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 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 Why to evaluate radiation damage in reactor structural Materials ? Commercial Reactors • Pressurised Heavy Water Reactor (PHWR) • Boiling Water Reactor (BWR) • Water Water Energy Reactor (WWER) Research Reactors • CIRUS • DHRUVA Structural Materials/ Components Zr-alloys • fuel cladding • pressure tube • calandria tube • garter spring : Zr-2/Zr-4 : Zr-2/Zr-2.5Nb : Zr-2 :Zr-0.5Cu-2.5Nb Steels • end fitting : 403 SS • end shield : 203D/304 SS • pressure vessel : 302B-Ni modified (A533B) WWER 1000 Components experience aggressive environment of : • Temperature • Stress • Corrosion • Radiation damage Primary radiation damage is from neutron population 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 End-Of-Life (EOL) fluence of components Component n-fluence (>1 MeV) dpa Time (years) Fuel cladding Pressure tube Calandria tube 2*1021 2*1022 2*1022 4.4 44 44 16 16 16 Garter spring End fitting End shield 2*1022 6*1019 5*1019 44 0.13 0.11 16 1 1 TAPS RPV WWER 1000 RPV 3.3*1018 3.7*1019 0.007 0.08 1/12 1 Saturation fluence : 1*1021 n/cm2 (>1MeV), 2.2 dpa Threshold fluence : 5*1017 n/cm2 (>1MeV), What life limiting structural materials were evaluated ? Components Evaluated Component Fuel Cladding Pressure Tube Garter Spring End Fitting Calandria Tube TAPS RPV Origin of specimens Operating reactor Operating reactor Operating reactor Research Reactor Research Reactor Operating reactor Types of Tests Conducted Type of Test Components Tension Pressure Vessel, Cladding, Garter Spring, End-fitting Pressure Vessel, End-fitting PressureVessel,Pressure Tube Impact Fracture Toughness Crush Test Irradiation Growth Delayed Hydride Crack(DHC) Garter Spring Calandria Tube Pressure Tube Life Limiting Phenomenon/Property Component Fuel Cladding Pressure Tube Garter Spring End Fitting Calandria Tube TAPS RPV Property Ductility Fracture Toughness, DHC Crush Strength DBTT Irradiation Growth DBTT (Fracture toughness) Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel 14 12 850 10 800 8 750 6 4 700 2 650 0 0 2000 4000 6000 8000 10000 12000 14000 16000 Burn up (MWd/T) Tensile Property of Claddings Uniform Elongation (%) Ultimate Tensile Strength (MPa) 900 Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Pressure Tubes Evaluated Reactor PT EFPY MAPS-2 N – 10 4.85 MAPS-1 P – 13 6.25 RAPS-2 K – 07 8.25 MAPS-1 J - 07 9.5 Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel 100 0 300 C 80 CCL (mm) Safe 60 50 mm 40 20 Unsafe 0 100 0 250 C CCL (mm) 80 Safe 60 50 mm 40 20 Unsafe 0 40 80 120 160 Equivalent hydrogen content (ppm) 200 CCL for various PTs Evaluated Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Material Temperature (0C) Zr-2 250 290 DHC velocity (mm/h) 0.07 0.12 Zr-2.5Nb 250 290 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 Garter Springs Evaluated Spring Identification Reactor EFPY 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) RAPS-2 8.50 14 ( stretch test) Not Identified Numbers Examined ( type of test ) Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel 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’ clock 728 a O-11(RAPS-2, 6.5) 6 O’ clock 845 b N-10(MAPS-2,4.8) 6 O’ clock 539 b K-14(MAPS-2,3.6) 6 O’ clock 410 b K-19(NAPS-1,1.8) 6 O’ clock 428 b 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 Irradiation details Specimens Charpy V-notch Location Tray rod location in CIRUS reactor Neutron Flux 2.4 x 1012n.cm-2.S-1 E > 1.0 Mev Duration of Irradiation 48 Days at Full Power Neutron Fluence 1 x 1019n.cm-2 E > 1.0 Mev Irradiation Temperature 290º C±10º C Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Energy Un -Irradiated Δ USE = 44J Irradiated Δ T=750C 1 X 1019n/cm2, >1 MeV Temperature At EOL fluence of 6 X 1019n/cm2, >1 MeV Δ T EOL = 75 X (6 X 1019/ 1 X 1019)0.33 =1360C RTNDT,EOL = 1700C Operating temperature : 2500C, 3000C Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Specimen Growth Strain (10-4) Seamless Longitudinal 4.70 Seam welded Long. 4.78 Seamless Transverse 2.78 Seam Welded Transverse 3.89 Inter-comparison of irradiation growth of seamless and seam welded calandria tube 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. Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Location of surveillance baskets in TAPS reactor Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Regulatory guides concerning the integrity of reactor vessels P – T LIMITS POWER PLANT SURVEILLANCE DATA -10 CFR 50, APP.G -Reg. Guide, ASME RTNDT + RTNDT Unirradiated USE CV RTNDT PTS LIMITS Irradiated Temperature re USNRC REGULATORY GUIDE 10 CFR 50.61 RTPTS 149C, 132C PTS Rule USE - USE USE LIMITS 10CFR50, App.G USE 68 J - Reg. Guide, ASME Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Credible Surveillance Data Sets Material Location and Fluence, n/cm2 (F > 1 MeV) T = CV41J USE, J C Base Weld HAZ Wall, 5.31 x 1017 - 6.5 EFPY 14 22 26 140 128 163 Base Weld Wall, 1.06 x 1018 – 13.0 EFPY 25 35 146 124 Base Weld HAZ Shroud, 4.88 x 1018 – 59.7 EFPY 38 40 36 135 113 163 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 2 Position C.2 (w.r.t G.E. prescribed limit on ART of 930C 0.25T Position Fluence, n/cm2 EFPY CF °C Δ RTNDT °C ART °C 3.27 x 1018 2.48 x 1018 40 50.1 31(37) 51(57) 3.27 x 1018 2.48 x 1018 40 52.9 33(39) 53(59) 3.27 x 1018 2.48 x 1018 40 54.1 33(39) 53(59) Wall Fluence n/cm2, E > 1 MeV Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Pressure - Temperature Limits Cladding/ Pressure Tube/ Garter Spring/ End fitting/ Calandria tube/ Pressure vessel Materials No. of Credible Surveillance Data Sets (Corresponding EFPY) CF RT NDT (0C) RTPTS (0C) RTPTS < SC Base 2 (6.5, 59.7) 47.4 33 68 Yes Base 3 (6.5,13.0, 59.7) 49.9 35 70 Yes Weld 2 (6.5, 59.7) 52.9 37 72 Yes Weld 3 (6.5,13.0, 59.7) 59.0 42 77 Yes 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 Fracture toughness Zr-alloys • fuel cladding •pressure tube • calandria tube • garter spring Steels • end fitting • end shield • pressure vessel Component Stress How to enhance the expertise in estimation of residual life/extension of life of components ? Crack size Enhancement of Data Base Test Results Correlation Miniature specimen results Component results Miniature specimen results Inter-compare Correlate for Unirradiated material Standard specimen results Inter-compare Inter-compare Correlate for irradiated material Standard specimen results Inter-compare Neutron irrdn. Particle irrdn. Enhancement of Database Inter-comparison of results from standard specimens and miniaturised specimens 800 700 Yield strength (MPa) 650 750 700 Small punch test Small punch test 600 550 500 450 400 350 650 600 550 500 450 Ultimate tensile strength (MPa) 300 250 250 300 350 400 450 500 550 600 650 400 400 700 450 550 600 650 700 750 800 Conventional tension test Conventional tension test 400 24 Uniform elongation (%) 22 2 Fracture toughness (kJ/m ) 350 20 Small punch test Small punch test 500 18 16 14 12 10 8 300 250 200 6 150 4 2 0 0 2 4 6 8 10 12 14 16 18 20 22 24 Conventional tension test 100 100 150 200 250 300 Standard specimen 350 400 Enhancement of Database Steps in Calculation of dpa Calculation of PKA energy (EPKA) Calculation of total lattice energy per incident neutron(ELattice) IRRADIATION ENVIRONMENT Damage rate dpa/s Cladding in PHWR 3.0110-8 Estimation of displacement cross section, d SS Cladding in FBR 500 1.310-6 Calculation of Displacement damage rate= d x flux SS with 3MeV Ni++ ion 510-3 Selection of displacement threshold energy (Ed) Calculation of Displacement damage, dpa = damage rate time of exposure Enhancement of Database Displacement Cross section (barns) DISPLACEMENT X-SECTION OF Zr in PHWR 6000 4000 E(MeV) barn 0.01 65 0.05 149 0.1 305 0.3 338 0.5 633 1.0 834 5.0 1632 Total Elastic 2000 Inelastic 0 0.01 0.1 1 Neutron Energy (MeV) 10 Summary Ductility PHWR Strength Fracture Toughness BWR Delayed hydride cracking Calandria tube Ductile Brittle Simulation tests Transition Temperature Crush strength End Fitting Pressure vessel Garter Spring Irradiation growth Fuel cladding Pressure tube Technique development Co-relation Fuel cladding Mini. Spn. Enhancement of data base Std. Spn. Neutron irradn. Accl. Irrdn. dpa coreln Co-relation Ageing management of structural components 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