Research Progress and Technological Development Trend of Accident Tolerant Fuel

Li Ziyi; Wang Xiaomin; Wang Kai; Zhang Ruiqian; Yin Chunyu; Chen Huan; Shi Haojiang; Pei Jingyuan; Lu Yonghong

doi:10.13832/j.jnpe.2024.05.0155

Volume 45 Issue 5

Oct. 2024

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Article Navigation > Nuclear Power Engineering > 2024 > 45(5): 155-164

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

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

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Research Progress and Technological Development Trend of Accident Tolerant Fuel

doi: 10.13832/j.jnpe.2024.05.0155

Nuclear Power Institute of China, Chengdu, 610213, China

Received Date: 2023-11-12
Rev Recd Date: 2023-12-27
Publish Date: 2024-10-14

Abstract

Abstract

The research and development of accident tolerant fuel (ATF) has become a new research direction in the international fuel industry in the post-Fukushima era, which involves the research and development of advanced cladding materials and new nuclear fuels. After more than ten years of comprehensive and systematic research, the international nuclear fuel industry represented by the U.S. and France have gained important progress, focusing further on medium and long-term technical solutions. In this paper, the important progress, challenges and development trend of subsequent technologies in ATF cladding materials (including Cr coating, FeCrAl alloy and SiC composite) and fuels (including enhanced UO₂, high uranium density fuel and ceramic dispersive matrix dispersion (CDM) fuel) at home and abroad are reviewed.
- ATF,
- Cr coated-zirconium alloy,
- Enhanced UO₂ pellet,
- CDM fuel

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References(73)

References

[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 eVinci^TM 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 ADOPT^TM 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/.