1. Study on Radiation Induced Ageing of Power Reactor Components S. Chatterjee, K.S. Balakrishnan, Priti Kotak Shah, D.N. Sah and Suparna Banerjee Post Irradiation Examination Division Bhabha Atomic Research Centre Trombay, Mumbai, India

2. Why to evaluate radiation damage in reactor structural Materials What life limiting structural materials were evaluated How to enhance the expertise for estimation of residual life/extension of life of components

3. Why to evaluate radiation damage in reactor structural Materials ?

4. Commercial Reactors Pressurised Heavy Water Reactor (PHWR) Boiling Water Reactor (BWR) Water Water Energy Reactor (WWER) Research Reactors CIRUS DHRUVA

5. Structural Materials/ Components Zr-alloys fuel cladding: Zr-2/Zr-4 pressure tube: Zr-2/Zr-2.5Nb calandria tube: Zr-2 garter spring:Zr-0.5Cu-2.5Nb Steels end fitting: 403 SS end shield: 203D/304 SS pressure vessel: 302B-Ni modified (A533B) WWER 1000

6. Components experience aggressive environment of : Temperature Stress Corrosion Radiation damage Primary radiation damage is from neutron population

7. Neutron Radiation Damage leads to changes in dimension (creep and growth) changes in mechanical properties  increase in yield strength and tensile strength  decrease in ductility  decrease in fracture toughness  increase in ductile-brittle transition temperature  increase in delayed hydride cracking velocity and also changes in microstructure and chemical composition One/ some of these changes may become life limiting for components

8. End-Of-Life (EOL) fluence of components Componentn-fluence (>1 MeV) dpaTime (years) Fuel cladding2*10 21 4.416 Pressure tube2*10 22 4416 Calandria tube2*10 22 4416 Garter spring2*10 22 4416 End fitting6*10 19 0.131 End shield5*10 19 0.111 TAPS RPV3.3*10 18 0.0071/12 WWER 1000 RPV3.7*10 19 0.081 Saturation fluence : 1*10 21 n/cm 2 (>1MeV), 2.2 dpa Threshold fluence : 5*10 17 n/cm 2 (>1MeV),

9. What life limiting structural materials were evaluated ?

10. ComponentOrigin of specimens Fuel Cladding Pressure Tube Garter Spring End Fitting Calandria Tube TAPS RPV Operating reactor Research Reactor Operating reactor Components Evaluated

11. Type of TestComponents Tension Impact Fracture Toughness Crush Test Irradiation Growth Delayed Hydride Crack(DHC) Pressure Vessel, Cladding, Garter Spring, End-fitting Pressure Vessel, End-fitting PressureVessel,Pressure Tube Garter Spring Calandria Tube Pressure Tube Types of Tests Conducted

12. ComponentProperty Fuel Cladding Pressure Tube Garter Spring End Fitting Calandria Tube TAPS RPV Ductility Fracture Toughness, DHC Crush Strength DBTT Irradiation Growth DBTT (Fracture toughness) Life Limiting Phenomenon/Property

13. Tensile Property of Claddings Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

14. ReactorPTEFPY MAPS-2N – 104.85 MAPS-1P – 136.25 RAPS-2K – 078.25 MAPS-1J - 079.5 Pressure Tubes Evaluated Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

15. CCL for various PTs Evaluated Safe Unsafe Safe Unsafe Equivalent hydrogen content (ppm) 40 80 120 160 200 Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

16. MaterialTemperature ( 0 C) DHC velocity (mm/h) Zr-2250 290 0.07 0.12 Zr-2.5Nb250 290 0.29 0.52 DHCV irr, Zr-2 = DHCV unirr, Zr-2 X 5 DHCV irr, Zr-2.5Nb = DHCV unirr, Zr-2.5Nb X 3 DHCV measurement on Zirconium alloys Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

17. Garter Springs Evaluated Spring Identification Reactor EFPYNumbers Examined ( type of test ) K-07 RAPS-2 8.26 1 ( tension, stretch, crush tests) O-11 RAPS-2 6.50 1 ( tension, stretch, crush tests) F-10 RAPS-1 3.60 2 (stretch test) N-10 MAPS-2 4.80 1 ( stretch, crush tests) K-14 MAPS-2 3.60 1 ( stretch, crush tests) K-19 NAPS-1 1.80 1 ( stretch, crush tests) Not Identified RAPS-2 8.50 14 ( stretch test) Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

18. Room Temperature Crush Test Results Spring Identification (Reactor, EPFY) Location of G.S. piece Maximum Load applied (N/coil)* Remarks** K-7(RAPS–2,8.26)6 O’ clock728 a O-11(RAPS-2, 6.5)6 O’ clock 845b N-10(MAPS-2,4.8)6 O’ clock 539b K-14(MAPS-2,3.6)6 O’ clock 410b K-19(NAPS-1,1.8)6 O’ clock 428b * Load values depicted are typically one order more in magnitude than the design load ** a: Specimen got crushed, b: Gap got closed Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

19. Irradiation details SpecimensCharpy V-notch Location Tray rod location in CIRUS reactor Neutron Flux 2.4 x 10 12 n.cm -2.S -1 E > 1.0 Mev Duration of Irradiation48 Days at Full Power Neutron Fluence 1 x 10 19 n.cm -2 E > 1.0 Mev Irradiation Temperature290º C±10º C Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

20. Temperature Un -Irradiated Irradiated 1 X 10 19 n/cm 2, >1 MeV Δ USE Δ T=75 0 C Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel = 44J Energy At EOL fluence of 6 X 10 19 n/cm 2, >1 MeV Δ T EOL = 75 X ( 6 X 10 19 / 1 X 10 19 ) 0.33 =136 0 C RT NDT,EOL = 170 0 C Operating temperature : 250 0 C, 300 0 C

21. SpecimenGrowth Strain (10 -4 ) Seamless Longitudinal4.70 Seam welded Long.4.78 Seamless Transverse2.78 Seam Welded Transverse3.89 Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Inter-comparison of irradiation growth of seamless and seam welded calandria tube

22. Residual Life Estimation of TAPS RPV SA302B (nickel modified) steel cladded with stainless steel is used as the pressure vessel material for the two 210 MWe boiling water reactors of the Tarapur Atomic Power Station. Charpy V-notch impact surveillance specimens representing the pressure vessel belt-line base, weld and the heat affected zone were irradiated at the wall and shroud locations. Some of these specimens from the wall and shroud locations were removed after 6.5 effective full power years (EFPY) of reactor operation. Subsequently additional specimens were also removed after 13 EFPY from the wall location. The surveillance data generated from these specimens were evaluated on the basis of USNRC Regulatory Guide 1.99, Revision 2. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

23. Location of surveillance baskets in TAPS reactor Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

24. Regulatory guides concerning the integrity of reactor vessels USNRC REGULATORY GUIDE POWER PLANT SURVEILLANCE DATA P – T LIMITS PTS LIMITS USE -  USE USE LIMITS RT NDT +  RT NDT 10 CFR 50.61 RT PTS  149  C,  132  C PTS Rule 10CFR50, App.G USE  68 J - Reg. Guide, ASME - 10 CFR 50, APP.G - Reg. Guide, ASME Temperature re Unirradiated Irradiated  USE  RT NDT CVCV Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

25. Credible Surveillance Data Sets MaterialLocation and Fluence, n/cm 2 (F > 1 MeV)  T =  C V41J  C USE, J Base Weld HAZ Wall, 5.31 x 10 17 - 6.5 EFPY14 22 26 140 128 163 Base Weld Wall, 1.06 x 10 18 – 13.0 EFPY25 35 146 124 Base Weld HAZ Shroud, 4.88 x 10 18 – 59.7 EFPY 38 40 36 135 113 163 Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

26. Adjusted Reference Temperature (ART) after 40 Years (60 years) Predicted Using Regulatory Guide 1.99, Revision 2 Position C.2 (w.r.t G.E. prescribed limit on ART of 93 0 C Wall Fluence n/cm 2, E > 1 MeV 0.25T Position Fluence, n/cm 2 EFPY CF °C Δ RT NDT °C ART °C 3.27 x 10 18 2.48 x 10 18 40 50.1 31(37)51(57) 3.27 x 10 18 2.48 x 10 18 40 52.9 33(39)53(59) 3.27 x 10 18 2.48 x 10 18 40 54.1 33(39) 53(59) Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

27. Pressure - Temperature Limits Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

28. MaterialsNo. of Credible Surveillance Data Sets (Corresponding EFPY) CF  RT NDT ( 0 C) RT PTS ( 0 C) RT PTS < SC Base 2 (6.5, 59.7) 3 (6.5,13.0, 59.7) 47.4 49.9 33 35 68 70 Yes Weld 2 (6.5, 59.7) 3 (6.5,13.0, 59.7) 52.9 59.0 37 42 72 77 Yes RT PTS = Initial RT NDT +  RT NDT + 33 Reference PTS Temperature (RT PTS ) after 40 years using PTS rule w.r.t SC of 132 0 C for base and 149 0 C for welds Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel

29. Stress Crack size Fracture toughness Component Zr-alloys fuel cladding pressure tube calandria tube garter spring Steels end fitting end shield pressure vessel How to enhance the expertise in estimation of residual life/extension of life of components ?

30. Component results Miniature specimen results Standard specimen results Inter-compare Correlate for Unirradiated material Miniature specimen results Standard specimen results Correlate for irradiated material I nter-compare Particle irrdn. I nter-compare Test Results Correlation Enhancement of Data Base Neutron irrdn.

31. Enhancement of Database Inter-comparison of results from standard specimens and miniaturised specimens

32. Calculation of PKA energy (E PKA ) Steps in Calculation of dpa Calculation of total lattice energy per incident neutron(E Lattice ) Selection of displacement threshold energy (E d ) Estimation of displacement cross section,  d Calculation of Displacement damage rate=  d x flux Calculation of Displacement damage, dpa = damage rate  time of exposure IRRADIATION ENVIRONMENT Damage rate dpa/s Cladding in PHWR 3.01  10 -8 SS Cladding in FBR 500 1.3  10 -6 SS with 3MeV Ni ++ ion 5  10 -3 Enhancement of Database

33. DISPLACEMENT X-SECTION OF Zr in PHWR Enhancement of Database

34. Technique development Summary Irradiation growth Simulation tests Ductile Brittle Transition Temperature Fracture Toughness Strength Fuel cladding Garter Spring Pressure tube End Fitting Calandria tube Ageing management of structural components PHWR BWR Pressure vessel Fuel cladding Std. Spn. Mini. Spn. Co-relation Crush strength Ductility Neutron irradn. Accl. Irrdn. Co-relation dpa coreln Enhancement of data base Delayed hydride cracking

35. CONCLUSIONS  Increasing demands on extending life of components calls for optimisation of evaluation techniques and analysis procedures, in addition to enhancement of data base  Input from R&D work towards identification and understanding of ageing degradation and establishing structure property correlations are key to ageing management of in-reactor structural materials