实验技术与管理

[1]高立本，沈健.高温气冷堆的发展与前景[J].中国核工业，2016(10):24–27.GAO L B, SHEN J. The development and perspective of HTGR[J]. China Nuclear Industry, 2016(10):24–27.(in Chinese)

[2]吴宗鑫.我国高温气冷堆的发展[J].核动力工程，2000,21(1):39–43, 80.WU Z X. The development of high temperature gas-cooled reactor in China[J]. Nuclear Power Engineering, 2000, 21(1):39–43, 80.(in Chinese)

[3]张作义，吴宗鑫，王大中，等.我国高温气冷堆发展战略研究[J].中国工程科学，2019, 21(1):12–19.ZHANG Z Y, WU Z X, WANG D Z, et al. Development strategy of high temperature gas cooled reactor in China[J]. Strategic Study of CAE, 2019, 21(1):12–19.(in Chinese)

[4]徐元辉，左开芬.高温气冷堆的过去、现在和将来[J].高技术通讯，1994(11):39–43.XU Y H, ZUO K F. The past, today and the future of HTGR[J].High Technology Letters, 1994(11):39–43.(in Chinese)

[5]吴宗鑫，张作义.世界核电发展趋势与高温气冷堆[J].核科学与工程，2000, 20(3):211–219, 231.WU Z X, ZHANG Z Y. World development of nuclear power system and high temperature gas-cooled reactor[J]. Chinese Journal of Nuclear Science and Engineering, 2000, 20(3):211–219, 231.(in Chinese)

[6] ZHANG Z Y, DONG Y J, LI F, et al. The Shandong Shidao Bay200 MWe high-temperature gas-cooled reactor pebble-bed module(HTR-PM)demonstration power plant:An engineering and technological innovation[J]. Engineering, 2016, 2(1):112–118.

[7]李晓伟，吴莘馨，张作义，等.高温气冷堆示范工程螺旋管式直流蒸汽发生器工程验证试验[J].清华大学学报（自然科学版），2021, 61(4):329–337.LI X W, WU X X, ZHANG Z Y, et al. Engineering test of HTR-PM helical tube once through steam generator[J]. Journal of Tsinghua University(Science and Technology), 2021, 61(4):329–337.(in Chinese)

[8]李顺洋，万力，桂南，等.基于弹塑性接触和渗流模型的静密封泄漏计算[J].清华大学学报（自然科学版），2023, 63(8):1264–1272.LI S Y, WAN L, GUI N, et al. Evaluation of leakage rates of static seals based on elastic-plastic contact theories and seepage theories[J]. Journal of Tsinghua University(Science and Technology),2023, 63(8):1264–1272.(in Chinese)

[9]任成，杨星团，李聪新，等.高温堆球床等效导热系数测量实验加热系统设计及可行性验证[J].清华大学学报（自然科学版），2014, 54(8):1068–1072.REN C, YANG X T, LI C X, et al. Heating system design and validation for the pebble bed effective thermal conductivity experiment in a high temperature gas-cooled reactor[J]. Journal of Tsinghua University(Science and Technology), 2014, 54(8):1068–1072.(in Chinese)

[10]雒晓卫，于溯源，张振声，等. HTR-10产生石墨粉尘量的估算及其尺寸分布[J].核动力工程, 2005, 26(2):203–208.LUO X W, YU S Y, ZHANG Z S, et al. Estimation of graphite dust quantity and size distribution of graphite particle in HTR-10[J]. Nuclear Power Engineering, 2005, 26(2):203–208.(in Chinese)

[11]陈志鹏，雒晓卫，于溯源.输送速度对高温气冷堆燃料装卸系统提升段石墨球磨损性能的影响[J].原子能科学技术，2012, 46（增刊2）:853–858.CHEN Z P, LUO X W, YU S Y. Effect of feeding velocity on wear behavior of graphite ball under elevating process with HTGR fuel handling system[J]. Atomic Energy Science and Technology, 2012, 46(S2):853–858.(in Chinese)

[12] ARANGANATHAN N, BIJWE J. Comparative performance evaluation of NAO friction materials containing natural graphite and thermo-graphite[J]. Wear, 2016, 358–359:17–22.

[13] GRATTAN P A, LANCASTER J K. Abrasion by lamellar solid lubricants[J]. Wear, 1967, 10(6):453–468.

[14]朱振国，王硕，任勇，等.机械密封用炭石墨材料摩擦磨损性能研究[J].炭素技术，2017, 36(6):34–37.ZHU Z G, WANG S, REN Y, et al. Friction and wear behavior of carbon graphite materials in mechanical seal[J]. Carbon Techniques,2017, 36(6):34–37.(in Chinese)

[15]徐宇轩.机械用炭石墨材料的结构设计与力学、摩擦学性能研究[D].长沙：湖南大学，2022.XU Y X. Structural design and mechanical and tribological properties of carbon-graphite materials for machinery[D].Changsha:Hunan University, 2022.(in Chinese)

[16]雒晓卫，于溯源，盛选禹，等.三种反应堆用石墨摩擦性能比较[J].高技术通讯，2001, 11(5):97–100.LUO X W, YU S Y, SHENG X Y, et al. Friction performance of three kinds of graphite used in nuclear reactors[J]. High Technology Letters, 2001,11(5):97–100.(in Chinese)

[17]全国钢标准化技术委员会（SAC/TC 183）.高温气冷堆堆内构件用核级等静压石墨：GB/T 40408—2021[S].北京：中国标准出版社，2021.National Steel Standardization Technical Committee(SAC/TC183). Nuclear grade isostatic graphite for high temperature gas-cooled reactor internals:GB/T 40408—2021[S]. Beijing:China Standards Publishing House, 2021.(in Chinese)

[18]李馨楠，周湘文，卢振明，等.球形燃料元件基体石墨摩擦磨损性能的研究[J].炭素技术，2016, 35(1):1–6.LI X N, ZHOU X W, LU Z M, et al, Friction and wear properties of matrix graphite in spherical fuel element for HTR[J]. Carbon Techniques, 2016, 35(1):1–6.(in Chinese)

[19] STANSFIELD O M. Friction and wear of graphite in dry helium at 25, 400, and 800℃[J]. Nuclear Applications, 1969, 6(4):313–320.