Irradiation Testing of Structural Materials in Fast Breeder Test Reactor
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1 Irradiation Testing of Structural Materials in Fast Breeder Test Reactor IAEA Technical Meet (TM 34779) Nov 17-21, 2008 IAEA, Vienna S.Murugan, V. Karthik, K.A.Gopal, N.G. Muralidharan, S. Venugopal, K.V. Kasiviswanathan, P.V.Kumar and Baldev Raj Indira Gandhi Centre for Atomic Research Kalpakkam INDIA
2 Irradiation Testing of Structural Materials in Fast Breeder Test Reactor Outline Introduction Fast Breeder Test Reactor Facilities available in FBTR Experiment with zirconium alloy specimens Experiment with D9 alloy specimens Development of non-instrumented gas-gap capsule Development of instrumented capsule Post irradiation testing of clad & wrapper tube samples Future plan of irradiation experiments Conclusion
3 Introduction Intense neutron flux High temperature Essential to assess performance of structural materials Effect of radiation on materials Irradiation hardening Embrittlement Void swelling Irradiation creep Irradiation testing on structural materials: (i) Irradiation experiments (ii) Testing of material samples from irradiated clad / wrapper tubes Zircaloy-2, Zr-2.5% Nb alloy Alloy D9 clad & wrapper tube of PFBR FBTR Main irradiation facility
4 FBTR
5 FBTR FBTR - Sodium cooled fast test reactor - Excellent facility for irradiation Neutron flux : n/cm 2 /s (order) Temperature: C (Present) : C (during next campaign) 1 st criticality - Mark I (70% PuC, 30% UC) - 22 FSAs (10.6 MWt) Core expansion - adding Mark-II (55% PuC, 45% UC) FSAs 8 nos. of high Pu MOX subassembly (44% PuO 2-56% UO 2 )-inducted into the core in Increase of LHR and burn up in Mark-I fuel to 400 W/cm - 25 to 155 GWd/t in stages
6 FBTR FBTR has completed 14 irradiation campaigns so far Core of 14th irradiation campaign: 27 Mark-I + 13 Mark-II + 8 MOX + 1 Test Total: 49 fuel subassemblies Evaluation of the performance of indigenously developed high Pu mono carbide driver fuel Irradiation of zirconium alloys used in Indian pressurised heavy water reactors for assessing their irradiation creep behavior Other physics and engineering experiments
7 Classification of irradiation experiments/capsules Irradiation experiments - Experiments carried out under intense neutron environment with simulated experimental conditions Irradiation experiments (i) Non-instrumented (ii) Instrumented Non-instrumented experiments - No online measurement/control facilities - Many positions available in FBTR - Calculations and/or PIE to give data on experimental parameters Instrumented experiments - With provision for on-line measurement/ control facilities - Experimental parameters (temperature) can be monitored/ controlled. Most of the structural material experiments in FBTR can be carried out using non-instrumented capsules Instrumented capsules for irradiation of structural materials are being developed.
8 Non-instrumented Irradiation Capsules In fuel spl. SA: 10 mm ID x 320 mm long Steel / Nickel: 12 to 18 mm ID x 320 mm long
9 Instrumented Irradiation Capsule Instrumented experiments Central 0-0 position of core of FBTR Leak tight access to the central axis of FBTR core CIPTEX offers of highest flux in FBTR Space available - 18 mm diameter x 320 mm long. Instrumented capsules for irradiation of structural materials are being developed.
10 Irradiation experiment carried out in FBTR on Zirconium alloys Pressure tube of PHWR - holds fuel bundles - hot pressurised heavy water - high fast neutron flux environment Critical component Zircaloy-2 and Zr-2.5% Nb alloy Zircaloy-2 Zr-2.5% Nb alloy Simplified cross section of PHWR Tin Iron Chromium Nickel Oxygen Niobium % % % % 140 ppm 100 ppm 100 ppm 1500 ppm 200 ppm 70 ppm 1300 ppm % Determination of irradiation creep rate of indigenously developed zirconium alloys using pressurised capsules in FBTR
11 Determination of Irradiation Creep Rate of Zirconium Alloys Pressurised capsules OD: 15.3 mm WT: 0.65 mm Length: 90 mm Filling gas: Argon + 2% helium Pressure: bar at RT Irradiation temp: C (Pressure: bar) Components of irradiation capsule
12 Irradiation capsule in special steel subassembly Irradiation locations in FBTR core 3 rd ring 0302,0305,0308,0311, th ring Post irradiation examination
13 0.35 (A) 0.30 CREEP STRAIN (%) E E E E E E E E E E+20 FLUENCE (E>1Mev), n/sq.cm 0.35 (B) Creep of Zircaloy-2 at 310ºC and average stress 1051 Kg/cm 2 CREEP STRAIN (%) RADIATION DAMAGE (dpa) (A) Creep strain vs Fluence Normalized to flux 3.2E13ncm -2 s -1 (E>1Mev) (Nominal value of flux in PHWR) (C) (B) Creep strain vs radiation damage (dpa) CREEP STRAIN (%) (C) Creep strain vs time EQUIVALENT EXPOSURE TIME IN PHWR (h) Fig.7 Creep of Zircaloy-2 at a temperature of 310 C and an average stress of 1051kg/sq.cm
14 1.20 (A) 1.00 CREEP STRAIN (%) E E E E E E E+21 FLUENCE (E>1Mev), n/sq.cm 1.20 (B) Creep of Zr-2.5% Nb at 314ºC and average stress 1500 Kg/cm 2 (A) Creep strain vs Fluence (B) Creep strain vs radiation damage (dpa) (C) Creep strain vs time CREEP STRAIN (%) CREEP STRAIN (%) Normalized to flux 3.2E13ncm -2 s -1 (E>1Mev) (Nominal value of flux in PHWR) RADIATION DAMAGE (dpa) EQUIVALENT EXPOSURE TIME IN PHWR (h) (C) Fig.8 Creep of Zr-2.5%Nb at a temperature of 314 C and an average stress of 1500 kg/sq.cm
15 Results Material Temperature (deg. C) Average stress (Kg/cm 2 ) [Fluence E>1Mev)] -1 Steady state creep rate [dpa] -1 [hr] -1 Zircaloy x x x 10-7 Zircaloy x x x 10-7 Zr-2.5% Nb alloy x x x 10-7 Zr-2.5% Nb alloy x x x 10-7 Zr-2.5% Nb alloy x x x 10-7
16 D9 Alloy Irradiation Experiment in FBTR C Cr Ni Mo Ti Mn Si S P B N Alloy D x C max max ppm max. Pressurised capsule for determination of irradiation creep behaviour of indigenously developed D9 alloy 6.6 mm OD & 0.45 mm WT D9 pressurised capsule (6.5 MPa pressure at RT)
17 D9 Alloy Irradiation Experiments in FBTR Present irradiation: Pressurised capsules of D9 clad tube 30 pressurised capsules Low temperature & low fluence tests Filling pressure at RT: 2.1, 4.2 and 6.3 MPa Temp 350 C Hoop stress at 350 C 30, 60 and 90 MPa Duration > 1 year
18 D9 alloy Irradiation experiments in FBTR Other D9 specimens: Longitudinal tubular tensile specimens Flat tensile specimens Swelling specimens Shear punch specimens
19 Gas-gap non-instrumented capsules Target temperatures: 400, 450, 500, 550, 600 C Sub capsule- specimens surrounded by static sodium Using nuclear heating in the specimens, gap width and composition of helium-argon mixture in the gap irradiation temperature is realised. Target damage: upto 90 dpa Duration of irradiation: 2 to 4 EFPYs Gas-gap capsule
20 Development of instrumented capsule HEATING COIL LEADS (OD 1.5/2.0) Outer tube (18 OD/ 16 ID) Sub capsule (14.5 OD/ 13 ID) Between outer tube and the equipment holder well of CIPTEX in FBTR- sodium Specimens - in sub capsule with static sodium Heating coil mm - around the sub capsule Grooves - machined on the outside surface of sub capsule Helium - in the gap between sub capsule and outer tube Thermocouples 3 nos. attached A 8 Ø18 A 22 TUBE FOR FILLING HELIUM HELIUM THERMOCOUPLE TUBE FOR FILLING UP OF SODIUM IN SUB CAPSULE (OD 2.5) (ID 1.5) (20mm long) ARGON SODIUM SPECIMENS (SS) HEATING COIL OD 1.5 / 2.0 THERMOCOUPLE ID HEIGHT 300 mm SUB CAPSULE OD Ø12 CERAMIC INSULATOR OUTER TUBE (IRRADIATION CAPSULE) OD 18 - ID 16 SPACER DETAIL - AA SPECIMENS(SS) HELIUM SUB CAPSULE THERMO COUPLES HEATING COIL OUTER TUBE
21 Development of Instrumented capsule Irradiation temperature measured by thermocouples Temperature to be kept as constant during irradiation Heating coil and output of T/c - connected to temperature controller Temperature of specimens -maintained within a small range during the period of irradiation (up to 250 W/cm) Development of critical joints by laser welding and nicrobrazing technique Capsule under development
22 Post Irradiation Testing of Samples from Clad and Wrapper tubes Clad & Wrapper tube: Type 316 SS 20% CW Change in properties due to high radiation damage levels Testing in hot cells Tubular specimens (70 mm long) from clad tube Damage: 0-83 dpa Temperature: 427 to 500 C Specimen after failure Remote tensile test machine in the hot cell facility
23 Results of PIE - Clad and Wrapper tubes RT - Room temperature, T test Test temperature, T irrad Irradiation temperature ( K) Trends in the UTS and % uniform elongation of SS316 cladding with dpa.
24 Results of PIE - Clad and Wrapper tubes Punched Specimen Specimen before Punching Shear punch test fixture, experimental setup and miniature specimens
25 Results of PIE - Clad and Wrapper tubes Trends in the RT strength and ductility of the irradiated SS316 wrapper of FBTR as a function of dpa. Volumetric swelling of FBTR cladding and wrapper determined from immersion density measurements as a function of dpa
26 Future Plan of Irradiation Experiments Irradiation using six nos. of non-instrumented gasgap capsules with irradiation temperatures from C D9, D9I and other materials Development of instrumented capsule with thermocouples Development of instrumented capsule with thermocouples and heating coil
27 Conclusions Facilities available in FBTR for structural material irradiation Details of pressurised capsules developed Irradiation experiment on zirconium alloy for thermal reactor programme Irradiation experiment being carried out on D9 alloy for fast reactor programme Development of non instrumented gas-gap capsule - completed Development of instrumented capsule in progress Some results from PIE of fuel cladding and fuel subassembly wrapper tubes of FBTR which have seen irradiation damage levels upto 83 dpa Future plan of irradiation experiments
28
29 Irradiation Testing on Structural Materials in Fast Breeder Test Reactor
30 Details of a sub capsule Sketch of noninstrumented gas gap capsule Photograph of mock up gas gap capsule
31 Sodium filled capsule with thermocouple attachments Exposure in electrical furnace Temperature (deg. C) Ambient Outer T1 Spci. In sodium Temperature (deg. C) Ambient Outer T1 Outer T2 Inner T1 Inner T2 Spec. Centre Time (s) Temperatures indicated by thermocouples Time (s) Results of temperature distribution by theoretical analysis
32 Machining of grooves to wind heating coil Sleeves attached to thermocouples using laser welding Heating coil wound over the tube
Irradiation Testing of Structural Materials in Fast Breeder Test Reactor. Abstract
Irradiation Testing of Structural Materials in Fast Breeder Test Reactor S. Murugan*, V. Karthik, K.A. Gopal, N.G. Muralidharan, S. Venugopal, K.V. Kasiviswanathan, P.V. Kumar and Baldev Raj Indira Gandhi
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