No 0 KAIST Seminar Corrosion of Structural Materials

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No. 0 KAIST Seminar Corrosion of Structural Materials and Electrochemistry in High Temperature Water

No. 0 KAIST Seminar Corrosion of Structural Materials and Electrochemistry in High Temperature Water of Nuclear Power Systems November 25 (Tue), 2008 Shunsuke Uchida Institute of Applied Energy Japan Atomic Energy Agency KAIST Seminar Nov. 25, 2008 S. Uchida

No. 1 ABSTRACT Latest experiences with corrosion in the cooling systems of nuclear power

No. 1 ABSTRACT Latest experiences with corrosion in the cooling systems of nuclear power plants are reviewed.   High temperature cooling water causes corrosion of structural materials, which often lead to adverse effects in the plants, e. g. , increasing shutdown radiation, generating defects in materials of major components and fuel claddings, and increasing the volume of radwaste sources.   Corrosion behaviors are much affected by water qualities and differ according to the combinations of water qualities and the materials themselves.   In order to establish reliable operation, each plant requires plant unique optimal water chemistry control based on careful consideration of its system, materials and operational history.     Electrochemistry is one of key issues to determine corrosion related problems but it is not only cause.   Most of phenomena of corrosion related problems, e. g. , flow accelerated corrosion (FAC), intergranular stress corrosion cracking (IGSCC), primary water stress corrosion cracking (PWSCC) and thinning of fuel cladding materials, can be understood based on the electrochemical index, e. g. , electrochemical corrosion potential (ECP), conductivities and p. H.   Major electrochemical index, ECP, can be measured at elevated temperature and applied as in situ sensors of corrosion conditions to detect anomaly conditions of structural materials and material abnormalities at their very early stages. KAIST Seminar Nov. 25, 2008 S. Uchida

No. 2 Table of contents 1. Nuclear power plants in the world 2. Interaction

No. 2 Table of contents 1. Nuclear power plants in the world 2. Interaction of cooling water and materials Major roles of cooling water in nuclear power plants and its adverse effects 3. Flow accelerated corrosion (Wall thinning)  4. Stress corrosion cracking (IGSCC and PWSCC) 5. Corrosion of zirconium alloy (Corrosion of fuel cladding) 6. Future subjects 7. Conclusion KAIST Seminar Nov. 25, 2008 S. Uchida

No. 3 World list of nuclear power plants PWR, VVER 69 58 24 13

No. 3 World list of nuclear power plants PWR, VVER 69 58 24 13 16 1 PHWR BWR 35 GCR, AGR LMFBR 31 17 1 4 18 18 11 15 15 9 3 6 LWGR 1 6 2 2 7 2 6 unit number Countries*1 total USA 104 France 59 Japan 55 Russia 31 South Korea 20 UK 19 Canada 18 Germany 17 India 17 Ukraine 15 China 11 Sweden 10 Spain 8 Belgium 7 Taiwan 6 Czech 6 Slovakia 5 Switzerland 5 Total*2 439 Share (%) 8 7 6 5 4 3 2 1 0 4 LWR in Japan BWR PWR >35 >30 >25 >20 >15 >10 >5 >0 on construction unit age (y) 5 3 260 (59) 43 (10) 2 93 (21) 18 (4) 23 (5) 2 (0. 5) *1: more than 5 plants Others: Finland, Hungary: 4 plants, Bulgaria, Argentina, Brazil, Mexico, Pakistan, South Africa: 2 plants, Romania, Armenia, Lithuania, Netherlands, Slovenia: 1 plant *2: 380 GWe Version 2008 Ref. Nuclear Energy Institute, “World Nuclear Generation and Capacity (2007)”, web site: http: //www. nei. org/resourcesandstats/documentlibrary/reliableandaffordableenergy/graphicsandcharts/worldnuclearpowerplantsinoperation/ KAIST Seminar Nov. 25, 2008 S. Uchida

No. 4 Table of contents 1. Nuclear power plants in the world 2. Interaction

No. 4 Table of contents 1. Nuclear power plants in the world 2. Interaction of cooling water and materials Major roles of cooling water in nuclear power plants and its adverse effects 3. Flow accelerated corrosion (Wall thinning)  4. Stress corrosion cracking (IGSCC and PWSCC) 5. Corrosion of zirconium alloy (Corrosion of fuel cladding) 6. Future subjects 7. Conclusion KAIST Seminar Nov. 25, 2008 S. Uchida

No. 5 Major roles and adverse effects of cooling water of nuclear power plants

No. 5 Major roles and adverse effects of cooling water of nuclear power plants BWR primary cooling water PWR primary and secondary cooling water major roles: energy transporting medium neutron moderating medium structural material integrity major roles: energy transport medium neutron moderating medium structural material integrity radioactive contamination radwaste source volume pre-filter demineralizer fuel integrity occupational exposure spent resin liquid waste (back wash, regeneration) recirculation system radwaste system main steam/ feed water systems fuel integrity occupational exposure primary cooling system secondary cooling system under line: items concerning adverse effects Ref. S. Uchida, “Latest Experience with Water Chemistry in Nuclear Power Plants in Japan”, Power Plant Chemistry, 8, 282 (2006). KAIST Seminar Nov. 25, 2008 S. Uchida

No. 6 Optimal water chemistry control (BWR and PWR) Fewer environmental impacts Reducing occupational

No. 6 Optimal water chemistry control (BWR and PWR) Fewer environmental impacts Reducing occupational exposure (radioactive contamination) iron, nickel & cobalt control Reducing radwaste sources Minimizing radioactive effluent Higher safety and higher reliability Improving reliability of structural materials p. H & hydrogen control Improving reliability of cladding materials radical control Establishing 4 targets Ref. S. Uchida, “Latest Experience with Water Chemistry in Nuclear Power Plants in Japan”, Power Plant Chemistry, 8, 282 (2006). KAIST Seminar Nov. 25, 2008 S. Uchida

Major accidents and incidents at nuclear facilities Plant (reactor type) Date Causes Three Mile

Major accidents and incidents at nuclear facilities Plant (reactor type) Date Causes Three Mile Island-2 (PWR) Mar. 1979 LOCA Chernobyl (LGR) Apr. 1986 RIA Surry-2 Fukushima Daini-3 Mihama-2 Monju (PWR) (BWR) (PWR) (LMFBR) JCO (conversion facility) Hamaoka-1 (BWR) Mihama-3 (PWR) *: related to material Dec. 1986 Jan. 1989 Feb. 1991 Dec. 1995 No. 7 Environmental effects <1 m. Sv 31 people died 16 k person Sv 4 people died none FAC* vibration CF parts defect (Na leakage ) Sep. 1999 critical 2 people died accident 130 residents received radiation dose Nov. 2001 H 2 explosion none Aug. 2004 FAC* 5 people died LOCA: loss of coolant accident RIA: reactivity initiated accident FAC: flow assisted corrosion CF: corrosion fatigue Ref. S. Uchida, “Latest Experience with Water Chemistry in Nuclear Power Plants in Japan”, Power Plant Chemistry, 8, 282 (2006). KAIST Seminar Nov. 25, 2008 S. Uchida

Major problems for structural materials in NPPs problem FAC reactor troubled location type PWR

Major problems for structural materials in NPPs problem FAC reactor troubled location type PWR feed water piping No. 8 countermeasures material exchange water chemistry improvement BWR feed water piping water chemistry improvement heater drain piping material exchange IGSCC BWR primary piping material exchange, stress improvement water chemistry improvement PWSCC PWR core internals material exchange, water chemistry improvement Fuel cladding PWR fuel material improvement corrosion BWR material improvement SG tubing defects PWR SG water chemistry improvement material exchange FAC: flow accelerated corrosion IGSCC: intergranular stress corrosion cracking PWSCC: pressurized water stress corrosion cracking Ref. S. Uchida, “Latest Experience with Water Chemistry in Nuclear Power Plants in Japan”, Power Plant Chemistry, 8, 282 (2006). KAIST Seminar Nov. 25, 2008 S. Uchida

Major materials in primary system No. 9 Their wetted surface zirconium alloy stainless steel

Major materials in primary system No. 9 Their wetted surface zirconium alloy stainless steel zirconium alloy carbon steel nickel alloy stainless steel  a) PWR (primary system)  b) BWR Ref. H. Takiguchi, H. Takamatsu, S. Uchida, K. Ishigure, M. Nakagami and M. Matsui, ” Water Chemistry Data Acquisition, Processing, Evaluation and Diagnosis Systems in Light Water Reactors”, J. Nucl. Sci. Technol. , 41, 212 (2004). KAIST Seminar Nov. 25, 2008 S. Uchida

Comparison of corrosion behaviors No. 10 and features of major materials Materials Carbon steel

Comparison of corrosion behaviors No. 10 and features of major materials Materials Carbon steel Corrosion rate (relatively) Oxide film Application high Problems magnetite/hematite piping of secondary system FAC Effects of strong electrochemistry Stainless steel (nickel alloy) medium Zirconium alloy Cr rich nickel ferrite piping and component of primary system IGSCC, PWSCC radioactivity accumulation medium zirconium oxide fuel cladding low clad thinning weak Ref. S. Uchida, “Corrosion of Structural Materials and Electrochemistry in High Temperature Water of Nuclear Power Systems”, Power Plant Chemistry, 10, (11) (2008, ) in press. KAIST Seminar Nov. 25, 2008 S. Uchida

Interaction between structural materials and cooling water materials water growth of oxide film composition

Interaction between structural materials and cooling water materials water growth of oxide film composition (impurities crystal structure local stress No. 11 oxide film temperature p. H conductivity oxidant release of metallic ions barrier for diffusion of oxidant and metallic ion consequence of corrosion/control corrosion Ref. S. Uchida, “Corrosion of Structural Materials and Electrochemistry in High Temperature Water of Nuclear Power Systems”, Power Plant Chemistry, 10, (11) (2008, ) in press. KAIST Seminar Nov. 25, 2008 S. Uchida

Major parameters of major corrosion induced phenomena No. 12 Flow accelerated corrosion(FAC) material factors

Major parameters of major corrosion induced phenomena No. 12 Flow accelerated corrosion(FAC) material factors Cr content flow dynamics factors environmental factors mass transfer due to flow turbulence temperature. p. H, [O 2] Intergranular stress corrosion cracking (IGSCC) Zirconium alloy corrosion material factors sensitization at heat affected zone stress factors residual stress at heat affected zone environmental factors radiolytic species, [O 2], [H 2 O 2] stress factors compression due to lattice constant environmental factors radiolytic species, [O 2], [H 2 O 2] Ref. S. Uchida, “Corrosion of Structural Materials and Electrochemistry in High Temperature Water of Nuclear Power Systems”, Power Plant Chemistry, 10, (11) (2008, ) in press. KAIST Seminar Nov. 25, 2008 S. Uchida

No. 13 Table of contents 1. Nuclear power plants in the world 2. Interaction

No. 13 Table of contents 1. Nuclear power plants in the world 2. Interaction of cooling water and materials Major roles of cooling water in nuclear power plants and its adverse effects 3. Flow accelerated corrosion (Wall thinning)  4. Stress corrosion cracking (IGSCC and PWSCC) 5. Corrosion of zirconium alloy (Corrosion of fuel cladding) 6. Future subjects 7. Conclusion KAIST Seminar Nov. 25, 2008 S. Uchida

No. 14 Rupture part of the piping hanger ruptured 2. 5 m orifice flange

No. 14 Rupture part of the piping hanger ruptured 2. 5 m orifice flange bent piping upstream (A) 1. 5 m 4. 0 m 0. 45 m upstream (B) 1. 9 m 1. 5 m 3. 4 m Ref. S. Uchida, “Latest Experience with Water Chemistry in Nuclear Power Plants in Japan”, Power Plant Chemistry, 8, 282 (2006), Final report about Mihama-3 secondary system piping failure, Nuclear and Industrial Safety Agency, Tokyo (2005). (in Japanese) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 15 Elemental models to evaluate FAC Determination of oxide film mass transfer coefficient

No. 15 Elemental models to evaluate FAC Determination of oxide film mass transfer coefficient by flow dynamics analysis (inner layer) (outer layer) Fe 2+ release to bulk Evaluation of ionic concentrations precipitation of oxide release of oxide Evaluation of erosion by water droplets base metal Evaluation of oxide film growth Evaluation of shear stress by flow boundary layer diffusion of oxidants Evaluation of corrosive conditions main flow cooling water : corrosion (chemical term) : flow dynamics (physical term) Major elemental models Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 16 Evaluation steps for FAC Step 1 1 D CFD code Distributions of

No. 16 Evaluation steps for FAC Step 1 1 D CFD code Distributions of flow velocity and temperature in the cooling system Step 2 1 D O 2 -hydrazine reaction code Corrosion conditions, e. g. , [O 2] and ECP, in the cooling system Step 3 3 D CFD code Distributions of mass transfer coefficients Step 4 Chart evaluation Danger zone evaluation Step 5 Wall thinning calculation code Distributions of wall thinning rate and ECP Step 6 final evaluation Evaluation of residual life and the effects of countermeasures Flow dynamics analysis Corrosion (chemical) analysis System analysis Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

FAC occurrence zone indicated by major parameters No. 17 Corrosive conditions Wall thinning rate

FAC occurrence zone indicated by major parameters No. 17 Corrosive conditions Wall thinning rate Mass transfer coefficient oxidant [O 2]<5 ppb ECP<-0. 3 V flow pattern mass transfer coeff. >threshold p. H<9. 2 temperature 120<T<180 FAC occurrence zone Evaluation of wall thinning Double oxide layer model [Fe 2+] [Fe]<1/2[Fe]sat Cr content [Cr]<0. 2% : coupled analysis : individual analysis Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 18 Corrosive condition calculation O 2 -N 2 H 4 reaction code, RADIOLYSIS-N

No. 18 Corrosive condition calculation O 2 -N 2 H 4 reaction code, RADIOLYSIS-N 2 H 4 100 50 0 SG heaters [O 2] in sampling line A B C 5 4 A B 3 150 100 B 2 1 0 C temperature C 0 50 A 0 20 40 60 80 100 distance from main pipe (s) Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 temperature (ºC) 150 condenser [O 2] (ppb) 200 A B C : to the sampling lime condenser outlet deaerator 0 5 without mixing temperature LPH 6 -0. 1 4 A deaerator -0. 2 3 ECP LPH 4 -0. 3 LPH 3 B 2 LPH 2 -0. 4 C 1 LPH 1 -0. 5 suitable interaction perfect mixture 0 -0. 6 50 100 150 200 0 time from condenser outlet (s) [O 2] (ppb) temperature (ºC) 250 [O 2] ECP temperature ECP (V-SHE) LPH 1 -4: 1 st-4 th low pressure heaters HPH 6: 6 th high pressure heaters A-C: sampling point S. Uchida

No. 19 Mass transfer coefficient calculation 3 D flow dynamics calculation, PLASHY 0º k-e

No. 19 Mass transfer coefficient calculation 3 D flow dynamics calculation, PLASHY 0º k-e calculation gave the almost same results as the LES calculation with lower mesh number and lower CPU time 90º 135º Pipe inner diameter: 50 cm Thickness of boundary layer: ~10 mm 180º mass transfer coefficient (m/s) Local calculation 3 3 D Local calculation 2 2 D Local calculation 1 PLASHY (a-FLOW) Whole system: 1 D RELAP 5 45º reattachment point orifice distance from inlet point (m) Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 20 Evaluation code system for FAC Calculation Flow pattern targets Input Computer programs

No. 20 Evaluation code system for FAC Calculation Flow pattern targets Input Computer programs [O 2], [Fe 2+] Wall thinning rate and ECP (anodic/cathodic (oxide film current density) formation) Reactor [O 2] , T, p. H, km, parameters: geometries, T, flow velocity (v), oxide film heat flux (Q), surface/volume rate, thickness, temperature (T) mixing rate properties icorr, 1 D CFD 2 -3 D k-e CFD 3 D LES Dynamic model (Oxide layer growth) N 2 H 4 -O 2 reaction code ECP Static model (Electro-chemistry) coupling calculation Output T, v distributions [O 2] and [Fe 2+] along distributions flow path along flow path icorr, ECP wall thinning rate oxide film thickness properties (Fe 2 O 3/Fe 3 O 4 ratio) Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Chemistry in surface boundary layers as related to flow accelerated corrosion of carbon steel in high temperature water”, ECS Transactions, 11 (27) 13 -26 (2008). S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

Schematic diagram of charge balance at surface oxide boundary film metal layer diffusion cathodic

Schematic diagram of charge balance at surface oxide boundary film metal layer diffusion cathodic reaction anodic reaction O 2 H+ H O 2 H 2 cathodic current e- M+ N 2 H 4 H+ eeanodic current a) Cathodic and anodic reactions 0. 5 potential (arbitrarily scale) bulk No. 21 total anodic current Fe Fe 2+ + eoxidation of H hydrazine 0 with oxide film without oxide film low [O 2] -0. 5 -1. 0 hydrogen generation potential L high [O 2] total cathdic current O 2 + e - O 2 10 -4 10 -3 10 -2 10 -1 current density (arbitrarily scale) b) Charge balance 100 Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 22 Modified double oxide layer model flow mass transfer dissolution release bulk water

No. 22 Modified double oxide layer model flow mass transfer dissolution release bulk water adsorption d boundary layer outer layer (hematite particles) base metal inner layer (magnetite particles) magnetite hematite particles oxidation Base Metal d. M/dt= -Ic Ferrous ion and oxide particle concentrations in the boundary layer d. C/dt=Ic/tb-dm. CTm 2/3 Cm 1/3 fmtb-dh. CTh 2/3 Ch 1/3 fhtb -kg. C/Csatfb(C)-k(C-Cb)+zm. Tm/t b+zh. Th/tb d. Cp/dt= kg. C/Csatfb(C)/Wm-kd. Cp-k(Cp-Cpb) Oxide layer (number of magnetite particles and thickness) d. Cm/dt= kg. C/Csat fb(C)/(Wm) -(c+km)Cm d. Tm/dt=dm. CTm 2/3 Cm 1/3 fmtb 2 +kg. C/Csat fb(C)tb-(zm+c+km)Tm+ F(ECP) Oxide layer (number of hematite particles and thickness) d. Ch/dt=c. Cm –kh. Ch d. Th/dt= c. Tm +dh. CTh 2/3 Ch 1/3 fhtb 2 -(zh +kh)Th Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

Calculated anodic polarization responses 101 d. V/dt=100 V/s 100 d. V/dt=1 V/s 10 -1

Calculated anodic polarization responses 101 d. V/dt=100 V/s 100 d. V/dt=1 V/s 10 -1 increasing potential 10 -2 d. V/dt=0. 01 V/s 10 -3 decreasing potential 10 -4 -1. 0 -0. 5 0 0. 5 potential (V-SHE) 1. 0 a) Potential increasing rate dependence current density (A/m 2) 101 100 No. 23 mass transfer coeff. , km= 0. 01 m/s increasing potential 10 -1 10 -2 10 -3 10 -4 -1. 0 km= 0. 002 m/s decreasing potential km= 0. 001 m/s -0. 5 0 0. 5 1. 0 potential (V-SHE) b) Mass transfer coefficient dependence Ref. S. Uchida, “Corrosion of Structural Materials and Electrochemistry in High Temperature Water of Nuclear Power Systems”, Power Plant Chemistry, 10, (11) (2008, ) in press. KAIST Seminar Nov. 25, 2008 S. Uchida

Results of wall thinning calculation 0. 4 101 0. 2 100 0 10 -1

Results of wall thinning calculation 0. 4 101 0. 2 100 0 10 -1 corrosion rate 10 -2 -0. 2 ECP (V-SHE) corrosion rate (mm/y) a) Mass transfer coefficient dependence 102 measured : Satoh, et al. ([O 2]<1 ppb) No. 24 -0. 4 ECP (V-SHE) corrosion rate (mm/y) c) p. H dependence-1: test specimens in a pipe -0. 6 10 -3 -4 -3 -2 -1 10 10 b) [O 2] dependence measured mass transfer coefficient, k (m/s) 1 : Heitmann and Schub 10 0. 4 101 ECP (k: 0. 005 m/s) k: 0. 05 m/s 0. 2 (k: 0. 001 m/s) 100 corrosion rate 0 -1 ([O ]: 0. 001 ppb) 10 0 2 corrosion rate (k: 0. 005 m/s) 10 -1 [O 2]: -0. 2 10 -2 -0. 2 -2 1 ppb 10 (k: 0. 001 m/s) -0. 4 [O 2]: 1 ppb -0. 4 10 -3 measured 10 -3 : Brush and Pearl ECP -0. 6 (p. H: 7, k: 0. 005 m/s) ([O ]: -4 -0. 6 2 0. 001 ppb) 10 10 -4 0 50 100 150 200 7. 5 8. 0 8. 5 9. 0 9. 5 10. 0 [O 2] (ppb) p. H (-) Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Corrosion Damage of Components in Cooling Systems of Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics (III), Evaluation of Pipe Wall Thinning Rate with the Coupled Model of Static Electrochemical Analysis and Dynamic Double Oxide Layer Analysis ”, J. Nucl. Sci. Technol. , 46 (1) (2009), in press. KAIST Seminar Nov. 25, 2008 S. Uchida

No. 25 Application of oxygen control to mitigate FAC 4 50 0 [O 2]

No. 25 Application of oxygen control to mitigate FAC 4 50 0 [O 2] (ppb) Temperature (ºC) -0. 1 4 200 50 [O 2] ECP temperature 0 5 250 100 1 [O 2] at condenser outlet: 5 ppb [N 2 H 4] at injection point: 100 ppb -0. 6 0 0 50 100 150 200 Time from condenser outlet (s) 0 150 2 FAC dangerous region 3 LPH 4 deaerator 2 1 -0. 2 -0. 3 -0. 4 -0. 5 -0. 6 0 0 50 100 150 200 Time from condenser outlet (s) Temperature (ºC) 100 3 N 2 H 4 injection point ECP (V-SHE) 150 T: LPH 6 -0. 1 180℃ -0. 2 LPH 4 deaerator -0. 3 LPH 3 ECP: -0. 4 LPH 2 -0. 3 V -0. 5 LPH 1 250 5 200 4 150 100 50 0 3 2 1 LPH 6 LPH 4 deaerator 0 -0. 1 -0. 2 -0. 3 -0. 4 -0. 5 ECP (V-SHE) 200 0 [O 2] (ppb) 5 ECP (V-SHE) 250 [O 2] (ppb) Temperature (ºC) in secondary cooling system at PWR -0. 6 0 0 50 100 150 200 Time from condenser outlet (s) Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 26 Evaluation of countermeasures Ia, Ic (A/m 2) Flow dynamics control ( k:

No. 26 Evaluation of countermeasures Ia, Ic (A/m 2) Flow dynamics control ( k: 0. 004 m/s) FAC rate ( k: 0. 002 m/s) 10 -2 ECP 0. 4 -. 2 10 -4 0 10 -6 -0. 2 10 -8 -0. 4 V (volt) ECP (V-SHE) FAC rate (mm/y) 100 1. 00 E+02 ECP 1. 00 E+01 1. 00 E+00 without oxide layer 1. 00 E-01 1. 00 E-02 with 1. 00 E-03 oxide layer 1. 00 E-04 1. 00 E-05 current 1. 00 E-06 (FAC rate) 1. 00 E-07 -0. 80 -0. 60 -0. 40 -0. 20 0. 00 0. 20 0. 40 b) FA rate and ECP with/without oxide layer -0. 6 0. 2 0. 4 0. 6 0. 8 1 [Fe 2+] Cb/Csat (-) a) Relationship of [Fe 2+], ECP and FAC rate 10 -10 0 Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 27 Table of contents 1. Nuclear power plants in the world 2. Interaction

No. 27 Table of contents 1. Nuclear power plants in the world 2. Interaction of cooling water and materials Major roles of cooling water in nuclear power plants and its adverse effects 3. Flow accelerated corrosion (Wall thinning)  4. Stress corrosion cracking (IGSCC and PWSCC) 5. Corrosion of zirconium alloy (Corrosion of fuel cladding) 6. Future subjects 7. Conclusion KAIST Seminar Nov. 25, 2008 S. Uchida

Theoretical determination of corrosive conditions of BWRs No. 28 based on water radiolysis model

Theoretical determination of corrosive conditions of BWRs No. 28 based on water radiolysis model steam flow steam water flow release of gaseous species H 2, O 2 decomposition of hydrogen peroxide 2 H 2 O 2→ 2 H 2 O + O 2 upper plenum feed water reactor core recirculation water radiolysis 2 H 2 O→ H 2 + H 2 O 2 down comer lower plenum cooled down recombination reactions 2 H 2 + O 2→ 2 H 2 O H 2 + H 2 O 2→ 2 H 2 O sampled water to determine oxidant concentrations Ref. E. Ibe. and S. Uchida, “Radiolytic Aspects in Boiling Water Reactor Primary Systems: Results from Numerical Simulation and Statistical Regression Analyses”, Nucl. Sci. Eng. , 89, 330 (1985). KAIST Seminar Nov. 25, 2008 S. Uchida

No. 29 Radiolysis Model for BWR primary circuit Basic equations and procedures for their

No. 29 Radiolysis Model for BWR primary circuit Basic equations and procedures for their solutions ∂Yi/∂t change in concentrations = giγQg + gin. Qn   direct generation [water radiolysis] +Σkimn. Ym. Yn  – YiΣkis. Ys   2 ndary generation disappearance + Vf/(1 -Vf)(ei*Ygi-ei. Yli) transfer between liquid and vapour Yi : concentration G: g-values   g ; g rays   n ; neutrons Q: energy absorption rate K: reaction constants Vf: void ratio e : transfer coefficient Numerical solution[Stiff equations] 1) Integration of the equations along with crack depth. 2) Backward differential formula (BDF) for initial guess and Newton method for stable solution Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 30 Maps of distribution of [O 2]eff in RPV (Effects of hydrogen injection

No. 30 Maps of distribution of [O 2]eff in RPV (Effects of hydrogen injection on suppression of [O 2]) [O 2]eff (ppb) 200 50 20 0 NWC HWC [H 2]eff : 50 ppb Ref. S. Uchida, M. Naitoh, Y. Uehara, H. Okada, S. Koshizuka and D. H. Lister, “Evaluation Methods for Flow Accelerated Corrosion in Nuclear Power Plants by Coupling Analysis of Corrosion and Flow Dynamics”, Proc. Int. Conf. on Flow Accelerated Corrosion 2008, EDF, Lyon, France, Mar. 18 -20 (2008). (DVD) KAIST Seminar Nov. 25, 2008 S. Uchida

Effects of H 2 injection in BWR plants No. 31 0. 2 200 0

Effects of H 2 injection in BWR plants No. 31 0. 2 200 0 -0. 2 -0. 4 -0. 6 6 5 ECP NMCA ECP 150 100 [O 2]eff 50 0 0 20 optimal [H 2]RW 4 MS dose rate MSDR limit [O 2]eff target 40 60 [H 2]RW (ppb) a) Measured in plants 80 3 2 1 0 102 [O 2]eff (ppb) 250 main steam line dose rate 0. 4 [O 2]eff (ppb) ECP (V-SHE) [O 2]eff= [O 2]+1/2 [H 2 O 2]: effective oxygen concentration [H 2]RW : [H 2] in the reactor water 101 100 0 [O 2]eff 20 40 60 80 100 [H 2]RW (ppb) b) Calculated for plants Ref. H. Takiguchi, M. Sekiguch, A. Abe, K. Akamine, M. Sakai, Y. Wada and S. Uchida, “Evaluation of Effectiveness of Hydrogen Water Chemistry for Different Types of Boiling Water Reactors”, J. Nucl. Sci. Technol. , 36, 179 (1999). KAIST Seminar Nov. 25, 2008 S. Uchida

No. 32 A model for crack growth rate oxide film bulk water anion OH-,

No. 32 A model for crack growth rate oxide film bulk water anion OH-, SO 42 crack mouth Cation Ni 2+, Fe 2+ release 2+ 2+ Ni , Fe crack tip lower potential due to bare surface、Ect higher potential due to oxide film, Ecm (measured potential) crack growth rate release rate Ecm-Ect Ref. P. L. Andresen, Y. J. Kim, T. P. Diaz and S. Hettiarachchi, “Mitigation of SCC by on-line NOBLECHEM”, Proc. 13 th Int Conf. Environmental Degradation of Materials in Nuclear Power Systems – Water Reactors, Whistler, BC, Canada, Aug. 19 -23, 2007, CNS, (2007) (CD-ROM). KAIST Seminar Nov. 25, 2008 S. Uchida

Temperature dependent Co release rate from stainless steel temperature (ºC) 270 240 200 170

Temperature dependent Co release rate from stainless steel temperature (ºC) 270 240 200 170 10 -2 240ºC 250ºC 200ºC 270ºC 10 -3 150ºC 10 -4 20 50 100 200 500 1000 exposure time (h) a) Co release rate 5000 oxide film thickness (arbitrary unit) 10 -1 cobalt release rate (g/m 2/month) No. 33 104 inner oxide layer (direct oxidation) 103 outer oxide layer (precipitation of oxide) 102 101 100 1. 6 2. 0 2. 2 2. 4 2. 6 1/T (x 10 -3 K-1) b) Thickness of oxide film 1. 8 Ref. M. Kitamura, Y. Ozawa and S. Uchida, “Temprature Dependence of Cobalt Release Rate from Stainless Steel in Neutral Water”, Proceedings of the International Symposium on Environmental Degradation of Materials in Nuclear Power Systems, Water Reactors, Myrtle Beach, SC. , Aug. 22 -25, 1983, NACE National As, ISBN 0 -91666, 623 -635(1984). KAIST Seminar Nov. 25, 2008 S. Uchida

No. 34 Schematic diagram of relationship 1 1 dissolution larger stress intensity factor crack

No. 34 Schematic diagram of relationship 1 1 dissolution larger stress intensity factor crack depth 0 exposure time (day) oxide film thickness (arbitrary unit) crack depth (arbitrary unit) dissolution rate (arbitrary unit) between crack growth rate and oxide film thickness at crack tip film thickness 0 Ref. S. Uchida, “Corrosion of Structural Materials and Electrochemistry in High Temperature Water of Nuclear Power Systems”, Power Plant Chemistry, 10, (11) (2008, ) in press. KAIST Seminar Nov. 25, 2008 S. Uchida

No. 35 Crack growth rate as a function of ECP crack growth rate (mm/s)

No. 35 Crack growth rate as a function of ECP crack growth rate (mm/s) 10 -5 Type 304 stainless steel (25 mm CT specimen) furnace sensitized: 15 C/cm 2 water temperature: 288 C constant load: 27. 5 MPa(m)1/2 10 -6 measured calculated 10 -7 10 -8 Conductivity (m. S/cm) : 0. 3 0. 2 0. 1 HWC 10 -9 -600 -400 -200 NWC 0 200 400 NWC: normal water chemistry (without hydrogen injection) HWC: hydrogen water chemistry (with hydrogen injection) ECP (m. V-SHE) Ref. F. P. Ford and P. L. Andresen, “Corrosion in Nuclear Systems: Environmentally Assisted Cracking in Kight Water Reactors” in Corrosion Mechanisms, Ed. P. Marcus and J. Ouder, Marcel Dekker, 501 -546 (1994). KAIST Seminar Nov. 25, 2008 S. Uchida

No. 36 Effects of hydrogen X-750, 360 C, 49 MPam 1/2 1. 0 Ni.

No. 36 Effects of hydrogen X-750, 360 C, 49 MPam 1/2 1. 0 Ni. O Ni metal 0 0 50 100 3 [H 2] (Ncm /kg) a) Crack growth rate 150 25 Ni/Ni. O line 2. 0 crack initiation time (kh) crack growth rate(mills/day) on IGSCC crack initiation and crack growth rate 20 15 10 tube diameter 3/4 inches 7/8 inches 5 0 0 5 10 15 20 25 30 Nml-H 2/kg-H 2 O (at 330 C) b) Crack initiation time 35 Ref. L. Wilson and J. Hickling, “Use of Primary Water chemistry in Pressurized Water Reactors to Mitigate PWSCC in Nickel Base Alloys”, Proc. International Conference on Water Chemistry of Nuclear Reactor Systems, 2006, Oct. 23 -26, 2006, Jeju Island, Korea, KAERI (2006). (in CD-ROM) A. Molander, A. Jenssen, M. Konig and K. Norring, “PWSCC Initiation and Crack Growth Data for Alloy 600 with Focus on Hydrogen Effects”, Proc. International Workshop on Optimization of Dissolved Hydrogen Content in PWR Primary Coolant, July 18 -19, 2008, Sendai, Japan (2007). KAIST Seminar Nov. 25, 2008 S. Uchida

No. 37 High Temperature G-values species PWR(305ºC) BWR(285ºC) g rays neutrons a rays 0.

No. 37 High Temperature G-values species PWR(305ºC) BWR(285ºC) g rays neutrons a rays 0. 60 H 3. 50 0. 90 H+ H 2 O 2 e- 0. 662 0. 453 0. 152 0. 50 3. 565 0. 927 3. 50 0. 60 1. 50 0. 612 3. 565 1. 278 0. 662 1. 974 0. 152 1. 14 0. 542 0. 836 1. 104 HO 2 0. 55 0. 00 0. 04 0. 000 0. 050 0. 300 OH OH- 4. 50 0. 00 1. 70 0. 00 4. 632 0. 000 1. 849 0. 000 1. 191 0. 000 0. 199 molecules, atoms /100 ev absorption Ref. H. Takiguchi, M. Ullberg and S. Uchida “Optimization of Dissolved Hydrogen Concentration for Control of Primary Coolant Radiolysis in Pressurized Water reactors”, J. Nucl. Sci. Technol. , 41, 601 (2004) KAIST Seminar Nov. 25, 2008 S. Uchida

Calculated results of PWR radiolysis model Comparison with INCA loop experiments 0 DH (Ncm

Calculated results of PWR radiolysis model Comparison with INCA loop experiments 0 DH (Ncm 3 kg-1) 4 6 8 10 2 14 10 -5 H 2 OH OH H O 2 HO 2 eaq- 10 -7 10 -9 10 -11 10 -13 10 -15 0 1 12 10 -5 10 -7 10 -9 10 -11 0 1 Comparison of calculated [Ox]/[Red] ratio and measured ECP(1600 ppm B and 2 ppm Li) 2 3 4 -6 x 10 [H 2] (mol/l) In-core region DH (Ncm 3 kg-1) (1600 ppm B and 2 ppm Li) 0 2 4 6 8 10 10 -3 concentrations (mol/l) 12 2 3 4 -6 X 10 [H 2] (mol/l) 14 250 150 50 -150 -250 -350 -450 DH (Ncm 3/kg) 5 10 15 0 20 calculated by the ratio of oxide species to redox species ECP (SHE) ECP (m. V SHE) concentrations (mol/l) Out-of-core region (1600 ppm B and 2 ppm Li) 10 -3 No. 38 ECP= 350·log ([Ox]/ [Red]) – 100 0 1 2 3 4 5 -6 x 10 [H 2] (mol/l) 6 Ref. H. Takiguchi, M. Ullberg and S. Uchida “Optimization of Dissolved Hydrogen Concentration for Control of Primary Coolant Radiolysis in Pressurized Water reactors”, J. Nucl. Sci. Technol. , 41, 601 (2004) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 39 Table of contents 1. Nuclear power plants in the world 2. Interaction

No. 39 Table of contents 1. Nuclear power plants in the world 2. Interaction of cooling water and materials Major roles of cooling water in nuclear power plants and its adverse effects 3. Flow accelerated corrosion (Wall thinning)  4. Stress corrosion cracking (IGSCC and PWSCC) 5. Corrosion of zirconium alloy (Corrosion of fuel cladding) 6. Future subjects 7. Conclusion KAIST Seminar Nov. 25, 2008 S. Uchida

No. 40 40 compression (MPa) 16 20 10 200 100 0 ratio of 14

No. 40 40 compression (MPa) 16 20 10 200 100 0 ratio of 14 cubic 12 10 20 30 40 50 60 70 80 90 100 exposure time (days) stress in horizontal direction -200 60 50 40 30 0 -100 20 18 30 0 Compression and weight gain 50 22 ratio of cubic (%) 60 weight gain 20 10 weight gain (mg/dm 2) Weight gain and cubic ratio weight gain (mg/dm 2) Properties of Zr. O 2 on zirconium alloy 0 0 10 20 30 40 50 60 70 80 90 100 exposure time (days) Ref. J. Godlewski and R. Cadalbert, “A New Method of Residual Stress Distribution Analysis for Corroded Zircaloy-4 Cladding”, Proc. Int. Symp. On Material Chemistry in Nuclear Environment, Tsukuba, Japan (1992). KAIST Seminar Nov. 25, 2008 S. Uchida

Depth profile of 18 O and total oxygen No. 41 before and after immersing

Depth profile of 18 O and total oxygen No. 41 before and after immersing in 288 C H 218 O for 100 h 106 before after 16 O+18 O counts (-) 105 18 O 104 growth 103 102 18 O/(16 O+18 O) (-) 101 1. 0 before after 0. 8 18 O (16 O+18 O) growth 0. 6 0. 4 0. 2 0 0 0. 2 0. 4 0. 6 0. 8 1. 0 1. 2 1. 4 1. 6 depth (mm) Ref. Y. Nishino, T. Yasuda, E. Iba and M. Endo, “Effects of Water radiolysis on Zircaloy Corrosion”, Proc. 1997 Int. Topical Meeting on LWR Fuel Performance, 232 -239, Portland, Oregon, Mar. 2 -6, 1997, American Nuclear Society (1997). KAIST Seminar Nov. 25, 2008 S. Uchida

Distribution of corrosive radical concentrations No. 42 [ calculated with radiolysis model ] 400

Distribution of corrosive radical concentrations No. 42 [ calculated with radiolysis model ] 400 spacers [zyrcaloy-2/ nickel alloy] O 2 HO 2 OH distance from core inlet (cm) 350 fuel cladding [zyrcaloy-2] 300 250 200 150 region with maximum effects of radicals 100 50 0 fuel assembly 0 1 2 3 4 5 concentration (ppb) 6 Ref. Y. Nishino, M. Endo, Y. Wada and T. Yasuda, “Interaction between Zircaloy-2 Oxide Film and Hiogh temperature Water under BWR Conditions”, Proc. 1998 JAIF Int. Conf. on Water Chemistry of Nuclear power Systems, 780, Oct. 13 -16, 1998, Kashiwazaki, Japan, JAIF (1998). KAIST Seminar Nov. 25, 2008 S. Uchida

Weight gain of zircaloy-2 weight gain (mg/dm 2) 100 No. 43 water steam super

Weight gain of zircaloy-2 weight gain (mg/dm 2) 100 No. 43 water steam super critical water 10 1 1. 2 1. 6 1/tempertaure (K-1) x 103 2. 0 Ref. H. Karasawa, K. Ishida, Y. Wada, M. Endou, Y. Nishina, M. Aizawa, M. Fuse, E. Kadoi and H. Takiguchi, “Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (III) Effects of Adding Hydrazine on Zircaloy-2 Corrosion ”, J. Nucl. Sci. Technol. , 43, 1218 (2006) KAIST Seminar Nov. 25, 2008 S. Uchida

Weight gain of the specimens in 450ºC No. 44 supercritical water with or without

Weight gain of the specimens in 450ºC No. 44 supercritical water with or without g-ray irradiation Ref. H. Karasawa, K. Ishida, Y. Wada, M. Endou, Y. Nishina, M. Aizawa, M. Fuse, E. Kadoi and H. Takiguchi, “Hydrazine and Hydrogen Co-injection to Mitigate Stress Corrosion Cracking of Structural Materials in Boiling Water Reactors, (III) Effects of Adding Hydrazine on Zircaloy-2 Corrosion ”, J. Nucl. Sci. Technol. , 43, 1218 (2006) KAIST Seminar Nov. 25, 2008 S. Uchida

No. 45 Table of contents 1. Nuclear power plants in the world 2. Interaction

No. 45 Table of contents 1. Nuclear power plants in the world 2. Interaction of cooling water and materials Major roles of cooling water in nuclear power plants and its adverse effects 3. Flow accelerated corrosion (Wall thinning)  4. Stress corrosion cracking (IGSCC and PWSCC) 5. Corrosion of zirconium alloy (Corrosion of fuel cladding) 6. Future subjects 7. Conclusion KAIST Seminar Nov. 25, 2008 S. Uchida

No. 46 Future subjects Flow accelerated corrosion (Wall thinning)  1. Mass transfer coefficient One

No. 46 Future subjects Flow accelerated corrosion (Wall thinning)  1. Mass transfer coefficient One of the key parameters for flow dynamics factors to determine FAC is the mass transfer coefficient around the attacked zone. Precise evaluation of flow pattern is required to determine the mass transfer coefficients. 2. Surface roughness The problem is mass transfer coefficients are determined not only by flow distribution in the liquid side but also by surface roughness, which varies as a result of FAC. 3. Ferrous ion concentration FAC rate is also affected by ferrous ion concentration in the water, which is also determined as a result of FAC upstream from the zone of interest. 4. Coupling analysis of flow dynamics and corrosion These facts require that FAC be evaluated in power plant by tight coupling analysis of flow dynamics and corrosion, especially electrochemical, models. KAIST Seminar Nov. 25, 2008 S. Uchida

No. 47 Future subjects Stress corrosion cracking Intergranular stress corrosion cracking (IGSCC) 1. Environmental

No. 47 Future subjects Stress corrosion cracking Intergranular stress corrosion cracking (IGSCC) 1. Environmental factors Procedures for water radiolysis analysis should be authorized as standard procedures to evaluate IGSCC mitigation. 2. Stress factors The stress factors are difficult to measure and apply to analysis of their effects on oxide film rupture. Precise evaluation and measurement of the stress distribution in structural materials are key issues to understand IGSCC behaviors. Dynamic stress might result in more effects on oxide film rupture than static stress. Primary water stress corrosion cracking (PWSCC) 1. The effects of hydrogen on PWSCC Procedures for water radiolysis analysis should be authorized as standard procedures to evaluate PWSCC mitigation. 2. Irradiation effects Irradiation of materials might enhance hydrogen acceptance due to distributed point defects in materials and also might enhance hydrogen injection due to increased proton recoil at the material surface. KAIST Seminar Nov. 25, 2008 S. Uchida

No. 48 Future subjects Corrosion of zirconium alloy 1. Increasing importance of corrosion of

No. 48 Future subjects Corrosion of zirconium alloy 1. Increasing importance of corrosion of cladding materials Increasing fuel burn-up, extension of the fuel cycle and raising the power rate are much related to fuel performance. 2. Development of cladding materials The development has been based on in-pile experiments and demonstration tests in operating plants over long development periods. It is desirable to carry out parametric experiments for corrosion properties in out-of-pile experiments and then confirm the abilities in in-pile experiments. Methodology for an acceleration test for zirconium alloy is one of the key issues. Application of a supercritical water loop for corrosion testing is one promising candidates for the acceleration test. KAIST Seminar Nov. 25, 2008 S. Uchida

No. 49 Table of contents 1. Nuclear power plants in the world 2. Interaction

No. 49 Table of contents 1. Nuclear power plants in the world 2. Interaction of cooling water and materials Major roles of cooling water in nuclear power plants and its adverse effects 3. Flow accelerated corrosion (Wall thinning)  4. Stress corrosion cracking (IGSCC and PWSCC) 5. Corrosion of zirconium alloy (Corrosion of fuel cladding) 6. Future subjects 7. Conclusion KAIST Seminar Nov. 25, 2008 S. Uchida

No. 50 Conclusions Problems related to corrosion experienced in nuclear power plants were reviewed.

No. 50 Conclusions Problems related to corrosion experienced in nuclear power plants were reviewed. Corrosion behaviors of structural materials were categorized into three groups, i. e. , carbon steel (FAC), stainless steel (IGSCC) and nickel alloy (PWSCC), and zirconium alloy, based on relative corrosion rates. Corrosion behaviors are fundamentally understood as electrochemical reactions but they are determined not only electrochemical reactions but also behaviors of oxide film, which is one of the products of corrosion and at the same time one of the control barriers for corrosion itself. As a result of dividing each corrosion behavior into two, electrochemical reactions and oxide film rupture and reformation, fundamental phenomena determining the process can be more easily understood and simulation for the process can be easily prepared. Most of phenomena of corrosion related problems, can be understood based on the electrochemical index, e. g. , ECP, conductivities and p. H. Major electrochemical index, ECP, can be measured at elevated temperature and applied as in situ sensor of corrosion conditions to detect anomaly conditions of structural materials and material abnormalities at their very early stages. KAIST Seminar Nov. 25, 2008 S. Uchida

No. 51 Thank you for your kind attention. KAIST Seminar Nov. 25, 2008 S.

No. 51 Thank you for your kind attention. KAIST Seminar Nov. 25, 2008 S. Uchida