隨著核電廠運轉時間的增加,在輕水式反應器(Light Water Reactor, LWR)漸發現有應
力腐蝕龜裂(Intergranular Stress Corrosion Cracking, IGSCC)的問題。針對溶氫濃度對於
組件腐蝕防治的效益仍需更多研究的基礎討論。而組件的劣化不僅會影響運轉安全亦
所費不貲。我國核三廠兩部機組為壓水式反應器,在其結構組件中,主要使用的是沃斯田
系不鏽鋼與鎳基合金。
本實驗研究鎳基合金與沃斯田系不鏽鋼材料的應力腐蝕龜裂行為，探討沃斯田系
不鏽鋼316、鎳基合金 600 與 X750,組件材料在壓水式反應器的水化學環境下，調整不
同溶氫濃度（0、0.45、2.58 ppm），對於防治應力腐蝕劣化效益的測試評估。腐蝕行為
研究以 U-bend 應力腐蝕試驗進行。試片測試後採用 SEM 觀察材料的表面,觀察表面
產生 crack 的分布情形及其裂縫長度、數目等。
結果顯示，DH 0 ppm 下之316 不鏽鋼經敏化處理後腐蝕較為嚴重，表面有較常與
較多之裂縫，氫氣的注入後對於316 不鏽鋼與鎳基600 合金之裂縫總數變化有降低的作
用，但整體而言對X750 而言數目影響不明顯。

As the operation time of nuclear power plants increases, the phenomenon of Intergranular
Stress Corrosion Cracking (IGSCC) is readily found in Light Water Reactor (LWR). Further
research is needed for the benefits of dissolved hydrogen concentration on component
corrosion prevention. The deterioration of components not only affects the safety of operation
but also costs. In Taiwan, Maanshan Nuclear Power Plant contains two units of Pressurized
Water Reactors (PWRs). For the structural components, the stainless steel and nickel-based
alloys are mainly used.
This experiment aimed to investigate the stress corrosion cracking behavior of different
heat-treated nickel-based alloys and stainless-steel materials, and to discuss their corrosion
behavior under the simulated environment of primary water of a pressurized water reactor.
Corrosion behavior studies were conducted by using the U-bend stress corrosion test under
different dissolved hydrogen concentration (0, 0.45, 2.58 ppm). After the test piece was tested,
the surface of the material was observed by SEM, and the distribution of cracks on the surface
and the length and number of cracks were observed.
The results showed that the sensitized 316 stainless steel tested at DH 0 ppm suffered the
most severe corrosion attack, and there were more and longer cracks on it. After the injection
of hydrogen, the trend of the total number of cracks for 316 stainless steel and nickel-based
600 alloy were changed. However, the effect of hydrogen injection on X750 series was
relatively insignificant.

[1] T. Root, J. Price, K. Hall, S. H Schneider, C. Rosenzweig, and A. Pounds, “Fingerprints
of global warming on wild animals and plants,” Nature, vol. 421, pp. 57–60, 2003.
[2] M. Dyurgerov and M. F Meier, “Glaciers and the Changing Earth System: A 2004
Snapshot,” vol. 58, 2004.
[3] S. Taskaev and V. V Kanygin, Boron Neutron Capture Therapy. 2016.
[4] R. C. Wiens, “EVIDENCE FOR AN ANCIENT EARTH Radiometric Dating - A
Christian Perspective Radiometeric Dating Does Work ,” 2002.
[5] M. B. Toloczko, M. J. Olszta, Z. Zhai, and S. M. Bruemmer, “Stress Corrosion Crack
Initiation Measurements of Alloy 600 in PWR Primary Water,” 17th Int. Conf. Environ.
Degrad. Mater. Nucl. Power Sytems - Water React., pp. 1–20, 2015.
[6] R. Ghafouri-Azar and S. S. Ho, “Analysis of Corrosion Fatigue for the Deaerator Heater
Tanks in Nuclear Power Plants,” in American Society of Mechanical Engineers, Pressure
Vessels and Piping Division (Publication) PVP, 2008, vol. 3.
[7] P. Skeldon, P. M. Scott, and P. Hurst, “Environmentally assisted cracking of alloy X-
750 in simulated PWR coolant,” Corrosion, vol. 48, no. 7, pp. 553–569, 1992.
[8] G. Furutani, N. Nakajima, T. Konishi, and M. Kodama, “Stress corrosion cracking on
irradiated 316 stainless steel,” J. Nucl. Mater., vol. 288, no. 2–3, pp. 179–186, 2001.
[9] X. Zhong, S. C. Bali, and T. Shoji, “Effects of Dissolved Hydrogen on the
Environmentally Assisted Cracking of 316 Stainless Steel in Pwr Primary Water At 325
O C,” pp. 1–20, 2015.
[10] L. Marchetti, F. Martin, F. Datcharry, and J. Chêne, “Kinetics of hydrogen permeation
through a Ni-base alloy membrane exposed to primary medium of pressurized water
reactors,” Corros. Sci., vol. 144, no. May, pp. 1–12, 2018.
[11] T. Kim, K. J. Choi, S. C. Yoo, and J. H. Kim, “Effects of dissolved hydrogen on the
crack-initiation and oxidation behavior of nickel-based alloys in high-temperature water,”
Corros. Sci., vol. 106, pp. 260–270, 2016.
[12] J. A. Roberts, “Structural materials in nuclear power systems,” Springer Sci. Bus. Media,
2013.
[13] I. K. R.W. Staehle, Anatomy of Proactivity, in: B.L. Eyre, “No Title,” in Int. Sym. on
Research for Aging Management of Light Water Reactors and Its Future Trend (The
15th Anniversary of INSS), 2008, pp. 29–115.
[14] A. Martinez-Ubeda, I. Griffiths, M. Karunaratne, P. Flewitt, C. Younes, and T. Scott,
“Influence of nominal composition variation on phase evolution and creep life of Type
98
316H austenitic stainless steel components,” Procedia Struct. Integr., vol. 2, pp. 958–
965, 2016.
[15] M. Michael, STAINLESS STEELSFOR DESIGN ENGINEERS. .
[16] R. M. G.S. Was, H.H. Tischner, “The Influence of Thermal Treatment on the Chemistry
and Structure of Grain Boundaries in Inconel 600,” LatanisionMetallurgical Trans., vol.
A. 12, pp. 1397–1408, 1981.
[17] G.S. Was, “Grain Boundary Chemistry and Intergranular Fracture in Austenitic Nickel-
Base Alloys,” Mater. Sci. Forum, vol. 46, pp. 335–358., 1989.
[18] J. R. SCARBERRY, R. C., PEARMAN, S. C., & CRUM, “Precipitation Reactions in
Inconel Alloy 600 and Their Effect on Corrosion Behavior.,” Corrosion, vol. 32, no. 10,
pp. 401–406, 1976.
[19] A. K. Sinha and J. J. Moore, “Precipitation of M23C6 carbides in an aged Inconel X-
750,” Metallography, vol. 19, no. 1, pp. 87–98, 1986.
[20] W. Gwan, G. Yu, and J. Huang, “Study of Stress Corrosion Cracking of Alloy X-750
Components in Nuclear Power Reactor,” vol. 14, no. 4, pp. 9–16, 2000.
[21] R. Staehle, “Definition of precursors, incubation, slow growth and propagation of SCC,”
SCC Initiat. Work., vol. 2, no. 3, pp. 79–87, 2008.
[22] T. T. K Arioka, T Miyamoto, T Yamada, “Formation of crack embryos prior to crack
growth in high temperature water,” in 14th International Conference on Environmental
Degradation of Materials in Nuclear Power Systems Water Reactors, 2009, pp. 895–909.
[23] T. M. K Arioka, T Yamada, T Terachi, “Temperature, potential and sensitization effects
on intergranular crack growth and crack-tip appearance of cold worked 316,” in 13th
International Conference on Environmental Degradation of Materials in Nuclear Power
Systems, 2007, pp. 1–13.
[24] S. F. H. Xu, “Laboratory Investigation of PWSCC of CRDM Nozzle 3 and Its J-Groove
Weld on the Davis-Besse Reactor Vessel Head,” in Proc. 12th Int. Conf. on
Environmental Degradation of Materials in Nuclear Power Plant.
[25] G. A. Y. D.S. Morton, S.A. Attanasio, E. Richey, “n Search of the True Temperature and
Stress Intensity Factor Dependencies for PWSCC,” in 12th Int. Conf. on Environmental
Degradation of Materials in Nuclear Power Systems–Water Reactors, 2005, pp. 977–
988.
[26] F. W. P. G. Economy, R.J. Jacko, “IGSCC Behaviour of Alloy 600 Steam Generator
Tubing in Water or Steam Tests above 360 °C,” Corrosion., vol. 43, pp. 727–734, 1987.
[27] J. A. G. R.W. Staehle, “Quantitative Assessment of Submodes of Stress Corrosion
Cracking on the Secondary Side of Steam Generator Tubing in Pressurized Water
Reactors,” Corrosion, vol. 60, pp. 115–180, 2004.
99
[28] M. A. K. Arioka, T. Yamada, T. Miyamoto, “Intergranular Stress Corrosion Cracking
Growth Behavior of Ni-Cr-Fe Alloys in Pressurized Water Reactor Primary Water,”
Corrosion, vol. 70, pp. 695–707, 2014.
[29] D. N. S. S.-I. Baik, M.J. Olszta, S.M. Bruemmer, “Grain-boundary structure and
segregation behavior in a nickel-base stainless alloy,” Scr. Mater., vol. 66, pp. 809–812,
2012.
[30] J. W. P.L. Andresen, J. Hickling, A. Ahluwalia, “Effect of Dissolved Hydrogen on SCC
of Ni Alloys and Weld Metals,” NACE Corros., 2009.
[31] M. K. S. E. Richey, D.S. Morton, “SCC Initiation Testing of Nickel-Based Alloys Using
In-Situ Monitored Uniaxial Tensile Specimens,” in , Proc. 12th Int. Conf. on
Environmental Degradation of Materials in Nuclear Power Plant.
[32] P.M. Scott, “An Overview of Internal Oxidation as a Possible Explanation of
Intergranular Stress Corrosion Cracking of Alloy 600 in PWRSProc. 9th Int. Conf. on
Environmental Degradation of Materials in Nucle,” in Proc. 9th Int. Conf. on
Environmental Degradation of Materials in Nuclear Power Plant.
[33] T. M. A. D.S. Morton, S.A. Attanasio, G.A. Young, P.L. Andresen, “The Influence of
Dissolved Hydrogen on Nickel Alloy SCC: A Window to Fundamental Insight,” NACE
Corros., 2001.
[34] N. Np-t-, “IAEA Nuclear Energy Series Stress Corrosion Cracking in Light Water
Reactors : Good Practices and Lessons Learned,” Int. At. Energy Agency, 2011.
[35] H. W. H. Coriou, L. Grall, P. Olivier, “Influence of Carbon and Nickel Content on Stress-
Corrosion Cracking of Austenitic Stainless Alloys in Pure or Chloride Containing Water
at 350 oC,” in n: R.W. Staehle (Ed.), Proc.of Conference of Fundamental As.
[36] K. O. T. Yonezawa, “Effect of chemical compositions and microstructure on the stress
corrosion cracking resistance of nickel-based alloys in high-temperature water.,” Proc.
Int. Conf. Eval. Mater. Perform.
[37] M. L. C. P.M. Scott, “Some Possible Mechanisms of Intergranular Stress Corrosion
Cracking of Alloy 600 in PWR Primary Water,” in Proc. 6th Int. Conf. on Environmental
Degradation of Materials in Nuclear Power Systems–Water Reactors.
[38] P. M. P. Combrade, P.M. Scott, M. Foucault, E. Andrieu, “Oxidation of Ni Base Alloys
in PWR Water: Oxide Layers and Associated Damage to the Base Metal,” in Proc. 12th
Int. Conf. on Environmental Degradation of Materials in Nuclear Power Plant.
[39] R. P. L. Fournier, O. Calonne, P. Combrade, P. Scott, P. Chou, “Grain boundary
oxidation and embrittlement prior to crack initiation in Alloy 600 in PWR primary water,”
in Proc. 15th Int. Conf. on Envir.
100
[40] P. M. S. R.C. Newman, T.S. Gendron, “Internal Oxidation and Embrittlement of Alloy
600,” in Environmental Degradation of Materials in Nuclear Power Systems–Water
Reactors, 1999, pp. 79–93.
[41] F. Scenini, “The Effect of Surface Preparation on the Oxidation and SCC Behaviour of
Alloy 600 and 690 in Hydrogenated Steam, Ph.D. Dissertation,” 2006.
[42] R. J. J. F. Scenini, R.C. Newman, R.A. Cottis, “Effect of Surface Preparation on
Intergranular Stress Corrosion Cracking of Alloy 600 in Hydrogenated Steam,”
Corrosion, vol. 64, pp. 824–835, 2008.
[43] W. K. Z. Zhang, J. Wang, E.-H. Han, “Analysis of Surface Oxide Films Formed in
Hydrogenated Primary Water on Alloy 690TT Samples With Different Surface States,”
J. Mater. Sci. Technol., vol. 30, pp. 1181–1192, 2014.
[44] W. K. Z. Zhang, J. Wang, E.-H. Han, “Influence of dissolved oxygen on oxide films of
Alloy 690TT with different surface status in simulated primary water,” Corrosion, vol.
53, pp. 3623–2635, 2011.
[45] D. V. R. Bandy, “Mechanisms of stress corrosion cracking and intergranular attack in
alloy 600 in high temperature caustic and pure water,” J. Mater. Energy Syst., vol. 7, pp.
237–245, 1985.
[46] D.H. Hur, M.S. Choi, D.H. Lee, M.H. Song, S.J. Kim, J.H. Han, “Effect of shot peening
on primary water stress corrosion cracking of Alloy 600 steam generator tubes in an
operating PWR plant,” Nucl. Eng. Des., vol. 227, pp. 155–160, 2004.
[47] S. Le Hong, “Influence of Surface Condition on Primary Water Stress Corrosion
Cracking Initiation of Alloy 600,” Corrosion, vol. 57, pp. 323–333, 2001.
[48] R. J. J. F. Scenini, R.C. Newman, R.A. Cottis, “Effect of Surface Preparation on
Intergranular Stress Corrosion Cracking of Alloy 600 in Hydrogenated Steam,”
Corrosion., vol. 64, pp. 824–835, 2008.
[49] S. M. B. M.B. Toloczko, M.J. Olszta, “One Dimensional Cold Rolling Effects of Stress
Corrosion Crack Growth in Alloy 690 Tubing and Plate Materials,” in Proc. 15th Int.
Conf. on Environmental Degradation.
[50] K. A. P.L. Andresen, M.M. Morra, “SCC of Alloy 690 and its Weld Metals,” in Proc.
15th Int. Conf. on Environmental Degradation of Materials in Nuclear Power Systems–
Water Reactors, 2011, pp. 161–176.
[51] L. T. S. Bruemmer, M. Olszta, M. Toloczko, “High-Resolution Characterizations of
Grain Boundary Damage and Stress Corrosion Crack Tips in Cold-Rolled Alloy 690,”
in Proc. 15th Int. Conf. on Environmental Degradation of Materials in Nuclear Power
Systems–Water Reactors, 2011, pp. 301–314.
101
[52] T. T. K. Arioka, T. Yamada, T. Miyamoto, “Dependence of Stress Corrosion Cracking
of Alloy 690 on Temperature, Cold Work, and Carbide Precipitation - Role of Diffusion
of Vacancies at Crack Tips,” Corrosion, vol. 67, pp. E1–E18, 2011.
[53] and C. H. H. J. S. M. Bruemmer, L. A. Charlot, “Microstructure and microdeformation
effects on IGSCC of Alloy 600 steam generator tubing.,” Corrosion, vol. 11, no. 44,
1988.
[54] R. W. S. and J. A. Gorman., “Quantitative assessment of submodes of stress corrosion
cracking on the secondary side of steam generator tubing in pressurized water reactors.,”
Corrosion, pp. 115–180, 2014.
[55] G. P. Airey, “Microstructural aspects of the thermal treatment of Inconel alloy 600,”
Metallography, vol. 13, pp. 21–41, 1980.
[56] F. S. R.C. Newman, “Another Way to Think About the Critical Oxide Volume Fraction
for the Internal-to-External Oxidation Transition,” Corrosion, vol. 64, pp. 721–726,
2008.
[57] J. R. M. G.S. Was, “The Influence of Grain Boundary Precipitation on the Measurement
of Chromium Redistribution and Phosphorus Segregation in Ni-16Cr-9Fe,” Metall.
Trans. A, vol. 16, pp. 349–359, 1985.
[58] G. S. W. S.M. Bruemmer, “Microstructural and microchemical mechanisms controlling
intergranular stress corrosion cracking in light-water-reactor systems,” J. Nucl. Mater.,
vol. 216, pp. 348–363, 1994.
[59] L. E. T. S.M. Bruemmer, M.J. Olszta, M.B. Toloczko, “Linking Grain Boundary
Microstructure to Stress Corrosion Cracking of Cold-Rolled Alloy 690 in Pressurized
Water Reactor Primary Water,” Corrosion, vol. 69, pp. 953–963., 2013.
[60] N. S. T. Yonezawa, K. Onimura, N. Sakamoto, “Effect of heat teatment on SCC
resistance of high Nickel alloys in high temperature water,” in Environmental
Degradation of Materials in Nuclear Power Systems–Water Reactors.
[61] T. K. and H. HANNINEN, “THE EFFECT OF HEAT TREATMENT ON THE
MICROSTRUCTURE AND CORROSION RESISTANCE OF INCONEL X-750
ALLOY,” Corros. Sci., vol. 25, pp. 789–803, 1985.
[62] T. S. V.S. Raja, “Stress Corrosion Cracking: Theory and Practice,” in Elsevier Science, .
[63] H. L. Logan, “Film-rupture mechanism of stress corrosion,” ,ournal Res. Natl. Bur.
Stand., vol. 48, pp. 99–105., 1952.
[64] R. W. Staehle, “Predicting failures in light water nuclear reactors which have not yet
been observed- microprocess sequence approach ( MPSA ),” Environ. Crack. Mater.,
vol. 2, pp. 3–53, 2008.
102
[65] F. P. F. P.L. Andresen, “Fundamental modeling of environmental cracking for improved
design and lifetime evaluation in BWRs.,” Int. J. Press. Vessel. Piping., no. 59, pp. 61–
70, 1994.
[66] J. R. Galvele, “A stress corrosion cracking mechanism based on surface mobility,”
Corros. Sci., vol. 27, pp. 1–33, 1987.
[67] R. B. Rebak and Z. Szklarska-Smialowska, “The mechanism of stress corrosion cracking
of alloy 600 in high temperature water,” Corros. Sci., vol. 38, no. 6, pp. 971–988, 1996.
[68] J. R. G. S.B. Farina, G.S. Duffó, “Stress corrosion cracking of copper and silver, specific
effect of the metal cations,” vol. 47, pp. 239–245, 2005.
[69] H. K. Birnbaum, “Mechanical properties of metal hydrides,” J. Less-Common Met., vol.
104, pp. 31–41, 1984.
[70] P. S. H.K. Birnbaum, “Hydrogen-enhanced localized plasticity--a mechanism for
hydrogen-related fracture,” Mater. Sci. Eng., vol. 76, pp. 191–202, 1994.
[71] D. J. U. S.L. Lee, “A decohesion model of hydrogen assisted cracking,” Eng. Fract.
Mech., vol. 31, pp. 647–660, 1988.
[72] H. K. B. S. Gahr, M.L. Grossbeck, “Hydrogen embrittlement of Nb I—Macroscopic
behavior at low temperatures,” Acta Metall., vol. 25, pp. 125–134, 1977.
[73] S. E. S. Myers, M. Baskes, H. Birnbaum, J. Corbett, G. DeLeo, “Hydrogen interactions
with defects in crystalline solids,” Rev. Mod. Phys., vol. 64, pp. 559–617, 1992.
[74] B. Lawn, Fracture of Brittle Solids. 1993.
[75] O. D. B. F. Foct, T. Magnin, “Stress corrosion cracking mechanisms of alloy 600
polycrystals and single crystals in primary water—Influence of hydrogen,” Metall.
Mater. Trans. A, vol. 31, pp. 2025–2036, 2000.
[76] ASTM G-30, “Making and Using U-Bend Stress-Corrosion Test Specimens 1,” vol. 97,
no. Reapproved 2003, pp. 1–7, 2008.
[77] H. Agrawal, P. Sharma, P. Tiwari, R. V. Taiwade, and R. Dayal, “Evaluation of Self-
Healing Behaviour of AISI 304 Stainless Steel,” Trans. Indian Inst. Met., vol. 68, 2015.
[78] A. S. M. International and A. Rights, ASM specialty handbook: nickel, cobalt, and their
alloys, vol. 38, no. 11. 2013.