Study on Radiation Induced Ageing of Power Reactor Components S. Chatterjee , K.S. Balakrishnan, Priti Kotak Shah, D.N.

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. Whyto 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. Whyto 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 Saturation fluence : 1*1021 n/cm2 (>1MeV), 2.2 dpa Threshold fluence : 5*1017 n/cm2 (>1MeV),

9. What life limiting structural materials were evaluated ?

10. Components Evaluated

11. Types of Tests Conducted

12. Life Limiting Phenomenon/Property

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

14. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Pressure Tubes Evaluated

15. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Safe Unsafe CCL for various PTs Evaluated Safe Unsafe 40 80120 160 200 Equivalent hydrogen content (ppm)

16. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel 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

17. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Garter Springs Evaluated

18. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Room Temperature Crush Test Results * Load values depicted are typically one order more in magnitude than the design load ** a: Specimen got crushed, b: Gap got closed

19. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Irradiation details

20. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Un -Irradiated Δ USE = 44J Irradiated 1 X 1019n/cm2, >1 MeV Energy Δ T=750C Temperature At EOL fluence of 6 X 1019n/cm2, >1 MeV ΔT EOL = 75 X (6 X 1019/ 1 X 1019)0.33 =1360C RTNDT,EOL = 1700C Operating temperature : 2500C, 3000C

21. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Inter-comparison of irradiation growth of seamless and seam welded calandria tube

22. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel 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.

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

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

25. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Credible Surveillance Data Sets

26. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Adjusted Reference Temperature (ART) after 40 Years (60 years) Predicted Using Regulatory Guide 1.99, Revision 2Position C.2 (w.r.t G.E. prescribed limit on ART of 930C

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

28. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel RTPTS = Initial RTNDT +  RT NDT + 33 Reference PTS Temperature (RTPTS) after 40 years using PTS rule w.r.t SC of 1320C for base and 1490C for welds

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

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

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

32. Enhancement of Database Steps in Calculation of dpa Calculation of PKA energy (EPKA) Calculation of total lattice energy per incident neutron(ELattice) Selection of displacement threshold energy (Ed) Estimation of displacement cross section, d Calculation of Displacement damage rate= d x flux Calculation of Displacement damage, dpa = damage rate  time of exposure

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

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

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

36. Thank you