    密度/(g·cm−3) 10.96 12.2 14.3
    铀密度/[g(U)·cm−3] 9.66 13.52 11.31
    热导率(673~1473 K)/(W·m−1·K−1) 6~2.5 13.0~22.3 15~27.3
    熔点/K 3130 1938 2953
    下载: 导出CSV

    表  5  UN-CDM燃料芯块部分堆外性能对比

    Table  5.   Out-of-pile Performance Comparison of UN-CDM Fuel Pallet between NPIC and Abroad

    物项 NPIC 国外
    UN微球 碳铀比≥2.6
    铀含量≥94.2%
    密度:90%T.D.
    碳铀比≥2.6(ORNL)
    铀含量≥94%
    密度:92%T.D.
    TRISO SiC涂层 密度:约3.0 g/cm3
    热膨胀系数:约5×10−6−1
    密度:约3.2 g/cm3(ORNL,法马通公司)
    热膨胀系数:约5.11×10−6−1
    CDM芯块 相体积分数:约38%
    基体密度:95%T.D.
    相体积分数:约45%(ORNL,法马通公司)
    基体密度:97%T.D.
    下载: 导出CSV

    表  6  ATF及包壳材料的主要特点及发展阶段

    Table  6.   Major Pros and Cons of ATF and Claddings and Corresponding Development Stage

    物项 优势 劣势 技术路线发展阶段
    包壳 Cr涂层锆合金  锆合金包壳体系成熟,商业化应用速度快  仍有锆水反应风险 近期
    FeCrAl合金  导热性好,热膨胀系数小,抗氧化、耐腐蚀  中子吸收截面较大,塑性加工困难,成型难度大 中远期
    SiC复合材料  基础性能优异,高温下蒸汽氧化速率低、强度高、化学性能稳定,极端工况下安全系数高  耐水热腐蚀性能不佳 中远期
    燃料 增强型UO2  密度更合理、晶粒尺寸更大,提升燃料芯块导热性能,减弱PCI  铀密度提升存在上限 近期
    高铀密度燃料  铀密度高、导热性能好  铀硅化物熔点偏低,UN燃料耐水腐蚀性差 中远期
    CDM燃料  具备超强的裂变气体包容能力,抗辐照性能优异  铀装量低 中远期
    下载: 导出CSV
  • [1] 郑中成,郭立平,唐睿. 超临界水冷堆燃料包壳材料的辐照损伤研究进展[J]. 原子核物理评论,2017, 34(2): 211-218.
    [2] CARMACK J, GOLDNER F, BRAGG-SITTON S M, et al. Overview of the U. S. DOE accident tolerant fuel development program: INL/CON-13-29288[R]. Idaho Falls: Idaho National Lab. , 2013.
    [3] TERRANI K A. Accident tolerant fuel cladding development: promise, status, and challenges[J]. Journal of Nuclear Materials, 2018, 501: 13-30. doi: 10.1016/j.jnucmat.2017.12.043
    [4] 杨红艳,陈寰,张瑞谦,等. 核电耐事故锆包壳表面涂层研究进展[J]. 表面技术,2022, 51(7): 87-97.
    [5] BISCHOFF J, DELAFOY C, VAUGLIN C, et al. AREVA NP’s enhanced accident-tolerant fuel developments: focus on Cr-coated M5 cladding[J]. Nuclear Engineering and Technology, 2018, 50(2): 223-228. doi: 10.1016/j.net.2017.12.004
    [6] LYONS J L, PARTEZANA J, BYERS W A, et al. Westinghouse chromium-coated zirconium alloy cladding development and testing[C]. Proceedings of Topfuel 2019, Seattle, United States of America: 2019. https://www.ans.org/pubs/proceedings/article-47064/.
    [7] STUCKERT J, AUSTREGESILO H, HOLLANDS T H, et al. IAEA fumac benchmark on KIT bundle test cora-15[C]. Proceedings of Topfuel 2018, Prague, Czech Republic: 2018.
    [8] 严俊,廖业宏,彭振驯,等. Cr涂层锆合金事故容错燃料包壳材料研究进展[J]. 表面技术,2023, 52(12): 206-224.
    [9] 中国广核集团. 中广核开始国内ATF燃料入堆辐照测试工作[EB/OL]. (2019-01-22)[2023-12-21]. http://www.cgnpc.com.cn/cgn/c100944/2019-01/22/content_916216b8da314c03b1785b9d798c163d.shtml.
    [10] WEI T G, ZHANG R Q, YANG H Y, et al. Microstructure, corrosion resistance and oxidation behavior of Cr-coatings on Zircaloy-4 prepared by vacuum arc plasma deposition[J]. Corrosion Science, 2019, 158: 108077. doi: 10.1016/j.corsci.2019.06.029
    [11] KIM H G, KIM I H, JUNG Y I, et al. Adhesion property and high-temperature oxidation behavior of Cr-coated Zircaloy-4 cladding tube prepared by 3D laser coating[J]. Journal of Nuclear Materials, 2015, 465: 531-539. doi: 10.1016/j.jnucmat.2015.06.030
    [12] YEOM H, DABNEY T, JOHNSON G, et al. Improving deposition efficiency in cold spraying chromium coatings by powder annealing[J]. The International Journal of Advanced Manufacturing Technology, 2019, 100(5-8): 1373-1382. doi: 10.1007/s00170-018-2784-1
    [13] DELAFOY C, BISCHOFF J, LAROCQUE J, et al. Benefits of framatome’s E-ATF evolutionary solution: Cr-coated cladding with Cr2O3-doped fuel[C]. Proceedings of Topfuel 2018, Prague, Czech Republic: 2018.
    [14] FIELD K G, HU X X, LITTRELL K C, et al. Radiation tolerance of neutron-irradiated model Fe-Cr-Al alloys[J]. Journal of Nuclear Materials, 2015, 465: 746-755. doi: 10.1016/j.jnucmat.2015.06.023
    [15] AYDOGAN E, WEAVER J S, MALOY S A, et al. Microstructure and mechanical properties of FeCrAl alloys under heavy ion irradiations[J]. Journal of Nuclear Materials, 2018, 503: 250-262. doi: 10.1016/j.jnucmat.2018.03.002
    [16] GAMBLE K A, HALES J D, BARANI T, et al. Behavior of U3Si2 fuel and FeCrAl cladding under normal operating and accident reactor conditions: INL/EXT-16-40059[R]. Idaho Falls: Idaho National Laboratory, 2016.
    [17] LIN Y P, FAWCETT R M, DESILVA S S, et al. Path towards industrialization of enhanced accident tolerant fuel[C]. Proceedings of Topfuel 2018, Prague, Czech Republic: 2018.
    [18] TERRANI K A, PINT B A, KIM Y J, et al. Uniform corrosion of FeCrAl alloys in LWR coolant environments[J]. Journal of Nuclear Materials, 2016, 479: 36-47. doi: 10.1016/j.jnucmat.2016.06.047
    [19] SWARTZ M M, BYERS W A, LOJEK J, et al. Westinghouse eVinciTM heat pipe micro reactor technology development[C]. Proceedings of the 28th International Conference on Nuclear Engineering. New York, United States of America, ASME, 2021.
    [20] ZHANG J S. A review of steel corrosion by liquid lead and lead–bismuth[J]. Corrosion Science, 2009, 51(6): 1207-1227. doi: 10.1016/j.corsci.2009.03.013
    [21] POPOVIC M P, CHEN K, SHEN H, et al. A study of deformation and strain induced in bulk by the oxide layers formation on a Fe-Cr-Al alloy in high-temperature liquid Pb-Bi eutectic[J]. Acta Materialia, 2018, 151: 301-309. doi: 10.1016/j.actamat.2018.03.041
    [22] YAMAMOTO Y, FIELD K, SNEAD L. Optimization of nuclear grade FeCrAl fuel cladding for light water reactors[C]. Michigan, United States of America: Proceedings of Oak Ridge National Laboratory IAEA Technical Meeting, 2015.
    [23] TERRANI K A, PARISH C M, SHIN D, et al. Protection of zirconium by alumina- and chromia-forming iron alloys under high-temperature steam exposure[J]. Journal of Nuclear Materials, 2013, 438(1-3): 64-71. doi: 10.1016/j.jnucmat.2013.03.006
    [24] TERRANI K A, ZINKLE S J, SNEAD L L. Advanced oxidation-resistant iron-based alloys for LWR fuel cladding[J]. Journal of Nuclear Materials, 2014, 448(1-3): 420-435. doi: 10.1016/j.jnucmat.2013.06.041
    [25] YAMAMOTO Y, YANG Y, FIELD K G, et al. Letter report documenting progress of second generation ATF FeCrAl alloy fabrication: ORNL/LTR-2014/219[R]. Oak Ridge: Oak Ridge National Lab. , 2014.
    [26] BACHHAV M, ROBERT ODETTE G, MARQUIS E A. Microstructural changes in a neutron-irradiated Fe–15at. %Cr alloy[J]. Journal of Nuclear Materials, 2014, 454(1-3): 381-386. doi: 10.1016/j.jnucmat.2014.08.026
    [27] REBAK R B. Accident-tolerant materials for light water reactor fuels[M]. Amsterdam: Elsevier, 2020:143-156.
    [28] DECK C P, JACOBSEN G M, SHEEDER J, et al. Characterization of SiC–SiC composites for accident tolerant fuel cladding[J]. Journal of Nuclear Materials, 2015, 466: 667-681. doi: 10.1016/j.jnucmat.2015.08.020
    [29] 伍浩松,郭志锋. 法马通耐事故燃料在美国试验堆中进行辐照试验[J]. 国外核新闻,2018(7): 23.
    [30] QIU B W, WANG J, DENG Y B, et al. A review on thermohydraulic and mechanical-physical properties of SiC, FeCrAl and Ti3SiC2 for ATF cladding[J]. Nuclear Engineering and Technology, 2020, 52(1): 1-13. doi: 10.1016/j.net.2019.07.030
    [31] XU P, LAHODA E J, LYONS J, et al. Status update on Westinghouse sic composite cladding fuel development[C]. Prague, Czech Republic: Proceedings of Topfuel 2018, 2018.
    [32] GERINGER J W, PETRIE C, JAMES A, et al. HFIR SiC-SiC composite clad tube bowing test: pre-irradiation characterization: ORNL/SPR-2021/2100[R]. Springfield: National Technical Information Service, 2021.
    [33] World Nuclear News, General Atomics, Framatome join for fuel channel work[EB/OL]. (2020-02-21)[2023-12-26]. https://www.world-nuclear-news.org/Articles/GA-Framatome-team-up-on-fuel-channel-development.
    [34] KOYANAGI T, KATOH Y, NOZAWA T. Design and strategy for next-generation silicon carbide composites for nuclear energy[J]. Journal of Nuclear Materials, 2020, 540: 152375. doi: 10.1016/j.jnucmat.2020.152375
    [35] FIELD K G, SNEAD M A, YAMAMOTO Y, et al. Handbook on the material properties of FeCrAl alloys for nuclear power production applications (FY18 Version: Revision 1): ORNL/SPR-2018/905[R]. Oak Ridge: Oak Ridge National Laboratory, 2018.
    [36] DOYLE P, SUN K C, SNEAD L, et al. The effects of neutron and ionizing irradiation on the aqueous corrosion of SiC[J]. Journal of Nuclear Materials, 2020, 536: 152190. doi: 10.1016/j.jnucmat.2020.152190
    [37] 尚新渊,张爱民. 碳化硅复合材料包壳燃料棒在LOCA事故中的特性研究[J]. 核技术,2019, 42(8): 080601.
    [38] COZZO C, RAHMAN S. SiC cladding thermal conductivity requirements for normal operation and LOCA conditions[J]. Progress in Nuclear Energy, 2018, 106: 278-283. doi: 10.1016/j.pnucene.2018.03.016
    [39] LI M, ZHOU X B, YANG H, et al. The critical issues of SiC materials for future nuclear systems[J]. Scripta Materialia, 2018, 143: 149-153. doi: 10.1016/j.scriptamat.2017.03.001
    [40] 刘俊凯,张新虎,恽迪. 事故容错燃料包壳候选材料的研究现状及展望[J]. 材料导报,2018, 32(11): 1757-1778.
    [41] SHARMA A S, FITRIANI P, YOON D H. Fabrication of SiCf/SiC and integrated assemblies for nuclear reactor applications[J]. Ceramics International, 2017, 43(18): 17211-17215. doi: 10.1016/j.ceramint.2017.09.126
    [42] KATOH Y, OZAWA K, SHIH C, et al. Continuous SiC fiber, CVI SiC matrix composites for nuclear applications: properties and irradiation effects[J]. Journal of Nuclear Materials, 2014, 448(1-3): 448-476. doi: 10.1016/j.jnucmat.2013.06.040
    [43] ENS. Framatome’s GAIA EATF technology completes its first-ever fuel cycle[EB/OL]. (2021-02-04)[2023-12-25]. https://www.euronuclear.org/news/framatome-gaia-eatf-nuclear-fuel/.
    [44] Framatome. Framatome’s GAIA Enhanced Accident Tolerant Fuel completes first-ever fuel cycle[EB/OL]. (2021-02-02)[2023-12-25]. https://www.framatome.com/medias/framatomes-gaia-enhanced-accident-tolerant-fuel-completes-first-ever-fuel-cycle/?lang=en.
    [45] 伍浩松,孟雨晨. 法马通先进燃料代码获得美核管会批准[J]. 国外核新闻,2023(5): 18.
    [46] 伍浩松,杨鹏. 法马通耐事故燃料组件在美机组完成首个换料周期辐照测试[J]. 国外核新闻,2023(9): 17.
    [47] Nuclear Newswire. Framatome receives NRC approval for transport of LEU+ fuel assemblies[EB/OL]. (2022-02-23)[2023-12-25]. https://www.ans.org/news/article-3690/framatome-receives-nrc-approval-for-transport-of-leu-fuel-assemblies/.
    [48] Westinghouse. Westinghouse EnCore® fuel[EB/OL]. (2017-06-12)[2023-12-25]. https://info.westinghousenuclear.com/blog/westinghouse-encore-fuel.
    [49] Westinghouse. Accident tolerant fuel: westinghouse ADOPTTM fuel achieves regulatory approval, moves closer to U. S. deployment[EB/OL]. (2023-03-14)[2023-12-25]. https://info.westinghousenuclear.com/news/adopt-nrc-approval.
    [50] JOHNSON K D, RAFTERY A M, LOPES D A, et al. Fabrication and microstructural analysis of UN-U3Si2 composites for accident tolerant fuel applications[J]. Journal of Nuclear Materials, 2016, 477: 18-23. doi: 10.1016/j.jnucmat.2016.05.004
    [51] LOPES D A, UYGUR S, JOHNSON K. Degradation of UN and UN–U3Si2 pellets in steam environment[J]. Journal of Nuclear Science and Technology, 2017, 54(4): 405-413. doi: 10.1080/00223131.2016.1274689
    [52] HARP J M, LESSING P A, HOGGAN R E. Uranium silicide pellet fabrication by powder metallurgy for accident tolerant fuel evaluation and irradiation[J]. Journal of Nuclear Materials, 2015, 466: 728-738. doi: 10.1016/j.jnucmat.2015.06.027
    [53] HOGGAN R E, TOLMAN K R, CAPPIA F, et al. Grain size and phase purity characterization of U3Si2 fuel pellets[J]. Journal of Nuclear Materials, 2018, 512: 199-213. doi: 10.1016/j.jnucmat.2018.10.011
    [54] LAHODA E J, AVALI R, BOYLAN F A. Program management plan EnCore® accident tolerant fuel (ATF) program: GATFT-LTR-19-013[R]. Cranberry Township: Westinghouse Electric Company LLC, 2019.
    [55] NRC. Longer term accident tolerant fuel technologies[EB/OL]. (2024-06-13)[2024-08-27]. https://www.nrc.gov/reactors/power/atf/technologies/longer-term.html.
    [56] MISHCHENKO Y, JOHNSON K D, JÄDERNÄS D, et al. Uranium nitride advanced fuel: an evaluation of the oxidation resistance of coated and doped grains[J]. Journal of Nuclear Materials, 2021, 556: 153249. doi: 10.1016/j.jnucmat.2021.153249
    [57] HE L F, KHAFIZOV M, JIANG C, et al. Phase and defect evolution in uranium-nitrogen-oxygen system under irradiation[J]. Acta Materialia, 2021, 208: 116778. doi: 10.1016/j.actamat.2021.116778
    [58] GONG B W, KARDOULAKI E, YANG K, et al. UN and U3Si2 composites densified by spark plasma sintering for accident-tolerant fuels[J]. Ceramics International, 2022, 48(8): 10762-10769. doi: 10.1016/j.ceramint.2021.12.292
    [59] HANSON W A, CAPPIA F, WHITE J T, et al. Post-irradiation examination of low burnup U3Si5 and UN-U3Si5 composite fuels[J]. Journal of Nuclear Materials, 2023, 578: 154346. doi: 10.1016/j.jnucmat.2023.154346
    [60] YANG K, KARDOULAKI E, ZHAO D, et al. Uranium nitride (UN) pellets with controllable microstructure and phase – fabrication by spark plasma sintering and their thermal-mechanical and oxidation properties[J]. Journal of Nuclear Materials, 2021, 557: 153272. doi: 10.1016/j.jnucmat.2021.153272
    [61] TERRANI K A, TRAMMELL M P, KIGGANS J O, et al. UN TRISO compaction in SiC for FCM fuel irradiations: ORNL/LTR-2016/702[R]. Oak Ridge: Oak Ridge National Lab. , 2016.
    [62] LEE H G, KIM D, LEE S J, et al. Thermal conductivity analysis of SiC ceramics and fully ceramic microencapsulated fuel composites[J]. Nuclear Engineering and Design, 2017, 311: 9-15. doi: 10.1016/j.nucengdes.2016.11.005
    [63] 葛维维,杨金凤. 核安全峰会·聚焦微堆低浓化[J]. 中国核工业,2016(4): 22.
    [64] 王兴春,张焰. 荷兰即将启动微堆MMR燃料辐照测试[J]. 国外核新闻,2021(7): 19.
    [65] 伍浩松,李晨曦. 美FCM燃料中试制造设施正式投运[J]. 国外核新闻,2022(9): 19.
    [66] 伍浩松,戴定. 加成功制造首批TRISO燃料[J]. 国外核新闻,2021(5): 19.
    [67] Nuclear Newswire. Framatome and USNC team up to produce TRISO fuel at framatome facility[EB/OL]. (2023-11-29)[2023-12-25]. https://www.ans.org/news/article-5571/framatome-and-usnc-team-up-to-produce-triso-fuel-at-framatome-facility/.
    [68] LINDEMER T B, VOIT S L, SILVA C M, et al. Carbothermic synthesis of 820 μm uranium nitride kernels: literature review, thermodynamics, analysis, and related experiments[J]. Journal of Nuclear Materials, 2014, 448(1-3): 404-411. doi: 10.1016/j.jnucmat.2013.10.036
    [69] BROWN N R, HERNANDEZ R, NELSON A T. High volume packing fraction TRISO-based fuel in light water reactors[J]. Progress in Nuclear Energy, 2022, 146: 104151. doi: 10.1016/j.pnucene.2022.104151
    [70] 李维杰. 压水堆应用高丰度低浓铀燃料铀浓缩关键环节研究建议[J]. 当代化工研究,2023(4): 130-132.
    [71] EATF. Framatome’s EATF program[EB/OL]. (2018-01-17)[2023-12-25]. https://nextevolutionfuel.com/framatome-eatf-program/.
    [72] Framatome, Framatome EATF[EB/OL]. (2023-07-31)[2023-12-25]. https://nextevolutionfuel.com/.
    [73] EATF. Fuel for tomorrow[EB/OL]. (2019-06-13)[2023-12-25]. https://nextevolutionfuel.com/2019/06/fuel-for-tomorrow/.
  • 加载中
图(3) / 表(6)
计量
  • 文章访问数:  53
  • HTML全文浏览量:  8
  • PDF下载量:  22
  • 被引次数: 0
出版历程
  • 收稿日期:  2023-11-12
  • 修回日期:  2023-12-27
  • 刊出日期:  2024-10-14

目录

    /

    返回文章
    返回