ML14251A014

H.B. Robinson, Unit 2 - Relief Request (RR)-11 for Relief from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1

ML14251A014
Person / Time
Site:	Robinson
Issue date:	08/27/2014
From:	Gideon W R Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/14-0092
Download: ML14251A014 (10)

Text

W. R. GideonH. B. Robinson SteamElectric Plant Unit 2Site Vice PresidentDuke Energy ProgressENERGY. -3581 West Entrance RoadHartsville, SC 295500:8438571701F: 843 857 1319Randy. Gideonc~duke-energy. com10 CFR 50.55aSerial: RNP-RA/14-0092AUG 2 7 2014Attn: Document Control DeskUnited States Nuclear Regulatory CommissionWashington, DC 20555-0001.H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23RELIEF REQUEST (RR)-11 FOR RELIEF FROM VOLUMETRIC/SURFACE EXAMINATIONFREQUENCY REQUIREMENTS OF ASME CODE CASE N-729-1Pursuant to 10 CFR 50.55a(a)(3)(i), Duke Energy Progress, Inc., hereby requests a relief fromperforming the required volumetric/surface examinations for the H. B. Robinson Steam ElectricPlant (RNP), Unit No. 2 reactor vessel closure head (RVCH) components identified in ASMECode, Section Xl, Code Case N-729-1. The details and justification for this request are providedby the enclosure to this letter.Duke Energy Progress, Inc. requests approval of the enclosed relief request by January 2015 tosupport the Spring 2015 refueling outage.This letter contains no new Regulatory Commitments.If you have any questions concerning this matter, please contact Mr. Richard Hightower,Manager -Regulatory Affairs at (843) 857-1329.Sincerely,Sharon W. PeavyhouseDirector -Nuc Org EffectivenessSWP/cac

Enclosure:

H. B. Robinson Steam Electric Plant, Unit No. 2, Relief Request (RR)-1 1 for Relieffrom Volumetric/Surface Examination Frequency Requirements of ASME CodeCase N-729-1cc: Mr. V. M. McCree, NRC Region IIMs. Martha Barillas, NRC Project Manager, NRRNRIC Resident Inspector United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)H. B. Robinson Steam Electric Plant, Unit No. 2, Relief Request (RR)-1 1 for Relieffrom Volumetric/Surface Examination Frequency Requirements of ASME CodeCase N-729-1 United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)1. American Society of Mechanical Engineers (ASME) Code Component(s) AffectedThe affected components are ASME Class 1 Pressurized Water Reactor (PWR) ReactorVessel Closure Head (RVCH) nozzles and partial-penetration welds fabricated with primarywater stress corrosion cracking (PWSCC) resistant materials. H. B. Robinson SteamElectric Plant (RNP) Unit 2 penetration tubes and vent pipe are fabricated from Alloy 690with alloy 52/152 attachment welds.2. Applicable Code Edition and AddendaThe 5th inservice inspection (ISI) interval Code of record for RNP Unit 2 is the 2007 Editionwith 2008 Addenda of ASME Boiler and Pressure Vessel Code,Section XI, "Rules forInservice Inspection of Nuclear Power Plant Components."3. Applicable Code RequirementThe Code of Federal Regulations (CFR) 10CFR50.55a(g)(6)(ii)(D)(1), requires (in part):All licensees of pressurized water reactors shall augment their ISI program with ASMECode Case N-729-1 (Ref. 1) subject to the conditions specified in paragraphs(g)(6)(ii)(D)(2) through (6) of this section. Licensees of existing operating reactors as ofSeptember 10, 2008 shall implement their augmented ISI program by December 31, 2008.10CFR50.55a(g)(6)(ii)(D)(3) conditions ASME Code Case N-729-1 by stating:Instead of the specified 'examination method' requirements for volumetric and surfaceexaminations in Note 6 of Table 1 of Code Case N-729-1, the licensee shall performvolumetric and/or surface examination of essentially 100 percent of the required volume orequivalent surfaces of the nozzle tube, as identified by Figure 2 of ASME Code Case N-729-1. A demonstrated volumetric or surface leak path assessment, through all J-groovewelds shall be performed. If a surface examination is being substituted for a volumetricexamination on a portion of a penetration nozzle that is below the toe of the J-groove weld[Point E on Figure 2 of ASME Code Case N-729-1], the surface examination shall be of theinside and outside wetted surface of the penetration nozzle not examined volumetrically.ASME Code Case N-729-1 specifies that the reactor vessel upper head components shallbe examined on a frequency in accordance with Table 1 of this code case.4. Reason for RequestTreatment of Alloy 690 RPV Closure Heads in Code Case N-729-1 was intended to beconservative and subject to reassessment once additional laboratory data and plantexperience on the performance of Alloy 690 and Alloy 52/152 weld metals becomeavailable. Using plant and laboratory data, Electric Power Research Institute (EPRI)document Materials Reliability Program (MRP) -375 was developed to support atechnically based volumetric / surface re-examination interval using appropriate analyticaltools. This technical basis demonstrates that the re-examination interval can be extendedto the requested interval length while maintaining an acceptable level of quality and safety.Duke Energy is requesting approval of this alternative to allow the use of the ISI intervalextension for the affected RNP Unit 2 component.

United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)The expedited nature of this request is the result of Duke Energy only just becoming awareof recent industry requests for relief from Code Case N-729-1 by plants required to performthe inspections, and the Company's interest in significant dose reduction during the Spring2015 refueling outage.5. Proposed Alternative and Basis for UseProposed AlternativePursuant to 1OCFR 50.55a (a)(3)(i), Duke Energy requests an alternative from performingthe required volumetric/surface examinations for the RNP RVCH components identifiedabove at the frequency prescribed in ASME Code,Section XI, Code Case N-729-1.Specifically, Duke Energy requests to extend the frequency of the volumetric/surfaceexamination of the RNP RVCH of Table 1, Item B4.40 of ASME Code Case N-729-1 forapproximately 3 years beyond the one inspection interval (nominally 10 calendar years)from installation of the RNP replacement RVCH. This request would extend thevolumetric/surface examination to the 31 st refueling outage which is scheduled tocommence in September of 2018. No alternative examination processes are proposed tothose required by ASME Code Case N-729-1, as conditioned by 1 OCFR50.55a(g)(6)(ii)(D).The visual examinations and acceptance criteria as required by Item B4.30 of Table 1 ofASME Code Case N-729-1 are not affected by this request and will continue to beperformed on a frequency of every 3RD refueling outage or 5 calendar years, whichever isless.Basis for UseThe original RNP RVCH, which was manufactured with Alloys 600/82/182 materials, wasreplaced with a new RVCH using Alloys 690/52/152 material during the refueling outagethat returned to operation in October 2005. In accordance with Table 1 of ASME CodeCase N-729-1, Item B4.40, as conditioned by 10CFR50.55a(g)(6)(ii)(D)(3), RNP will berequired to perform a volumetric and/or surface examination of essentially 100% of theRVCH by the end of 2015.The basis for the inspection frequency for ASME Code Case N-729-1 comes, in part, fromthe analysis performed in EPRI Materials Reliability Program (MRP)-1 11 (Ref. 2 ) whichwas summarized in the safety assessment for RVCHs in EPRI MRP-1 10 (Ref. 3). Thematerial improvement factor for Primary Water Stress Corrosion Cracking (PWSCC) ofAlloys 690/52/152 materials over that of mill annealed Alloys 600/82/182 was shown by thisreport to be in the order of 26 or greater.Additional Evaluations Performed under EPRI MRP-375Further evaluations were performed to demonstrate the resistance of Alloys 690/52/152 toPWSCC under a recent EPRI MRP initiative provided in EPRI MRP-375 (Ref. 4). Thisreport presents both deterministic and probabilistic evaluations that assess the improvedPWSCC resistance of Alloys 690/52/152.Operating experience to date for replacement and repaired components using Alloys690/52/152 has shown a proven record of resistance to PWSCC during numerousexaminations in the 20+ years of its application. This includes steam generators,pressurizers, and RVCHs.

United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)In particular, at the completion of the spring 2014 refueling outage season, Alloys690/52/152 operating experience includes inservice volumetric/surface examinationsperformed on thirteen of the 40 plant replacement RVCHs in the US in accordance withASME Code Case N-729-1.In France in 2013, a second 10 year nondestructive examination (NDE) inspection wasperformed on one of the first reactor vessel (RV) heads to be replaced with alloy690/52/152 material. There were no reports of PWSCC having been detected afterapproximately 20 years of service.The evaluation performed in MRP-375 considers a simple Factor of Improvement (FOI)approach applied in a conservative manner to model the increased resistance of Alloy 690compared to Alloy 600 at equivalent temperature and stress conditions. Even though basemetal and welding variability of test data exist (i.e. heat affected zones, weld dilution zones,etc.), relative, but conservative, FOls were estimated for the material improvements ofAlloys 690/52/152 materials using an extensive database of test data. Results for bothcrack initiation and crack growth conclude a higher resistance to PWSCC for Alloy 690base material and Alloy 52/152 weld materials. EPRI MRP-375, Figures 3-2, 3-4, and 3-6provide crack growth data for Alloy 690/52/152 materials and heat affected zones withrepresented curves plotting FOls of 1, 5, 10, and 20. A FOI of 20 bounds most of the dataplotted, however, a FOI of 10 or less bounds all of the data.EPRI MRP-375, Table 3-6 provides a summary of crack growth rate (CGR) and crackinitiation data. For crack initiation, FOls reported although significant, are conservativebecause, in many cases, crack initiation of Alloys 690/52/152 was not observed duringtesting; instead, the initiation time was assumed to be equivalent to the test duration.Additionally, many of the Alloy 690 crack growth rate tests were performed on specimenswith considerable amounts of cold work (up to 40%), which is known to accelerate CGRs torates that are not representative of cold work levels applicable to reactor vessel headpenetrations.EPRI MRP-375 then performed a combination of deterministic and probabilisticevaluations to establish a reasonable inspection interval for Alloy 690 RVCHs. Thedeterministic technical basis applies industry-standard crack growth calculationprocedures to predict time to certain adverse conditions under various conservativeassumptions. A probabilistic evaluation is then applied to make predictions for leakage andejection risk generally using best-estimate inputs and assumptions, with uncertaintiestreated using statistical distributions.The deterministic crack growth evaluation provides a precursor to the probabilisticevaluation to directly illustrate the relationship between the improved PWSCC growthresistance of Alloys 690/52/152 and the time to certain adverse conditions. Theseevaluations apply conservative CGR predictions and the assumption of an existing flaw(which is replaced with a PWSCC initiation model for probabilistic evaluation). Theevaluations provide a reasonable lower bound on the time to adverse conditions, fromwhich a conservative inspection interval may be recommended. This evaluation draws fromvarious EPRI MRP and industry documents which evaluate, for Alloys 600/82/182, the timefrom a detectable flaw being created to leakage occurring and from a leaking flaw to thetime that net section collapse (nozzle ejection) would be predicted to occur.

United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)Applying a conservative crack growth FOI of 20 to circumferential and inside diameter (ID)axial cracking and of 10 to outside diameter (OD) axial cracking for Alloys 690/52/152versus Alloys 600 and 182, the results show that more than 20 years is required forleakage to occur and that more than 120 years would be required to reach the critical cracksize subsequent to leakage. The probabilistic model in EPRI MRP-375 was developed topredict PWSCC degradation and its associated risks in RVCHs.The model utilized in this probabilistic evaluation is modified from the model presented inAppendix B of EPRI MRP-335, Rev. 1 (Ref. 5) that evaluated surface stress improvementof Alloy 600 RVCHs for surface stress improvement. The integrated probabilistic model inEPRI MRP-375 includes submodels for simulating component and crack stress conditions,PWSCC initiation, PWSCC growth, and flaw examination. The submodels for crackinitiation and growth prediction for Alloy 600 reactor pressure vessel head penetrationnozzles (RPVHPNs) in MRP-335, Rev. 1 were adapted for Alloy 690 RVCHs by applyingFOls to account for its superior PWSCC resistance. The probabilistic calculations arebased on a Monte Carlo simulation model including PWSCC initiation, crack growth, andflaw detection via ultrasonic testing. The average leakage frequency and average ejectionfrequency were determined using conservative FOI assumptions. The results show thatusing only modest FOls for Alloys 690/52/152 RVCHs, the potential for developing a safetysignificant flaw (risk of nozzle ejection) is acceptably small for a volumetric/surfaceexamination period of 20 years.The evaluations performed in EPRI MRP-375 were prepared to bound all PWRreplacement RVCH designs that are manufactured using Alloy 690 base material and Alloy52/152 weld materials. The evaluations assume a bounding continuously operating RVCHtemperature of 613°F and a relatively large number of RVCH penetrations (89).While Duke Energy is not requesting NRC review and approval of EPRI MRP-375 toapprove this request for alternative, the insights gained in this technical report helpsubstantiate the limited extension duration being requested for RNP of approximately 3years beyond the 10 year examination frequency established in ASME Code Case N-729-1. In particular, the tabulation of CGR data for Alloys 690/52/152 (Section 3 of EPRI MRP-375) and review of inspection experience for Alloys 690/52/152 plant components (Section2 of EPRI MRP-375) are sufficient to demonstrate the acceptability of the limited extensionduration being requested. This request is not dependent on the more detailed probabilisticcalculations presented in Section 4 of EPRI MRP-375.RNP Unit 2 RVCH Design and OperationThe analysis performed by EPRI MRP-375 bounds the design and operation of the RNPreplacement RVCH. The RVCH contains forty-seven (47) nozzle penetrations of whichforty-five (45) are used for control element drive mechanisms (CEDMs), and two (2) smalldiameter penetrations near the center of the RVCH are used for the Reactor Head Vent(RHV) and Reactor Vessel Level Indication System (RVLIS). The Replacement RVCH wasmanufactured by Mitsubishi and placed in service in October 2005. The replacement RVCHwas manufactured as a single forging which eliminated the center disc and flangecircumferential weld in the original RNP RVCH. The replacement RVCH is fabricated fromSA-508, Grade 3, Class 1 steel and clad with an initial layer of 309 L stainless steelfollowed by subsequent layers of 308 L stainless steel.

United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)The nozzle housing penetrations on the replacement RVCH are fabricated from InconelSB-1 67 (Alloy 690) UNS N06690 and the vent pipe was made from SB-1 67 (Alloy 690) andSA-312 Type 316.The nozzle J-groove welds utilized ERNiCrFe-7 (UNS N06052) andENiCrFe-7 (UNS W86152) weld materials.A preservice volumetric examination of the RNP replacement RVCH J-groove weldedCEDM, RHV and RVLIS nozzles was performed by Westinghouse prior to installation. Thevolumetric examinations included scanning the nozzles to the fullest extent possible, fromthe end of the nozzle to a minimum of 2 inches above the root of the J-groove weld on theuphill side. There were no ultrasonic examination (UT) responses indicative of planar flawsidentified during the volumetric examinations. Additionally, a preservice eddy currentexamination of the CEDM, RHV and RVLIS nozzle welds was performed. There were noresponses indicative of planar flaws identified during the eddy current examinations.A bare metal visual examination was performed in 2010 of the RNP replacementRVCH in accordance with ASME Code Case N-729-1, Table 1, Item B4.30. This visualexamination was performed by visual examination (VT-2) qualified examiners on the outersurface of the RVCH including the annulus area of the penetration nozzles. Thisexamination did not reveal any surface or nozzle penetration boric acid that would beindicative of nozzle leakage. This examination will be performed again in the upcoming29th refueling outage scheduled to commence in May 2015.The EPRI MRP-375 analyses assume a reactor vessel head operating temperature of613°F to bound the known RV head temperatures of all PWRs currently operating. Thenominal operating hot leg temperature for RNP is 604.1°F. Core bypass flow is expected toreduce the upper head temperature by approximately 4.35°F, which would result in anaverage RVCH temperature of approximately 599.750F. Based on this, the RNP RVCHaverage operating temperature (which is the measure of temperature relevant to potentialPWSCC degradation) is bounded by the EPRI MRP-375 evaluation results, which assumes613°F for its main deterministic and probabilistic calculations.FOI Implied by Inspection PeriodDuke Energy has also assessed the representative Alloy 690/52/152 FOI for the requestedRNP extension period for comparison with the full set of laboratory CGR data. ASME CodeCase N-729-1 is based upon conclusions reached that a head with Alloy 600 nozzles andoperating at a temperature of 605°F is safe to operate up to 2 years (one 24 monthoperating cycle) between volumetric/surface examinations. The same period for Alloy 690RVCHs in N-729-1 is 10 years which represents a factor of 5 over the Alloy 600 RVCHs. Asimple extension of that improvement factor to 13 years would be a factor of 6.5 for theproposed period between volumetric/surface examinations for RNP.However, the RVCH operating temperature assumed in the technical basis for heads withAlloy 600 nozzles (References 3, 6, & 7) for ASME Code Case N-729-1 was 605.0 F,compared to an assumed operating temperature of 599.75°F for RNP. Code Case N-729-1addresses the effect of differences in operating temperature on the requiredvolumetric/surface re-examination interval for heads with Alloy 600 nozzles on the basis ofthe re-inspection years (RIY) parameter. The RIY parameter adjusts the effective full poweryears (EFPYs) of operation between inspections for the effect of head operatingtemperature using the thermal activation energy appropriate to PWSCC crack growth.

United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)For heads with Alloy 600 nozzles, ASME Code Case N-729-1 as conditioned by1OCFR50.55a limits the interval between subsequent volumetric/surface inspections toRIY = 2.25. The RIY parameter, which is referenced to a head temperature of 600°F, limitsthe time available for potential crack growth between inspections. As discussed in thetechnical basis documents for heads with Alloy 600 nozzles, effective time for crack growthis the principal basis for setting the appropriate reexamination interval to detect anyPWSCC in a timely fashion. U.S. PWR inspection experience for heads with Alloy 600nozzles has confirmed that the RIY = 2.25 interval results in a suitably conservativeinspection program.There have been no reports of nozzle leakage or of safety significant circumferentialcracking subsequent to the time that the Alloy 600 nozzles in a head were.first examined bynon-visual inservice non-destructive examination for plants conforming to the 2.25 RIYInterval (References 8 & 9).The representative RNP RVCH operating temperatures of 599.75°F would result in an RIYtemperature adjustment factor of 0.994 (versus the reference temperature of 600.0 F) usingthe activation energy of 31 kcal/mol for crack growth of ASME Code Case N-729-1.Laboratory PWSCC crack growth rate testing for Alloy 690 wrought material by multipleinvestigators (References 10, 11, & 12) has shown thermal activation energy valuescomparable to the standard activation energy applied to model growth of Alloys 600/82/182(31 kcal/mol or 130 kJ/mol). Thus, it is appropriate to apply this standard activation energyfor modeling crack growth of Alloy 690/52/152 plant components. Conservatively assumingthat the EFPYs of operation accumulated at RNP since RVCH replacement is equal to thecalendar years since replacement, the RIY for the requested extended period at RNPwould be (0.994) x (13 years) = 12.92 RIY. The FOI implied by this RIY value for RNP is(12.92)/(2.25) = 5.7 FOI. Considering the statistical compilation of data provided in Figures3-2, 3-4, and 3-6 of EPRI MRP-375, this factor of improvement is conservatively less thanthe FOI of 10 that bounds the crack growth rate data presented. Furthermore, as discussedin Sections 2 and 3 of EPRI MRP-375, PWR plant experience and laboratory testing havedemonstrated a large improvement in resistance to PWSCC initiation of Alloys 690/52/152in comparison to that for Alloys 600/82/182. Hence, the demonstrated improvements inPWSCC initiation and growth confirm on a conservative basis the acceptability of thelimited requested period of extension.ConclusionsDuke Energy believes that the Alloy 690 nozzle base and Alloy 52/152 weld materials usedin the RNP replacement RVCH provide for a clearly superior reactor coolant systempressure boundary where the potential for PWSCC has been shown by analysis and byyears of positive industry experience to be remote. This is further supported by visualexamination of the RNP RVCH in 2010 and the volumetric examinations performed byother Westinghouse designed plants during their nominal 10-year examination undersimilar operating conditions which did not reveal PWSCC.The FOI implied by the requested extension period represents a level of reduction inPWSCC crack growth rate versus that for Alloys 600/82/182 that is completely bounded ona statistical basis by the laboratory data compiled in EPRI MRP-375. Given the lack ofPWSCC detected to date in any PWR plant applications of Alloys 690/52/152, the simpleFOI assessment clearly supports the limited requested period of extension.

United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)Therefore, the RNP RVCH FOI corresponding to the requested period of extension toperform a volumetric/ surface examination provides an acceptable level of quality andsafety in accordance with 1 OCFR50.55a(a)(3)(i).6. Duration of Proposed AlternativeThe proposed alternative is requested for the duration up to and including the 31st RNPrefueling outage that is schedule to commence in September of 2018 and which will occurin the fifth ten-year ISI inspection interval which began July 22, 2012 and ends July 30,2021.7. PrecedentsML14118A477 -Request for Alternative from Volumetric/Surface Examination FrequencyRequirements of ASME Code Case N-729-1, Arkansas Nuclear One, Unit 1 -Currentlyunder NRC review.ML14206A939 -Request for Alternative from Volumetric/Surface Examination FrequencyRequirements of ASME Code Case N-729-1, St Lucie Unit 1 -Currently under NRC review.8. References1 ASME Code Case N-729-1, "Alternative Examination Requirements for PWR ReactorVessel Upper Heads With Nozzles Having Pressure-Retaining Partial-PenetrationWelds,Section XI, Division 1," Approved March 28, 2006.2 EPRI MRP-1 11, "Resistance to Primary Water Stress Corrosion Cracking of Alloys 690,52, and 152 in Pressurized Water Reactors," Report No. 1009801, March 2004(ML041680546).3 EPRI MRP-1 10, "Reactor Vessel Closure Head Penetration Safety Assessment for U.S.PWR Plants," Report No. 1009807, April 2004 (ML041680506).4 EPRI MRP-375, "Technical Basis for Reexamination Interval Extension for Alloy 690PWR Reactor Vessel Top Head Penetration Nozzles", Report No. 3002002441,February 2014 (publically available at www.epri.com)5 EPRI MRP-335 (Rev. 1), "Topical Report for Primary Water Stress Corrosion CrackingMitigation by Surface Stress Improvement," Report No. 3002000073, January 2013.6 EPRI MRP-117, "Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S.PWR Plants," Report No. 1007830, December 2004 (ML043570129).7 EPRI MRP-105, "Probabilistic Fracture Mechanics Analysis of PWR Reactor PressureVessel Top Head Nozzle Cracking," Report No. 1007834, April 2004 (ML041680489).8 EPRI MRP Letter 2011-034, "Tcold RV Closure Head Nozzle Inspection ImpactAssessment," dated December 21, 2011 (ML1 2009A042)9 G. White, V. Moroney, and C. Harrington, "PWR Reactor Vessel Top Head Alloy 600CRDM Nozzle Inspection Experience," presented at EPRI International BWR and PWRMaterial Reliability Conference, National Harbor, Maryland, July 19, 2012.10 U.S. NRC, "Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld inSimulated PWR Environment- 2009," NUREG/CR-7137, ANL-10/36, published June2012 (ML1 2199A41 5).11 EPRI MRP-237 (Rev. 2), "Resistance of Alloys 690, 152, and 52 to Primary WaterStress Corrosion Cracking: Summary of Findings Between 2008 and 2012 fromCompleted and Ongoing Test Programs," Report No. 3002000190, April 2013(publically available at www.epri.com)

United States Nuclear Regulatory CommissionEnclosure to Serial: RNP-RA/14-00929 Pages (including cover sheet)12 M. B. Toloczko, M. J. Olszta, and S. M. Bruemmer, "One Dimensional Cold RollingEffects on Stress Corrosion Crack Growth in Alloy 690 Tubing and Plate Materials,"15th International Conference on Environmental Degradation of Materials in NuclearPower Systems -Water Reactors, TMS (The Minerals, Metals & Materials Society)