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0 Research on Corrosion Fatigue of Structure Materials in LWR Environment XU Xuelian SHI Xiuqiang DING Yaping Shanghai Nuclear Engineering Research and Design Institute Workshop on FAC/EAC in NPPs 21-23 April 2009, Moscow, Russian Federation

Research on Corrosion Fatigue of Structure … Research on Corrosion Fatigue of Structure Materials in LWR Environment XU Xuelian SHI Xiuqiang DING Yaping Shanghai Nuclear Engineering

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Page 1: Research on Corrosion Fatigue of Structure … Research on Corrosion Fatigue of Structure Materials in LWR Environment XU Xuelian SHI Xiuqiang DING Yaping Shanghai Nuclear Engineering

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Research on Corrosion Fatigue of Structure Materials in LWR Environment

XU Xuelian SHI Xiuqiang DING YapingShanghai Nuclear Engineering Research and Design Institute

Workshop on FAC/EAC in NPPs21-23 April 2009, Moscow, Russian Federation

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上海核工程研究设计院 SNERDI ----1111----

Contents� Introduction �Experimental Specimen and Apparatus �Experiment and Results

� Austenitic Stainless Steel F316Ti� Low Alloy Steel� Carbon Steel�Conclusions

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上海核工程研究设计院 SNERDI

Introduction� Environment effect on fatigue performance of materials used for pressurized boundary, including fatigue life and fatigue crack growth rate, are of importance to nuclear safety. The fatigue design of pressurized boundary components of light water reactor (LWR) is in accordance with the ASME Boiler and Pressure Vessel Code Section .� It is well known that the design fatigue curves adopted in the ASME Code are those that have been determined by modifying the pertinent best-fit curves, which were obtained through the laboratory test performed in ambient atmospheric air, with a factor of 2 for the stress and the factor of 20 for the cyclic life.

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上海核工程研究设计院 SNERDI

Introduction

� The factor 20 to be applied to the cyclic life comprises a factor of 2.0 for the scatter of data, 2.5 for the size effect, and 4.0 for the effects of the surface finish, testing environment, and some others.� It should be noted that, the environment mentioned therein related to the environment of workshop or the work site is distinguished from the controlled environment of laboratory, but does not refer to the effects of corrosive medium.

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上海核工程研究设计院 SNERDI

Introduction

� This research is undertaken on low cycle corrosion fatigue performance of structure material manufactured in China in boiling water reactor (BWR) and pressure water reactor (PWR) primary environments.� The purposes is to predicting the fatigue

life of nuclear materials and improving the design of nuclear materials.

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上海核工程研究设计院 SNERDI

Experimental Materials

� Three kinds of structure material manufactured in China had been selected for the low cycle corrosion fatigue performance investigation� Austenitic Stainless Steel F316Ti� Low alloy Steel SA508 Class 3� Carbon Steel TU48

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上海核工程研究设计院 SNERDI

Experimental Specimen

Schematic Diagram of the Specimen

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上海核工程研究设计院 SNERDI

Apparatus

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上海核工程研究设计院 SNERDI

Austenitic Stainless Steel

Two kinds of testing material were used, one was named as F316Ti(S) and another was named as F316Ti(F) which were manufactured in deferent factories of China.

0.040.632.3413.1718.350.0220.00150.631.760.072F316Ti(F)0.040.492.1412.1017.160.0220.0140.711.450.044F316Ti(S)

//2.00~~~~

3.00

10.00~~~~

14.00

16.00~~~~

18.00≤≤≤≤

0.040≤≤≤≤

0.030≤≤≤≤

1.00≤≤≤≤

2.00≤≤≤≤

0.080SA336----F316

CoTiMoNiCrPSSiMnCMaterial

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上海核工程研究设计院 SNERDI

Test Condition for Austenitic Stainless Steel

/Li 2ppm<0.1μμμμs/cm/<0.2μμμμs/cmConductivity

/B 1200ppm6.2~~~~6.5/6.2~~~~6.5pH

/≤≤≤≤5ppb≤≤≤≤100ppb/≤≤≤≤100ppbDO/11.76MPa7.8MPa/9.2MPaPressure

320℃℃℃℃320℃℃℃℃288℃℃℃℃300℃℃℃℃300℃℃℃℃Temperature

Water Chemistry

StrainStrainStrainStrainStrainControl Mode0.1%S-10.1%S-10.1%S-10.5~~~~

0.8%S-10.1%S-1Strain Rate

0.5~~~~1.5%0.6~~~~1.5%0.8~~~~1.6%0.6~~~~1.5%0.6~~~~1.5%Strain Range

Triangle Wave

Triangle Wave

Triangle Wave

Triangle Wave

Triangle WaveWave Form

AIRPWRBWRAIRBWRConditionsF316Ti(F)F316Ti(S)Material

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上海核工程研究设计院 SNERDI

Austenitic Stainless Steel

Fatigue data for F316Ti in simulated LWR primary environment and high temperature air

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上海核工程研究设计院 SNERDI

Austenitic Stainless Steel

Fracture surface of F316Ti(F) in PWR environment (∆εt=0.6%) 500××××a. initiation of cracking b.striation and the second cracking

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上海核工程研究设计院 SNERDI

Low Alloy Steel

� The material used in this work was 508-III(equivalent to SA508class3) manufactured in China.

0.030.0020.0060.480.120.740.170.0020.0031.410.19508-III

≤≤≤≤

0.10≤≤≤≤

0.05/0.45~0.60

≤≤≤≤

0.250.40~1.00

0.15~0.40

≤≤≤≤

0.018≤≤≤≤

0.0151.20~1.50

≤≤≤≤

0.25SA508

CuVCoMoCrNiSiSPMnCMaterial

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上海核工程研究设计院 SNERDI

Test Condition for Low Alloy Steel

/Li+ 2ppm/B 1200ppm/≤≤≤≤5ppbDO/9.2MPaPressure

300℃℃℃℃300℃℃℃℃Temperature

Water Chemistry

StrainStrainControl Mode0.1~0.6%S-10.1%S-1Strain Rate0.4~2.4%0.5~1.8%Strain Range

Triangle WaveTriangle WaveWave FormAIRPWRConditions

508-IIIMaterial

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上海核工程研究设计院 SNERDI

Low Alloy Steel

Fatigue data for 508-III in simulated PWR primary environment and high temperature air

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上海核工程研究设计院 SNERDI

Low Alloy Steel

Fracture surface of 508-III in PWR environment (∆εt=0.5%) 500××××

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上海核工程研究设计院 SNERDI

Carbon Steel

� A kind of carbon steel TU 48 manufactured in China was chosen for the test. The carbon steel sample was taken from middle wall thickness of the pipe as received.

0.0120.150.070.120.140.0130.0050.251.290.17TU48

SnCuMoNiCrPSSiMnCMaterial

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上海核工程研究设计院 SNERDI

Test Condition for Carbon Steel

/5ppb, 100ppb, 200ppb,400ppb, 8ppmDO

/9.2MPaPressure300℃℃℃℃300℃℃℃℃TemperatureStrainStrainControl Mode0.1%S-10.1%S-1Strain Rate

0.5%~2.2%0.5%~2.2%Strain RangeTriangle WaveTriangle WaveWave Form

AIRWaterConditionsCarbon SteelMaterial

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上海核工程研究设计院 SNERDI

Carbon Steel

Fatigue data for carbon steel in high temperature water and high temperature air

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上海核工程研究设计院 SNERDI

Carbon Steel

Fracture surface of carbon steel in high temperature water (∆εt=0.7%)a. initiation of cracking 50×××× b. second cracking 500××××

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上海核工程研究设计院 SNERDI

Carbon Steel

The effect of dissolved oxygen on fatigue life

1.00E+03

1.00E+04

1.00E+00 1.00E+01 1.00E+02 1.00E+03 1.00E+04

Dissolved oxygen / ppb

Fatig

ue lif

e / cy

cles

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上海核工程研究设计院 SNERDI

Carbon Steel

a. DO=5ppb,,,,50×××× b. DO=8ppm,,,,50××××

The effect of dissolved oxygen on pitting

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上海核工程研究设计院 SNERDI

Carbon Steel

The effect of dissolved oxygen on cracking, ΔεΔεΔεΔεt=0.7%,,,,500××××a. DO=5ppb b. DO=400ppb c. DO=8ppm

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上海核工程研究设计院 SNERDI

Conclusions � The ambient environment is one of the most

important effective factors on corrosion fatigue properties. The low cycle corrosion fatigue life for the same batch of materials in high temperature air is longer than that in simulated LWR primary environment.

� In simulated BWR and PWR environment, the fatigue property has no obvious difference in high strain range. With the decrease of strain amplitude, the difference occurred gradually.

� All of the test data are scattered around ASME best-fit curve and above design fatigue curve.

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上海核工程研究设计院 SNERDI

Conclusions� The characteristic of stress corrosion cracking

(SCC) and corrosion fatigue (CF) can be observed obviously on all of the fracture surface.

� Dissolved oxygen has effect on corrosion fatigue of carbon steel. In the condition of high dissolved oxygen, the fatigue life of specimen decreased significantly. Higher dissolved oxygen could cause more second cracking on the fracture surface.

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上海核工程研究设计院 SNERDI

Thanks!