發展輻射促進溶質原子偏析數值模擬程式，探討不同劑量質子輻射對奧斯田不□鋼溶質元素偏析，以及輻射促進應力腐蝕裂縫成長速率的影響，為本研究主要目的。

本研究以反科肯道爾效應為基礎，發展輻射促進溶質原子偏析數值模擬程式，並補充考慮輻射傷害效率、元素缺陷耦合效應及差排等缺陷造成之空缺－填隙原子對流失效應，可合理預測輻照敏化熱處理 (thermally sensitized , SEN)及工廠退火(mill annealled or as received, AR)304不□鋼之晶界溶質原子濃度分布。模擬程式根據目前質子校正的熱力學參數，並輔以適當的FMD計算，可以延伸用以評估中子輻射敏化，與文獻上中子實驗數據比對可合理預測鉻乏趨勢。

實驗研究以加速器產生的高能質子束照射AR304SS及SEN304SS，模擬中子輻射對不同起始狀態材料的影響，使用電子顯微鏡測量晶界微化學成份分佈的情形，分別使用電化學雜訊監測法及反轉直流電位降方法量測質子照射304不□鋼試片之裂縫成長速率，並輔以再鈍化電流量測，以探討輻射促進應力腐蝕裂縫成長的機制。

電化學雜訊法量測輻射促進應力腐蝕，無論使用被動式固定應變施力法或主動式施力法，即使塗佈適當材料遮蔽試片表面，或是拉近配對電極，均無法獲得電化學雜訊。可能原因是輻射促進應力腐蝕為膜破裂/金屬溶解過程，裂縫尖端的陽極電流在裂縫兩邊陰極區消耗，或是測試環境為純水，阻抗太高，以致電流無法流出至裂縫外配對電極被零電阻安培計檢出。

以反轉直流電位降方法則可以成功測得輻射促進應力腐蝕裂縫起始及成長。結果發現輻射會增進裂縫成長速率，輻射劑量愈高，裂縫成長速率愈高；但原有熱處理歷史會影響輻照304SS之敏化程度，進而影響裂縫成長速率。而較低的晶界鉻濃度是造成高輻射敏化程度及高裂縫成長速率的主要原因。輻照前鉻濃化者，輻照將加強其鉻乏的速度，反之，輻照前鉻耗乏者，輻照將減緩其鉻乏的速度。但不管原熱處理歷史或輻照歷史，輻照304SS之裂縫成長速率與晶界鉻濃度有極佳線性關連性。 不同前熱處理之輻照304SS，其EPR值可作為評估裂縫成長速率的定性指標。

綜合飽和氧環境及正常水化學環境下之輻射促進應力腐蝕裂縫成長結果，發現輻射促進應力腐蝕裂縫成長除受輻照敏化影響外，並受水環境中導電率及電化學電位影響，輻照敏化304不□鋼的保護電位約在－200 mVSHE附近。其結果可以使用膜破裂/金屬溶解模式預測。

輻照引發的不同敏化程度，改變了輻照304SS的再鈍化能力，因而影響裂縫成長速率；輻照304SS的再鈍化電流比未輻照304SS為高，這與裂縫成長速率量測發現的結果一致。使用累積電荷來評估輻照效應，可與裂縫成長速率有良好對應。而且再鈍化電流轉換為裂縫成長速率的計算結果也與實驗量測值相符，再次驗證輻射促進應力腐蝕裂縫成長受膜破裂/金屬溶解過程影響。輻射除了影響氧化膜再鈍化能力，可能也提升了鈍化膜破裂頻率，二者綜合的效應使得輻照不鏽鋼裂縫成長速率增加。

未來研究建議：1.增加輻照劑量及輻照深度，探討高劑量輻射材料在低電位水質環境應力腐蝕裂縫成長行為2. 改良本計算模式，併合熱處理及輻射引發溶質原子偏析，加入FMD自動產生項，以及考慮高劑量輻射引發之潛變及空洞等機制，用以預測高劑量輻射引發溶質原子偏析行為。3. 建議進行輻照不鏽鋼在不同溫度下之裂縫成長測試，驗證輻照與非輻照不鏽鋼之活化能差異。

Developing the model for simulating radiation induced segregation (RIS) process, investigating the effect of proton irradiation on RIS and stress corrosion cracking propagation are the objectives of this study.

The RIS model which is based on inverse Kirkendall effect incorporates radiation damage rate effect, atom-defect coupling effect, and interstitial-vacancy pair loss term toward dislocations, and therefore can predicts the RIS concentration profile of irradiated SEN304SS and AR304SS reasonably. According to the thermodynamic parameters benchmarked in proton irradiation data, the updated RIS can be extended to evaluate the Cr depletion evolution in neutron irradiated materials.

Experimental works include proton irradiations on SEN304SS and AR304SS in tandem accelerator, grain boundary microchemistry examinations by FEG-TEM/EDX, SCC crack growth rate measurements by RDC and ECN techniques. The repassivation kinetics is analyzed for elucidating the cracking mechanism of irradiation assisted stress corrosion cracking.

Electrochemical noise technique failed to measure the propagation of stress corrosion crack, both under load application by passive bolt loaded and active tension loaded. Applying ZrO2 coating to shielding extra surface, and shorten the distance between counter and working electrodes made no improvements. The anodic current in crack tip was probably consumed on the crack flank, or couldn’t flow out toward counter electrode because of high impedance of the testing solution.

Reversing DC techniques successfully measured the crack propagation of proton irradiated stainless steels. The crack growth rate is enhanced by proton irradiation, but also affected by prior heat treatments. The Cr depletion rate was enhanced by irradiation when initial grain boundary Cr concentration was enriched before irradiation (in SA condition), but radiation induced Cr decrement developed more slowly if the initial Cr concentration was depleted (in SEN condition). The intergranular crack growth rate of proton irradiated SS correlates linearly with the Cr concentration at grain boundary regardless of its prior thermal or irradiation histories. The EPR value can be treated as a qualitative indicator to evaluate the crack growth rate of proton irradiated SS304 with different prior thermal treatments.

In simulating BWR environments of oxygen saturated or normal water chemistry, Irradiation assisted stress corrosion cracking are affected by radiation sensitization, solution conductivity and ECP. The protection ECP is -200mVSHE for proton irradiated SEN304 stainless steel. The IASCC growth rate was reasonably predicted by PLEDGE model which was developed on the basis of film rupture/metal dissolution mechanism.

Repassivation kinetics analysis verified that radiation sensitization alter the repassivation ability and consequently enhance the stress corrosion crack propagation. The crack growth rate of proton irradiated stainless steel can be predicted by converting the repassivation current according to film rupture/metal dissolution model. However, irradiation affects not only repassivation current, but also film rupture frequency.

Three future studies are suggested. 1. Increase the radiation dose and irradiation depth. Investigate the effect of higher radiation dose on the SCC cracking behavior. 2. Refine RIS model by considering thermal and radiation sensitization coupling effect, FMD source term, and incorporating high dose effect of creep and void. 3. Perform the SCC test of irradiated stainless steel at various temperatures.

參考文獻
[1.] S.M. Bruemmer, “Quantitative Modeling of Sensitization Development in Austenitic Stainless Steel”, CORROSION 89, paper #561
[2.] E.A. Kenik, et al, 15th Sym. On Effects of Radiation on Materials, June 1990
[3.] D.I.R. Norris, et al, 15th Sym. On Effects of Radiation on Materials, June 1990
[4.] A.J. Jacobs, in: Effects of Radiation on Materials: 16th International Symposium, ASTM STP 1175, ASTM, Philadelphia, January, 1994, p. 902-918
[5.] M. Kodama, Sym. Of 6th Environmental, p.583
[6.] K. Asano, Sym. Of 5th Environmental, p.838
[7.] S.M. Bruemmer, B.W. Arey, L.A. Charlot, in: Proc. of 6th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, San Diego, California, August 1-5, 1993, The Minerals, Metals & Materials Society (TMS), Warrendale, PA, 1993, p. 277.
[8.] A. J. Jacobs 16th ASTM Sym Radiation Effects on Materials 1992.
[9.] J. Cookson, et al, Proc. of 6th Environmental Degradation, 1993, p
[10.] Y. Isobe, Personal Communication
[11.] A. Jenssen, et al ,“The Importance of Mo on IASCC in SS” corrosion 96, paper #101.
[12.] S. Watanabe, et al, Journal of Nuclear Materials, 224(1995)158
[13.] T. Onchi, et al, Corrosion 96, Paper #117.
[14.] A.J. Jacobs, Corrosion Science, vol.46, No. 8, p.634, 1990
[15.] W.G. Johnston, et al, Radiation Effects in Breeder Reactor Structural Materials, p.421
[16.] J.L. Brimhall, et al, J. Nucl. Mater. Vol. 117, p. 218, 1983
[17.] J.L. Brimhall, et al, J. Nucl. Mater. Vol. 122, p. 196, 1984
[18.] C.M. Stepherd, J. Nucl. Mater. Vol. 175, p. 170, 1990
[19.] N.H. Packan, et al, “ Radiation Induced Segregation in Light Ion Bombarded Ni-8 at% Si”, 13th Sym. On Effects of Radiation on Materials, June, 1986
[20.] R.D. Carter, et al, J. Nucl. Mater. Vol. 205, p. 361, 1993
[21.] J.M. Cookson, et al, J. Nucl. Mater. Vol. 202, p. 104, 1993
[22.] K. Fukuya, et al, J. Nucl. Mater. Vol. 191-194, p. 1007, 1992
[23.] D.L. Damcott, et al, J. Nucl. Mater. Vol. 225, p. 97, 1995
[24.] S.M. Bruemmer, et al, Sym of 5th Environmental, p.821, 1991
[25.] 劉文仁博士論文，質子輻射對商用304與304L不□鋼合金微觀分析與應力腐蝕破裂之研究，國立清華大學，中華民國八十三年六月九日
[26.] P.R. Okamoto, et al, J. Nucl. Mater. Vol. 53, p. 336, 1974
[27.] H. Takahashi, et al, J. Nucl. Mater. Vol. 103-104, p. 1415, 1981
[28.] H. Takahashi, et al, ASTM-STP 1047(1990)p.93
[29.] T. Ezawa, et al, Ultramicroscopy, 19(1991)p.187
[30.] K. Fukuya, et al, Scripta Metall., 19(1985)p.959
[31.] K. Nakata, et al, J. Nucl. Mater. Vol. 133-134, p. 575, 1985
[32.] K. Nakata, et al, J. Nucl. Mater. Vol. 150, p. 186, 1987
[33.] T.R. Anthony, “Solute Segregation and Stresses around Growing Voids in Metals”, Proc. Radiation induced Voids in Metals and Alloys, USAEC Sym. Series 26, CONF. 710601, p.630, 1971
[34.] Marwick, A. D.; Piller, R. C.; Sivell, P. M. Source, “Mechanisms Of Radiation-Induced Segregation In Dilute Nickel Alloys.”, Journal of Nuclear Materials, v 83, n 1, Aug, 1979, p 35-41
[35.] H. Wiedersich, et al, J. Nucl. Mater. Vol. 83, p. 98, 1979
[36.] N.Q. Lam, et al, Effects of Radiation on Materials, 11th Conference, p.985, 1982
[37.] J.M. Perks, et al, Proc. Conf. On Radiation Induced Sensitization of Stainless Steels, ed. D.I.R. Norris (CEGB, Berkeley Nuclear Labs, Berkeley), 1987, p.15
[38.] E.P. Simonen, et al, Corrosion 93, paper #615
[39.] G.S. Was, et al, J. Nucl. Mater. Vol. 216, p. 326, 1994
[40.] M.A. Kenik, et al, J. Nucl. Mater. 80(1979)159
[41.] L.E. Rehn, et al, J. Nucl. Mater. 205(1993)31
[42.] H. Wiedersich, P.R. Okamoto and N.Q. Lam, " A Theory of Radiation Induced Segregation in Condensed Alloys", J. Nucl. Mater. 83(1979) 98-108.
[43.] C. H. Woo, A. A. Semenovb and B. N. Singhc, Journal of Nuclear Materials, 206(1993)170-199
[44.] V. Naundorf, et al, Journal of Nuclear Material, 182(1991)254
[45.] P. Andresen anf F.P. Ford, “Modeling of Irradiation Effects on SCC Growth Rates”, Corrosion/89, paper #497, April 17-21, 1989, New Orleans, USA
[46.] K.S. Ramp, G.M. Gordon, “Fabrication and Operating History Considerations in Assessing Relative SCC Susceptiblity of BWR Components”, 5th Environmental Degradation Conference, 1991, P.365.
[47.] A.J. Jacobs, et al , “IASCC and Grain Bound Segregation in Heat Treated 304SS ”, Proc. 14th Int. Sym on Effects of Radiation on Materials, June, 1988.
[48.] A.J. Jacobs, et al, “Radiation Effects on SCC and Other Selected Properties of 304 & 316 S.S.” , Metallurgical Society, 1988., P.673.
[49.] K. Fukuya, et al , “An IASCC Study Using High Energy Ion Irradiation”, 5th Environmental Degradation Conference, 1991, P.281.
[50.] M. Kodama, et al, “Influence of Impurities and Alloying Elements on IASCC in N Irradiated S.S. ”, 16th Int. Sym. on Effects of Rad. on Materials, 1992.
[51.] G. Was, et al, “Effects of irradiation on IGSCC ”, Journal of Nuclear Materials, 216(1994)P.326.
[52.] Y. Isobe, et al, Submitted to J. of Null Mat.
[53.] O. Sudo, et al, 5th Environmental Degradation , P.251.
[54.] A. J. Jacobs, corrosion, vol. 50 , P.731, 1994.
[55.] H. Chung, 6th Environmental Degradation Conf. P511.
[56.] M. Tsubota, et al, 7th Environmental Degradation , P.519.
[57.] P.L. Andresen, “Modeling & Prediction of IASCC”, in: Proc. of 7th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Breckenridge, Colorado, August 7-10, 1995, NACE International, Houston, TX, 1995, p. 893.
[58.] USNRC / NUREG / 0313, rev 2. 1988.
[59.] F.P. Ford, EPRI-NP-5064S, 1987.
[60.] 核能反應器及附屬設備壽命評估79年度期末報告，工業技術研究院工業材料研究所，MRL TPC 79-A2, (1990), P.2-22.
[61.] S. Uchida, et al, 7th Environmental Degradation ,P.959.
[62.] R.L. Jones, G.M. Gordon , and G.H. Neils, “Environmental Degradation of Materials in Boiling Water Reactors ”. Proceedings of the Fourth International Symposium of Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, NACE(1990),P.1-1~1-10.
[63.] P.L. Andresen, “Effects of Specific Anionic Impurities on Environmental Cracking of Austenitic Material in 288℃ Water”, Proceedings of the Fifth International Symposium on Environmental Degradation of materials in nuclear Power Systems-Water Reactors, American nuclear Society, (1992) P.209-218.
[64.] L. G. Ljungberg, D. Cubicciotti, and M. Trolle, “Effect of Water Impurities in BWR on Environmental Crack Growth under Realistic Load Conditions”, Proceeding of the Fourth International Symposium on Environmental Degradation of materials in nuclear Power Systems-Water Reactors, NACE, (1990), P.4-56-4-75.
[65.] S. Tahtinen, H. Hanninen, and M. Trolle, “Stress Corrosion Cracking of Cold Worked Austenitic Stainless Steel Pipes in BWR Reactor Water”, Proceedings of the Sixth International Symposium on Environmental Degradation of materials in nuclear Power Systems-Water Reactors, TMS (1993). P.265-270.
[66.] D.D. Macdonald and M. Urquidi-Macdonald, “Modeling of the Electrochemistry of Stress Corrosion Cracks in Sensitized Type 304 S.S. in Boiling Water Reactors”, Proceeding of The Fourth International Symposium on Environmental Degradation of Materials in nuclear Power Systems-Water Reactors, NACE, (1990), P.4-1~4-10.
[67.] P. L. Andresen, “Effect of Noble Metal Coating & Alloying on Stress Corrosion Crack Growth Rate of Stainless Steel in 288℃ Water”, Proceedings of the Sixth International Symposium on Environmental Degradation of materials in nuclear Power Systems-Water Reactors, TMS (1993) P.245-252.
[68.] L. G. Ljungberg, D. Cubicciotti, and M. Trolle, “The Effect of Sulfate on Environmental Cracking in Boiling Water Reactors under Constant Load or Fatigue Corrosion 46, (1990), P.641.
[69.] W. E. Ruther, W. K. Soppet, and T. F. Kassner, “Effect of Temperature and Ionic Impurities at Very Low Concentrations on Stress Corrosion Cracking of Type 304 Stainless Steel”, Corrosion 44, (1988). P.791.
[70.] P. L. Andresen, “Environmentally Assisted Crack Growth Rate Response of Non-sensitized Stainless Steels in High Temperature Water” Corrosion, Vol. 44, (1988). P.450-460.
[71.] T. Shoji, K. Yamaki, R. G. Ballinger, and I. S. Hwang, “Grain Boundary Segregation and Intergranular Stress Corrosion Cracking Susceptibility of Austenitic Stainless Steels in High Temperature Water”, Proceedings of the Fifth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, ANS, (1992), P.827-837.
[72.] U . S. Nuclear Regulatory Commission, “Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping” Final Report, NUREG-0313-Revision 2, 1988.
[73.] F. P. Ford, “Corrosion-Assisted Cracking of Stainless and Low-Alloy Steels in LWR Environments”, EPRI NP-5064S, 1987.
[74.] K. Fukuya, S. Shima, K. Nakata, S. Kasahara, A. J. Jacobs, G. P. Wozadlo, S. Suzuki, and M. Kitamura, “Mechanical Properties and IASCC Susceptibility in Irradiated Stainless Steels”, the Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power System-Water Reactors, TMS (1993), P.565-572.
[75.] J. Takagi et al, “Evaluation of Corrosion Environment in BWR Primary Circuit by Water Radiolysis Model”, Proceedings off Water Chemistry in Nuclear Power Plants, JAIF, Tokyo, (1988).
[76.] P. L. Andresen, F. P. Ford, S. M. Murphy, and J. M. Perks, “State of Know ledge of Radiation Effects on Environmental Cracking in Light Water Reactor Core Materials”, Proceedings off the Fourth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, NACE, (1990).P.1-83~1-120.
[77.] H. M. Chung, W. E. Ruther, J. E. Sanechi, and T. F. Kassner, “Irradiation-Induced Sensitization and Stress Corrosion Cracking of Type 304 Stainless Steel Core-Internal Components”, Proceedings of The Fifth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, American Nuclear Society, (1992). P.795-805.
[78.] H. M. Chung, W. E. Ruther, J. E. Sanechi, A. Hins, and T. F. Kassner, “Effects of Water Chemistry on Intergranular Cracking of Irradiated Austenitic Stainless Steels”, Proceedings of The Seventh International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, NACE, (1995). P.1133-1143.
[79.] A. Jenssen and L. G. Ljungberg, “Irradiation Assisted Stress Corrosion Cracking of Stainless Alloys in BWR Normal Water Chemistry and Hydrogen Water Chemistry”, Proceedings of The Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, TMS (1993), P.547-553.
[80.] M. Kodama, S. Nishimura, J. Morisawa, S. Suzuki, S. Shima, and M. Yamamoto, “Effect of Fluence and Dissolved Oxygen on IASCC in Austenitic Stainless Steels”, Proceedings of The Fifth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, TMS (1993), P.948-954.
[81.] W. L. Clarke and A. J. Jacobs, “Effect of Radiation Environment on SCC of Austenitic Materials”, Proceedings of First International Symposium on Environmental Degradation of Materials in Nuclear Power System-Water Reactors, NACE (1983), P.451.
[82.] R. Katsura, J. Morisawa, M. Kodama, S. Nishimura, S. Suzuki, S. Shima, and M. Yamamoto, “Effect of Stress on IASCC in Irradiated Austenitic Stainless Steels”, Proceedings of The Sixth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors, TMS (1993), P.625-63.
[83.] P. Andresen, P. Ford, R. Davis, W. Miller, and P. Scott, “Irradiation Effects on SS Crack Growth Behavior”, 12th ICG-IASCC (`1996).
[84.] J. Noggerath and G. Prantl, “Uncertainties in Root Cause Analysis and Crack Growth Predictions in the Case of NPP Muhleberg‘s Core Shroud Problem”, 12th ICG-IASCC (1996).
[85.] 核一廠二號機反應器爐心側板H3焊道龜裂評估報告，台灣電力公司，83年4月
[86.] S. Dumbill, W. Hanks. Sixth International Symposium on Environmental Degradation of materials in Nuclear Power System - Water Reactors. (Eds. R.E. Gold and E.P. Simonen) The Minerals, Metals & Materials Society (1993).
[87.] P.J.Maziasz, and C.J.McHargue, International Materials Reviews, 32(1987)190.
[88.] E.E.Bloom, W.R.Martin, J.O.Stiegler, and J.R.Weir, J.Nucl. Mater., 22(1967)68.
[89.] S.J.Zinkle, P.J.Maziasz, and R.E.Stoller, J.Nucl. Mater., 206(1993)266.
[90.] F.A.Garner, J.Nucl.Mater., 205(1993)98.
[91.] P.J.Maziasz, J.Nucl.Mater., 191-194(1992)701.
[92.] S. M. Bruemmer, et al, Mat. Res. Soc. Sym. Proc., Vol. 439, 1997, P.437
[93.] J. Gan, et al, 同上，P.445
[94.] O. Dimitrov, et al，Journal of Nuclear Materials 105 (1982)39－47。
[95.] S.J. Rothman, et al, "Self-diffusion in austenitic Fe-Cr-Ni alloys", J. Phys. F:Metal Phys., 10 (1980) 383-98.
[96.] E. P. Simonen, et al, JNM 225(1995)117-122
[97.] J. M. Perks and S.M. Murphy, Proc. Conf. on Materials for Nuclear Reactor Core Applications. Vol. 1 (BNES, 1987) P.119.
[98.] T.R. Allen et al, Proc. of 7th Environmental Degradation, P.997, 1995.
[99.] J. Ruzickova, B. Million, Material Science and Engineering 50 (1985) P.59.
[100.] B. Million, et al, Materials Science and Engineering 72(1985) P.85.
[101.] T.R. Allen, G.S. Was, Mat. Res. Soc, Symp. Proc., Vol. 439, P.539, 1997.
[102.] E.P. Simonen, S.M. Bruemmer, Mat. Ros. Soc. Symp Proc, Vol. 373, 1995, P.95.
[103.] S. M. Bruemmer, “Composition-Based Correlations to Predict Sensitization Resistance of Austenitic Stainless Steels”, Corrosion, Vol. 42, No.1 (1986), P.27-35.
[104.] P. Novak, R. Dtepec, and F. Franz, “Testing the Susceptibility of Stainless Steel to Intergranular Corrosion by a Reactivation Methods”, Corrosion, Vol.31, No.10, (1975), P.344-347.
[105.] R.W. Hertzberg, “Deformation and Fracture Mechanics of Engineering Materials”, 3rd edition, p.656, 1989
[106.] P.L. Andresen, in “ Stress Corrosion Cracking – Material Performance and Evaluation”, Ed. R. H. Jones (ASM 1992), p. 181.
[107.] M. Kodama, Y. Ishiyama, S. Namatame, S. Suzuki, K. Fukuya, H. Sakamoto, K. Nakata, T. Kato, in: Proc. of 9th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Newport Beach, California, August 1-5, 1999, The Minerals, Metals & Materials Society (TMS), Warrendale, PA, 1999, p. 923.
[108.] T. Onchi, K. Hide, M. Mayuzumi, K. Dohi, T. Niiho, Corrosion, 53 (1997) 778.
[109.] G.S. Was, T. Allen, J. Nucl. Mater. 205 (1993) 332.
[110.] O. Okada, K. Nakata, S. Kasahara and T. Aoyama, in: Proc. of the 8th International Symposium on Environmental Degradation of Materials in Nuclear Power Systems--Water Reactors, Amelia Island, Florida, August 10-14, 1997, American Nuclear Society (ANS) Inc., La Grange Park, Il, 1997, p. 743.
[111.] S.M. Bruemmer, L.A. Charlot, B.W. Arey, Corrosion 44 (1988) 328.
[112.] P.L. Andresen, Corrosion/95, paper #412
[113.] T. Shibata, S. Fujimoto, Transactions of the Japan Institute of Metals, Vol. 28, No. 5, 1987, p.424
[114.] T. Shibata, S. Fujimoto, Transactions of the Japan Institute of Metals, Vol. 28, No. 3, 1987, p.224
[115.] P. Andresen, CORROSION, Vol. 49, 1993, p.714
[116.] Q.J. Peng, J. Kwon, T. Shoji, “Development of a fundamental crack tip stran rate equation and its application to quantitative prediction of stress corrosion cracking of stainless steels in high temperature oxygenated water”, Journal of Nuclear Materials, 324(2004)52-61
[117.] Q.J. Peng, J. Kwon, T. Shoji, “Inverse Analysis of Crack Growth Kinetics of Sensitized 304L Stainless Steel in a Simulated Boling Water Reactor Environment for Qyuantitative Prediction of Stress Corrosion Cracking Behavior”, 10th Environmemtal Degradartion, 2002
[118.] F.P. Ford, “Quantitative Prediction of Environmentally Assisted Cracking”, Corrosion, Vol52, 1996, P.375
[119.] M. Gomez-Duran, D.D. Macdonald, “ Stress Corrosion Cracking of sensitized Type 304 stainless steel in thiosulfate solution:I. Fate of the coupling current”, Corrosion Science, 45(2003)1455-1471
[120.] D.D. Macdonald, P.-C. Lu, M. Urquidi-Macdonald, and T.-K. Yeh, “Theoretical Estimation of Crack Growth Rates in Type 304 Stainless Steel in Boiling-Water Reactor Coolant Environments”, 1996, CORROSION, p.768
[121.] M. Vankeerberghen, D.D. Macdonald, “Predicting crack growth rate vs. temperature behaviour of Type 304 stainless steel in dilute sulfuric acid solutions”, Corrosion Science, 44(2002)1425-1441
[122.] W.L. Clarke, The EPR method for the detection of sensitization in stainless steels, US Nuclear Regulatory Commission, Washington, DC, NUREG/CR-1095, February 1981.
[123.] P.L. Andresen, “Review of current research and understanding of irradiation-assisted stress corrosion cracking”, Proc. 5th Int. Symp. On Environmental Degradation of Materials in Nuclear Power systems, NACE., P.10, 1991
[124.] L.H.Wang, C.Fong, C.H.Tsai , J.J.Kai; “Comparison of SCC Susceptibility on Thermal Sensitized and Irradiated 304S.S. With ECN Techniques”, Presented at the 1997 (13th) ICG-EAC Annual Meeting, Bristol, UK, April, 1997
[125.] J.J.Kai, L.H.Wang, D.S.Tu, F.R.Chen, C.Fong; “Effects of Preexisting Composition Profiles on the Irradiation Assisted Stress Corrosion Cracking of Austenitic Stainless Steels”, Presented at the 8th international symposium on environmental degradation of materials in nuclear power systems- water reactors, Amelia Island, Florida, August, 1997
[126.] L.H.Wang, J.J.Kai, C.H.Tsai, C.Fong; “Comparison of Stress Corrosion Cracking Susceptibility of Thermally-Sensitized and Proton-Irradiated 304 Stainless Steel using Electrochemical Noise Techniques”, Presented at ICFRM-8, Eighth International Conference on Fusion Reactor Materials, Sendai, Japan, October, 1997
[127.] L.H.Wang, J.J.Kai, C.H.Tsai, C.Fong; “Comparison of Stress Corrosion Cracking Susceptibility of Thermally-Sensitized and Proton-Irradiated 304 Stainless Steel using Electrochemical Noise Techniques”, Journal of Nuclear Materials, 258-263(1998)2046-2053
[128.] L.H.Wang, C.Fong, C.H.Tsai , J.J.Kai; “Crack Growth Rate Measurement on Proton Irradiated Stainless Steels”, Presented at the 1998 (14th) ICG-EAC Annual Meeting, Hangzhou, China, April, 1998
[129.] L.H.Wang, C.Fong, C.H.Tsai , J.J.Kai; “Electrochemical Noise (ECN) Measurement on Constant-loaded Specimens”, Presented at the 1998 (14th) ICG-EAC Annual Meeting, Hangzhou, China, April, 1998
[130.] L.H.Wang, C.H.Tsai , J.J.Kai, D.S.Duh; “Effects of Pre-existing thermal sensitization on the radiation induced sensitization in type 304 stainless steels”, in “Microstructural Processes in Irradiated Materials” of Materials Research Society Symposium Proceeding, The Materials Research Society, Warrendale, PA, Nov. 30-Dec. 2 1999, Volume 540, p. 489.
[131.] L.H.Wang, C.H.Tsai , J.J.Kai; “Effect of Prior Thermal Treatment on the Crack Growth of Proton Irradiated AISI 304 Stainless Steels in High Temperature Water”, accepted by Journal of Nuclear Materials, April 2000
[132.] L.H.Wang, “Repassivation kinetics of irradiated stainless steel in high temperature water”, Presented at the 2001 ICG-EAC Annual Meeting, Kyongju, South Korea, April 23 - 27, 2001
[133.] L.H.Wang, “Repassivation Kinetics of Irradiated Stainless Steelin High Temperature Water”, Seminar on water chemistry of Nuclear Reactor System 2001, October 2001, Luntan, Taiwan, INER
[134.] T.S. Duh, J.J. Kai, F.R. Chen,; L.H. Wang, “Effects of grain boundary misorientation on the solute segregation in austenitic stainless steels”, Journal of Nuclear Materials, v 258-263, n pt B, Oct, 1998, p 2064-2068
[135.] T.S. Duh, J.J. Kai, F.R. Chen, L.H. Wang, “Numerical simulation modeling on the effects of grain boundary misorientation on radiation-induced solute segregation in 304 austenitic stainless steels”, Journal of Nuclear Materials, v 294, n 3, April, 2001, p 267-273
[136.] L.H. Wang, C.H. Tsai, J.J. Kai, “Effect of prior thermal treatment on the microchemistry and crack propagation of proton-irradiated AISI 304”, Journal of Nuclear Materials, 328 (2004) 11–20
[137.] NSC84-2212-E-007-008, 高能離子對低碳不鏽鋼微結構與應力腐蝕破裂影響研究 (I), 1995
[138.] NSC85-2212-E-007-011, 高能離子對低碳不鏽鋼微結構與應力腐蝕破裂影響研究 (II), 1996
[139.] NSC86-2212-E007-051, 質子與中子對輻射引發偏析機制之比較(I), 1997
[140.] NSC87-2212-E-007-067, 質子與中子對輻射引發偏析機制之比較(II), 1998
[141.] 陳彥和碩士論文，質子輻射對316不□鋼微結構及微化學成份偏析影響之研究，國立清華大學，中華民國八十五年六月二十一日
[142.] 杜定賢博士論文，304 不□鋼材料相互晶向差對晶界偏析的影響研究，國立清華大學，中華民國九十年七月