ML18267A094

Irradiation Effects

ML18267A094
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Issue date:	09/24/2018
From:	Tregoning R L NRC/RES/DE
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Effects of Irradiation on Fatigue LifeRobert TregoningNuclear Regulatory CommissionPublic Meeting on Environmentally Assisted Fatigue (EAF) Research and Related ASME ActivitiesNuclear Regulatory Commission HeadquartersSeptember 25, 2018This presentation was prepared as an account of work conducted by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, apparatus, product, or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights. The views expressed in this paper are not necessarily those of the U.S. Nuclear Regulatory Commission.

Background

• Irradiation effects are not considered in ASME Section III fatigue evaluations

• Little current experimental data on irradiation effects for nuclear materials under conditions representative of commercial light

-water reactorsMost data dates to mid

-1970s or earlierMuch of the irradiation conducted in fast flux reactors at either higher (> 400 o C) or lower (< 100 oC) temperatures than service conditions

• Changes in microstructure resulting from irradiation have been observed but not correlated with fatigue lifeMicrostructure and microchemistry changesCavities and He bubble formation

• Some fatigue tests have been performedFatigue crack growth rate (FCGR)-N) September 25, 2018Public Meeting on EAF and Related ASME Activities 2

Effects of Irradiation on FCGR

• Experimental Breeder Reactor II (EBR

-II) TestsType 304 & 316 SS irradiated at 410 oC to 1.2x10 22n/cm 2and tested at 427 oC and 593 o C427 oC results*CGRs in irradiated specimens up to factor of 2 higher for K < 44 *CGRs in irradiated specimens lower for K > 44 m593 oC results*Opposite trends compared to the 427 oC results*Advanced Test Reactor (ATR) TestsType 304 and 316 SS irradiated at 288 oC to 1.8x10 21n/cm 2and tested at 427 o C.Results*CGRs in irradiated specimens 2

-4 times lower

• No significant deleterious effects of irradiation in these tests and improved resistance to fatigue cracking observed under some conditionsSeptember 25, 2018Public Meeting on EAF and Related ASME Activities 3

-N*Engineering Test Reactor (ETR) TestsType 347 SS irradiated at 60 o C to 5.5x10 25n/m 2and aboveTested at room temperature

• Low strain amplitudes (< 0.35%): Irradiation appears to increase fatigue life*High strain amplitudes: Irradiation appears to decrease fatigue lifeSeptember 25, 2018Public Meeting on EAF and Related ASME Activities 4

-N, cont.*Tests on CW Type 316 SS Tubes Removed from PWR PlantIrradiated to fluences< 10 22n/m 2to > 3x10 26n/m 2Tested in air and PWR

-simulated water at 325 oC and various strain ratesNo baseline unirradiatedtests for comparison

• Fatigue life for tube specimens decreased by approximately 1.5

• -N relationship of irradiated material appears flatter than the ASME Code mean curveSeptember 25, 2018Public Meeting on EAF and Related ASME Activities 5

-N, cont.*Tests on Solution Annealed Type 304 SS Bars Removed from PWR PlantIrradiated to fluencesfrom 0.5 to 1x10 26n/m 2Tested in air and PWR

-simulated water at 325 oC and various strain ratesNo baseline unirradiatedtests for comparison

• As with CW 316 results, irradiated life curve appears flatter than ASME mean curve

• Implication is that irradiation may increase fatigue life at low strain amplitudes while decreasing life at higher strain amplitudesSeptember 25, 2018Public Meeting on EAF and Related ASME Activities 6

Summary*Effects of irradiation on fatigue crack growth rates (CGRs)

• Conclusive trends cannot be established due to limited LWR-representative data*No significant increase in FCGRs due to irradiation observed in existing test data

• Decreases in FCGRs due to irradiation observed under some conditions

• Effects of irradiation on fatigue life (i.e., -N)*Conclusive trends cannot be established due to limited LWR

-representative data*However, consistent trends have been observed in existing test data

• Decreases in fatigue life at higher (i.e., > 0.35 to 0.6%) strain levels

• Increases in fatigue life at low (i.e., < 0.35 to 0.6%) strain levels

• Trends appear consistent with material changes caused by irradiation

• Hardening causes increased tensile strength, leading to increased high cycle fatigue resistance

• Reduced ductility could lead to decreases in low

-cycle fatigue lifeSeptember 25, 2018Public Meeting on EAF and Related ASME Activities 7

Discussion

-Next Steps

• ResearchEffects of irradiation on fatigue life are one of the least understood of all potential contributing factors to fatigue lifeInformation on any additional past, current, or planned research in this area would be importantRelevance/feasibility of using FCR portion of IASCC testing as additional data.

• ImplicationsAny deleterious effects of irradiation on fatigue life will increase with operating timeIssue is only relevant to a limited scope of reactor internal componentsScoping would be useful to identify components of interest

• Potential resolutionLimited test program would be useful in confirming/predicting significant trendsTesting representative materials under LWR conditions Initial focus should be on low

-cycle fatigue effects and the transition between low

-cycle and high

-cycle regimes.Other thoughts?September 25, 2018Public Meeting on EAF and Related ASME Activities 8