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 Centre for Atomic ResearchKalpakkam- 603 102

INDIA

IAEA Technical Meet (TM – 34779)Nov 17-21, 2008

IAEA, Vienna

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

IntroductionEffect of radiation on materials� Irradiation hardening� Embrittlement� Void swelling� Irradiation creep

� Zircaloy-2, Zr-2.5% Nb alloy

� Alloy D9 – clad & wrapper tube of PFBR

� Intense neutron flux � High temperature

� Essential to assess performance of structural materials

Irradiation testing on structural materials:(i) Irradiation experiments(ii) Testing of material samples from

irradiated clad / wrapper tubes

FBTR – Main irradiation facility

FBTR

FBTR

� Neutron flux : 1015 n/cm2/s (order)Temperature: 330- 425°C (Present)

: 380-488°C (during next campaign)

� FBTR - Sodium cooled fast test reactor- Excellent facility for irradiation

� Increase of LHR and burn up in Mark-I fuel

- 250 to 400 W/cm - 25 to 155 GWd/t

in stages

� 1st 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% PuO2- 56% UO2)-inducted into the core in 2007.

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 heavywater reactors for assessing their irradiation creep behavior

� Other physics and engineering experiments

Classification of irradiation experiments/capsules

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.

Instrumented capsules for irradiation of structural materials are being developed.

Irradiation experiments(i) Non-instrumented(ii) Instrumented

Irradiation experiments -Experiments carried out under intense neutron environment with simulated experimental conditions

Most of the structural material experiments in FBTR can be carried out using non-instrumented capsules

Non-instrumented Irradiation CapsulesIn fuel spl. SA: 10 mm ID x 320 mm long

Steel / Nickel: 12 to 18 mm ID x 320 mm long

Instrumented capsules for irradiation of structural materials are being developed.

� 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 Irradiation Capsule

Irradiation experiment carried out in FBTR on Zirconium alloys

Pressure tube of PHWR- holds fuel bundles - hot pressurised heavy water- high fast neutron flux environmentCritical component

Simplified cross section of PHWR

Determination of irradiation creep rate of indigenously developed zirconium alloys using pressurised capsules in FBTR

Zircaloy-2 and Zr-2.5% Nb alloy

2.4-2.8%1300 ppm70 ppm200 ppm1500 ppm100 ppmZr-2.5% Nb alloy

100 ppm140 ppm0.03-0.08%0.05-0.15%0.7-0.20%1.2-1.7%Zircaloy-2

NiobiumOxygenNickelChromiumIronTin

Components of irradiation capsule

Pressurised capsulesOD: 15.3 mmWT: 0.65 mmLength: 90 mmFilling gas: Argon + 2% heliumPressure: 50-65 bar at RTIrradiation temp: 306-319°C(Pressure: 100-130 bar)

Determination of Irradiation Creep Rate of Zirconium Alloys

Irradiation capsule in special steel subassembly

Irradiation locations in FBTR core3rd ring – 0302,0305,0308,0311,03144th ring - 0403

Post irradiation examination

Creep of Zircaloy-2 at 310ºCand average stress 1051 Kg/cm2

(A) Creep strain vs Fluence(B) Creep strain vs radiation

damage (dpa)(C) Creep strain vs time

Fig.7 Creep of Zircaloy-2 at a temperature of 310 C and an average stress of 1051kg/sq.cm

0.000.050.100.150.200.250.300.35

0.0E+00 1.0E+20 2.0E+20 3.0E+20 4.0E+20 5.0E+20 6.0E+20 7.0E+20 8.0E+20 9.0E+20FLUENCE (E>1Mev), n/sq.cm

CREE

P STR

AIN

(%)

(A)

0.000.050.100.150.200.250.300.35

0 1000 2000 3000 4000 5000 6000 7000 8000EQUIVALENT EXPOSURE TIME IN PHWR (h)

CREE

P STR

AIN

(%)

Normalized to flux 3.2E13ncm-2s-1(E>1Mev)(Nominal value of flux in PHWR)

(C)

0.000.050.100.150.200.250.300.35

0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8RADIATION DAMAGE (dpa)

CREE

P STR

AIN

(%)

(B)

Creep of Zr-2.5% Nb at 314ºCand average stress 1500 Kg/cm2

(A) Creep strain vs Fluence(B) Creep strain vs radiation

damage (dpa)(C) Creep strain vs time

Fig.8 Creep of Zr-2.5%Nb at a temperature of 314 C and an average stress of 1500 kg/sq.cm

0.00

0.20

0.40

0.60

0.80

1.00

1.20

0.0E+00 2.0E+20 4.0E+20 6.0E+20 8.0E+20 1.0E+21 1.2E+21FLUENCE (E>1Mev), n/sq.cm

CREE

P STR

AIN

(%)

(A)

0.00

0.20

0.40

0.60

0.80

1.00

1.20

0 0.5 1 1.5 2 2.5RADIATION DAMAGE (dpa)

CREE

P STR

AIN

(%)

(B)

0.00

0.20

0.40

0.60

0.80

1.00

1.20

0 1000 2000 3000 4000 5000 6000 7000 8000 9000 10000EQUIVALENT EXPOSURE TIME IN PHWR (h)

CREE

P STR

AIN

(%)

(C)Normalized to flux 3.2E13ncm-2s-1(E>1Mev)(Nominal value of flux in PHWR)

12.067 x 10-7

5.725 x 10-7

5.125 x 10-71.300 x 10-75.100 x 10-7

[hr]-1

4.050 x10-3

2.600 x10-3

2.450 x10-30.500 x10-31.625 x10-3

[dpa]-1[Fluence E>1Mev)]-1

1.0530 x 10-23

0.5000 x 10-23

0.4670 x 10-230.1100 x 10-230.4467 x 10-23

Steady state creep rate

1457

1467

150010511062

Average stress (Kg/cm2)

319

318

314310306

Tempe-rature(deg. C)

Zr-2.5% Nb alloy

Zr-2.5% Nb alloy

Zr-2.5% Nb alloy

Zircaloy-2Zircaloy-2

Material

Results

D9 Alloy Irradiation Experiment in FBTRPressurised 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)

10-20 ppm.B

5 - 7.5 x CTi

0.005 max.N

0.02 max.P

Alloy D9

0.01 max.S0.50 - 0.75Si1.65 - 2.35Mn

2.0 - 2.5Mo14.5 - 15.5Ni13.5 - 14.5Cr0.035 - 0.05C

30 pressurised capsules

> 1 year30, 60 and 90 MPa350°CDurationHoop stress at 350°C Temp

D9 Alloy Irradiation Experiments in FBTRPresent irradiation:Pressurised capsules of D9 clad tube

Low temperature &low fluence tests

Filling pressure at RT: 2.1, 4.2 and 6.3 MPa

D9 alloy Irradiation experiments in FBTR Other D9 specimens:Longitudinal tubular tensile specimensFlat tensile specimens Swelling specimensShear punch specimens

Gas-gap non-instrumented capsules

Gas-gap capsule

� 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

Development of instrumented capsule HEATING COIL LEADS (OD 1.5/2.0)

OUTER TUBE (IRRADIATION CAPSULE)

Ø18

25

8

18

2011

THERMOCOUPLE

(OD 2.5) (ID 1.5) (20mm long)TUBE FOR FILLING UP OF SODIUM IN SUB CAPSULE

ARGONSODIUM

SPECIMENS (SS)HEATING COILOD 1.5 / 2.0

THERMOCOUPLE

22 ID 10.5 - HEIGHT 300 mmSUB CAPSULE OD Ø12

CERAMIC INSULATOR

A A

HELIUM

OD 18 - ID 16

SPACER

TUBE FOR FILLING HELIUM

THERMO COUPLES

HELIUM SUB CAPSULE

SPECIMENS(SS)

HEATING COIL

OUTER TUBEDETAIL - AA

300

� 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 - 1.5 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

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

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

Results of PIE - Clad and Wrapper tubes

Trends in the UTS and % uniform elongation of SS316 cladding with dpa.

RT - Room temperature, T test – Test temperature, Tirrad –Irradiation temperature (703–773K)

Results of PIE - Clad and Wrapper tubes

Shear punch test fixture, experimental setup and miniature specimens

Specimen before Punching

Punched Specimen

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

Future Plan of Irradiation Experiments

� Irradiation using six nos. of non-instrumented gas-gap capsules with irradiation temperatures from 400-600 ̊C – D9, D9I and other materials

� Development of instrumented capsule with thermocouples

� Development of instrumented capsule with thermocouples and heating coil

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

Irradiation Testing on Structural Materials in Fast Breeder Test Reactor

Details of a sub capsule

Sketch of non-instrumented gas gap capsule

Photograph of mock up gas gap capsule

Sodium filled capsule with thermocouple

attachments

Exposure in electrical furnace

0

50

100

150

200

250

0 1000 2000 3000 4000 5000Time (s)

Temp

eratu

re (d

eg. C

)

AmbientOuter T1Spci. In sodium

0

50

100

150

200

250

0 1000 2000 3000 4000 5000Time (s)

Temp

eratur

e (de

g. C)

AmbientOuter T1Outer T2Inner T1Inner T2Spec. Centre

Temperatures indicated by thermocouples

Results of temperature distribution by

theoretical analysis

Machining of grooves to wind heating coil

Heating coil wound over the tube

Sleeves attached to thermocouples using laser welding