L-2017-026, Inservice Inspection Plans Fourth Ten-Year Interval Unit 1 Relief Request 12

Inservice Inspection Plans Fourth Ten-Year Interval Unit 1 Relief Request 12

ML17045A357
Person / Time
Site:	Saint Lucie
Issue date:	02/14/2017
From:	Synder M Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2017-026
Download: ML17045A357 (46)

Text

• F=PL.

FEB 1 4 2017 L-2017-026 10 CFR 50.4 10 CFR 50.55a U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Re: St. Lucie Unit 1 Docket No. 50-335 In-Service Inspection Plans Fourth Ten-Year Inte1-val Unit 1 Relief Request 12 Pursuant to 10CFR50.55a(z)(1), Florida Power & Light (FPL) is requesting relief from from certain requirements of the ASME Section XI, Code Case N -729-1 for examination of the reactor vessel closure head (RVCH) at St. Lucie (PSL) unit 1. The relief request provides a proposed alternative to the ASME Boiler & Pressure Vessel Code, Code Case N-729-1 examination frequency for the 4th inset-vice inspection inte1-val.

The justification for this relief is contained in Attachment 1, Attachment 2 provides a calculation summary for the minimum factor of improvement (FOI) on crack growth, and the Dominion Engineering Report, TN-5696-00-02, contained in Attachment 3 provides further support for the requested alternative inspection intet-val based PWSCC crack growth rate data, the FOI approach, and addresses requests for additional information that the NRC has transmitted to other licensees in the context of similar relief requests. it,l!

Please contact Ken Frehafer at 772-467-77 48 if there are any questions about this submittal.

Michael J. Snyder Licensing Manager St. Lucie Plant Attachments MJS/KWF cc: USNRC Regional Administrator, Region II USNRC Senior Resident Inspector, St. Lucie Units 1 and 2 Florida Power & Light Company 6501 S. Ocean Drive, Jensen Beach, FL 34957

L-2017-026 Attachment 1 Page 1 of 9 Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety Fourth Ten-Year Interval Unit 1 Relief Request 12

1. ASME Code Component(s) Affected The affected components are ASME Class 1 Pressurized Water Reactor (PWR)

Reactor Vessel Upper Head (Closure Head) (RVCH) nozzles and partial-penetration welds fabricated with primary water stress corrosion cracking (PWSCC)-resistant materials. St. Lucie Unit 1 penetration tubes and vent pipe are fabricated from Alloy 690 with alloy 52/152 attachment welds.

2. Applicable Code Edition and Addenda

The 4th ISI interval Code of record for St. Lucie Unit 1 is the 2001 Edition with 2003 Addenda of ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components.

Examination of the reactor vessel closure head (RVCH) penetrations are performed in accordance with 10 CFR50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-1, with conditions.

3. Applicable Code Requirement

The Code of Federal Regulations 10CFR50.55a(g)(6)(ii)(D)(1), requires (in part):

All licensees of pressurized water reactors shall augment their inservice inspection program with ASME Code Case N-729-1subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section. Licensees of existing operating reactors as of September 10, 2008 shall implement their augmented inservice inspection program by December 31, 2008.

10CFR50.55a(g)(6)(ii)(D)(3) conditions ASME Code Case N-729-1 [1] by stating:

Instead of the specified examination method requirements for volumetric and surface examinations in Note 6 of Table 1 of Code Case N-729-1, the licensee shall perform volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent 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-groove welds shall be performed. If a surface examination is being substituted for a volumetric examination 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 the inside and outside wetted surface of the penetration nozzle not examined volumetrically.

ASME Code Case N-729-1, -2410 specifies that the reactor vessel upper head penetrations (nozzles and partial-penetration welds) shall be examined on a

L-2017-026 Attachment 1 Page 2 of 9 frequency in accordance with Table 1 of this code case. The basic inspection requirements of Code Case N-729-1, as amended by 10 CFR 50.55a, for partial-penetration welded Alloy 690 head penetration nozzles are as follows:

Volumetric or surface examination of all nozzles, not to exceed one inspection interval (nominally 10 calendar years) provided that flaws attributed to primary water stress corrosion cracking (PWSCC) have not been identified.

Direct visual examination (VE) of the outer surface of the head for evidence of leakage every third refueling outage or 5 calendar years, whichever is less.

4. Reason for Request

Code Case N-729-1 [1] as conditioned by 10 CFR 50.55a(g)(6)(ii)(D) requires volumetric and/or surface examination of the RVCH penetration nozzles and associated welds no later than nominally 10 calendar years after the head was placed into service. This examination schedule was intended to be conservative and subject to reassessment once additional laboratory data and plant experience on the performance of Alloy 690 and Alloy 52/152 weld metals became available [2]. Using plant and laboratory data, Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) Report MRP-375 was developed to support a technically based volumetric or surface re-examination interval using appropriate analytical tools. This technical basis demonstrates that the re-examination interval can be extended to a 20 year interval length while maintaining an acceptable level of quality and safety. FPL is requesting approval of this alternative to allow the use of the ISI interval extension for the St Lucie Unit 1 Alloys 690/52/152 reactor vessel closure head penetrations.

5. Proposed Alternative and Basis for Use Florida Power and Light Company (FPL) is requesting relief from the exam frequency requirements of Code Case N-729-1 [1], Item B4.40 for performing volumetric and/or surface exams of the St Lucie Unit 1 RVCH penetrations not to exceed one inspection interval (nominally 10 calendar years). The St. Lucie Unit 1 replacement RVCH was placed into service when the unit started up during December 2005.

Specifically, this would allow volumetric or surface examinations currently scheduled for the Spring 2018 refueling outage to be extended to the Spring 2021 refueling outage (approximately, 15.5 calendar years from installation). This request applies to the Item B4.40 inspection frequencies only.

As discussed in the original ASME technical basis document [2], the inspection frequency of ASME Code Case N-729-1 [1] for heads with Alloy 690 nozzles and Alloy 52/152 attachment welds is based, in part, on the analysis of laboratory and plant data presented in report MRP-111 [3], which was summarized in the safety assessment for RVCHs in MRP- 110 [4]. The material factor of improvement (FOI) for primary water stress corrosion cracking (PWSCC) of Alloy 690 materials over that of mill-annealed Alloy 600 material was shown by this report to be on the order of 26 or greater. The current inspection regime was established in 2004 as a conservative approach and was intended to be subject to reassessment upon the availability of

L-2017-026 Attachment 1 Page 3 of 9 additional laboratory data and plant experience on the performance of Alloy 690 and Alloy 52/152 [2].

Further evaluations were performed to demonstrate the acceptability of extending the inspection intervals for Code Case N-729-1, Item B4.40 components and documented in MRP-375 [5]. In summary, the basis for extending the intervals from once each interval (nominally 10 calendar years) to once every second interval (nominally 20 calendar years) is based on plant service experience, factor of improvement studies using laboratory data, deterministic study results, and probabilistic study results.

Per MRP-375, much of the laboratory data indicated a factor of improvement (FOI) of 100 for Alloys 690/52/152 versus Alloys 600/182/82 (for equivalent temperature and stress conditions) in terms of crack growth rates (CGRs). In addition, laboratory and plant data demonstrate a factor of improvement (FOI) in excess of 20 in terms of the time to PWSCC initiation. This reduced susceptibility to PWSCC initiation and growth supports elimination of all volumetric exams throughout the plant service period. However, since work is still ongoing to determine the performance of Alloys 690/52/152 metals, the determination of the proposed inspection interval is based on conservatively smaller factors of improvement.

Deterministic calculations demonstrate that the alternative volumetric reexamination schedule is sufficient to detect any PWSCC before it could develop into a safety significant circumferential flaw that approaches the large size (i.e., more than 300°)

necessary to produce a-nozzle ejection. The deterministic calculations also demonstrate that any base metal PWSCC would likely be detected prior to a through-wall flaw occurring. Probabilistic calculations based on a Monte Carlo simulation model of the PWSCC process, including PWSCC initiation, crack growth, and flaw detection via ultrasonic testing, show a substantially reduced effect on nuclear safety compared to a head with Alloy 600 nozzles examined per current requirements.

Service Experience As documented in MRP-375, the resistance of Alloy 690 and corresponding weld metals Alloy 52 and 152 is demonstrated by the lack of any PWSCC indications reported in these materials, in up to 24 calendar years of service for thousands of Alloy 690 steam generator tubes, and more than 22 calendar years of service for thick-wall and thin-wall Alloy 690 applications. This excellent operating experience includes service at pressurizer and hot-leg temperatures and includes Alloy 690 wrought base metal and Alloy 52/152 weld metal. This experience includes ISI volumetric or surface examinations performed in accordance with ASME Code Case N-729-1 on 16 of the 40 replacement RVCHs currently operating in the U.S. fleet.

This data supports a factor of improvement in time of at least 5 to 20 to detectable PWSCC when compared to service experience of Alloy 600 in similar applications.

Two of the replacement heads that were volumetrically examined in accordance with N-729-1 were Turkey Point Units 3 and 4, owned by FPL. The Turkey Point heads were replaced in 2004 and 2005 respectively, and examined during their 2014 refueling outages. The St. Lucie Unit 1 head and the Turkey Point Units 3 & 4 head

L-2017-026 Attachment 1 Page 4 of 9 were fabricated by the same manufacturer (AREVA), using thermally treated Alloy 690 nozzle material produced by the same material supplier (Valinox Nucleaire), per the same ASME SB-167 nozzle material specifications with identical supplemental requirements as the previously examined Turkey Point Units 3 and 4 heads. The nozzle J-groove attachment welds for the Turkey Point and St. Lucie Heads utilized PWSCC resistant ERNiCrFe-7 (UNS N06052 and/or ENiCrFe-7 UNS W86152) weld materials. The St. Lucie Unit 1 and Turkey Point Units 3 & 4 were all procured to ASME Section III, 1989 Edition, no addenda. As stated above, none of the prior examinations of replacement RVCHs with Alloy 690 nozzles has revealed any indications of PWSCC or service-induced cracking.

Factors of Improvement (FOI) for Crack Initiation Alloy 690 is highly resistant to PWSCC due to its approximate 30% chromium content. Per MRP-115 [6], it was noted that Alloy 82 CGR is 2.6 slower than Alloy 182. There is no strong evidence for a difference in Alloy 52 and 152 CGRs.

Therefore data used to develop factors of improvement for Alloy 52/152 were referenced against the base case Alloy 182, as Alloy 182 is more susceptible to initiation and growth when compared to Alloy 82. A simple factor of improvement approach was applied in a conservative manner in MRP-375 using multiple data. As discussed in MRP-375, laboratory and plant data demonstrate a factor of improvement in excess of 20 in terms of the time to PWSCC initiation.

Conservatively, credit was not taken for the improved resistance of Alloys 690/52/152 to PWSCC initiation in the main MRP-375 analyses.

Factors of Improvement (FOI) for Crack Growth MRP-375 also assessed laboratory PWSCC crack growth rate data for the purpose of assessing FOI values for growth. Data analyzed to develop a conservative factor of improvement include laboratory specimens with substantial levels of cold work. It is important to note that much of the data used to support Alloy 690 CGRs was produced using materials with significant amounts of cold work, which tends to increase the CGR. Similar processing, fabrication, and welding practices apply to the original (Alloy 600) and replacement (Alloy 690) components. MRP-375 considered the most current worldwide set of available PWSCC CGR data for Alloys 690/52/152 materials.

Figure 3-2 of MRP-375, compares data from Alloy 690 specimens with less than 10% cold work and the statistical distribution from MRP-55 [7] describing the material variability in CGR for Alloy 600. Most of the laboratory comparisons were bounded by a factor of improvement of 20, and all were bounded by a factor of improvement of 10. Most data support a FOI of much larger than 20. This is similar for testing of the Alloy 690 Heat Affected Zone (HAZ) as shown in Figure 3-4 of MRP-375 (relative to the distribution from MRP-55) and for the Alloy 52/152 weld metal (relative to the distribution from MRP-115 [6]) as shown in Figure 3-6 of MRP-375. Based on the data, it is conservative to assume a FOI of between 10 and 20 for CGRs.

Note that for a head with Alloy 600 nozzles and Alloy 82/182 attachment welds operating at a temperature of 605°F, the reinspection years (RIY) = 2.25 constraint

L-2017-026 Attachment 1 Page 5 of 9 on the volumetric or surface reexamination interval of ASME Code Case N-729-1 correspond to an interval of approximately 2.0 EFPYs or a 2 year operating cycle.

Thus, a nominal interval of 15.5 calendar years for the St. Lucie Unit 1 replacement head implies a FOI of 7.35 (see Attachment 2) versus the standard interval for heads with Alloy 600 nozzles. It is emphasized that the FOI of 7.35 implied by the requested extension period represents a level of reduction in PWSCC crack growth rate versus that for Alloys 600/82/182 that is completely bounded by the laboratory data compiled in EPRI MRP-375 when material variability is accounted for. Given the lack of PWSCC detected to date in any PWR plant applications of Alloys 690/52/152, the simple FOI assessment clearly supports the requested period of extension. , Dominion Engineering Inc. Technical Note TN-5696-00-02 Rev 0, provides further support for the requested alternative inspection interval based on the available laboratory PWSCC crack growth rate data and the FOI approach and addresses requests for additional information that NRC has transmitted to other licensees in the context of similar relief requests (see Section 7 Precedent). describes the materials tested for data points that are located above the curves that are a factor of 12 below the MRP-55 [7] and MRP-115 [6] crack growth rate curves for the 75th percentile of material variability. As Attachment 3 discusses data points above the curves that are a factor of 12 below the MRP-55 and MRP-115 curves, the discussion bounds the needed FOI of 7.35 for this St.

Lucie Unit 1 inspection interval extension. also identifies that much of the Alloy 690 CRDM material included in the MRP-375 data compilation was supplied by the Valinox Nucleaire, the same material supplier for the St. Lucie Unit 1 head (and the previously examined Turkey Point Unit 3 & 4 heads). Therefore this crack growth rate data and the FOI approach are applicable as a basis for the St. Lucie Unit 1 requested frequency extension. It is concluded that the available crack growth rate data do not indicate any susceptibility concerns specific to the nozzle or weld materials of the St. Lucie Unit 1 replacement head.

Design Features Further Increasing the Resistance of the St. Lucie Unit 1 Replacement Head to PWSCC In addition to the standard Alloy 690 materials (plate and CRDM nozzle material) test data reported in MRP-375, FPL imposed supplemental requirements on the St. Lucie Unit 1 nozzle materials (identical to and the previously examined Turkey Point Unit 3

& 4 heads) to increase the material resistance to PWSCC. These supplemental requirements included; thermal treatment (TT), prohibition of cold straightening after TT, ingot remelting to reduce impurities, additional chemistry requirements, microstructure and grain size requirements. These methods substantially reduce PWSCC susceptibility beyond that assumed in the generic MRP-375 study, resulting in additional assurance that the St. Lucie Unit 1 head can be operated for 15.5 years from replacement prior to their next volumetric and/or surface examination with an acceptable level of quality and safety.

Previous Examinations of the St. Lucie Unit 1 Replacement Head

L-2017-026 Attachment 1 Page 6 of 9 A preservice volumetric examination of the replacement RVCH partial-penetration welded nozzles was performed prior to head installation that went into service with the start up in 2005 at St Lucie Unit 1. There were no recordable indications identified during the preservice volumetric examinations of the nozzle tube in the area of the J-groove welds. A bare metal visual examination (VE) was performed of the St Lucie Unit 1 replacement RVCH in 2010 and 2015 in accordance with ASME Code Case N-729-1, Table 1, Item B4.30. This visual examination was performed by VT-2 qualified examiners on the outer surface of the RVCH including the annulus area of the penetration nozzles. This examination did not reveal any surface or nozzle penetration boric acid that would be indicative of nozzle leakage.

Deterministic Modeling A deterministic crack growth evaluation is commonly applied to assess PWSCC risks for specific components and operating conditions. The deterministic evaluation is intended to demonstrate the time from an assumed initial flaw to some adverse condition. Deterministic crack modeling results were presented in MRP-375 for previous references in which both growth of part-depth surface flaws and through-wall circumferential flaws were evaluated and normalized to an adjusted growth of 613 degrees Fahrenheit (°F) to bound the PWR fleet. The time for through-wall crack growth in Alloy 600 nozzle tube material, when adjusted to a bounding temperature of 613°F, ranged between 1.9 and 3.8 Effective Full Power Years (EFPY). Assuming a growth FOI of 10 to 20 as previously established for Alloys 690/52/152 materials, the median time for through-wall growth was 37.3 EFPY. In a similar manner, crack growth results for through-wall circumferential flaws were tabulated and adjusted to a temperature of 613°F. Applying a growth FOI of 20 resulted in a median time of 176 EFPYs for growth of a through-wall circumferential flaw to 300 degrees of circumferential extent.

The results of the generic evaluation are summarized in Table 4-1 of MRP-375. All cases were bounding and support an inspection interval greater than is being proposed. It is important to note that the upper head operating temperature of the St Lucie Unit 1 is 602.6°F and the FOI for the extension to a 15.5 year examination frequency is 7.35 and is well within the bounds of the assumptions.

Deterministic calculations performed in MRP-375 demonstrate that the alternative volumetric re-examination interval is sufficient to detect any PWSCC before it could develop into a safety significant circumferential flaw that approaches the large size necessary to produce a nozzle ejection. The deterministic calculations also demonstrate that any base metal PWSCC would likely be detected prior to a through-wall flaw occurring.

Probability of Cracking or Through-Wall Leaks Probabilistic calculations are based on a Monte Carlo simulation model of the PWSCC process, including PWSCC initiation, PWSCC crack growth, and flaw detection via ultrasonic testing and visual examinations for leakage. The basic structure of the probabilistic model is similar to that used in the MRP-105 [8]

technical basis report for inspection requirements for heads with Alloy 600 nozzles,

L-2017-026 Attachment 1 Page 7 of 9 but the current approach includes more detailed modeling of flaw initiation and growth (including multiple flaw initiation for each nozzle on base metal and weld surfaces), and the initiation module has been calibrated to consider the latest set of experience for U.S. heads. The outputs of the probabilistic model are leakage frequency (i.e., frequency of through-wall cracking) and nozzle ejection frequency.

Even assuming conservatively small factors of improvement for the crack growth rate for the replacement nickel-base alloys (with no credit for improved resistance to initiation), the probabilistic results with the alternative inspection regime show:

1) An effect on nuclear safety substantially within the acceptance criterion applied in the MRP-117 [9] technical basis for Alloy 600 heads,

2) And a substantially reduced effect on nuclear safety compared to that for a head with Alloy 600 nozzles examined per current requirements.

Furthermore, the results confirm a low probability of leakage if modest credit is taken for improved resistance to PWSCC initiation compared to that for Alloys 600 and 182.

Conclusion In summary, the basis for extending the intervals from once each interval (nominally 10 calendar years) to a 15.5 year examination frequency is based on plant service experience, factor of improvement studies using laboratory initiation and growth data, deterministic modeling, and probabilistic study results. The results of the analysis show that the alternative proposed extension to a 15.5 year examination frequency results in a substantially reduced effect on nuclear safety when compared to a head with Alloy 600 nozzles and examined per the current requirements. The minimum FOI of 7.35 implied by the requested extension to a 15.5 year examination frequency period represents a level of reduction in PWSCC crack growth rate versus that for Alloys 600/82/182 that is completely bounded by the laboratory data compiled in MRP-375 when accounting for heat-to-heat variability of Alloy 600 and weld-to-weld variability of Alloy 82/182/132. The proposed revised interval will continue to provide reasonable assurance of structural integrity.

Additional assurance of structural integrity is provided by the design features of the St Lucie Unit 1 replacement head such as thermal treatment of nozzle material and by the 2014 indication free inspection results of the Turkey Point Unit 3 and 4 heads with similar material. Furthermore, the visual examinations and acceptance criteria as required by Item B4.30 of Table 1 of ASME Code Case N-729-1 are not affected by this request and will continue to be performed on a frequency of every third refueling outage or 5 calendar years, whichever is less. As discussed in Section 5.2.3 of MRP-375, the visual examination requirement of the outer surface of the head for evidence of leakage supplements the volumetric and/or surface examination requirement and conservatively addresses the potential concern for boric acid corrosion of the low-alloy steel head due to PWSCC leakage.

For the reasons noted above, it is requested that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1) as the alternative provides an acceptable level of quality and safety.

L-2017-026 Attachment 1 Page 8 of 9

6. DURATION OF PROPOSED ALTERNATIVE:

The proposed Alternative is requested for the remainder of the Fourth Inservice Inspection Interval which began 2/11/2008 and ends 2/10/2018 and the Fifth Inservice Inspection Interval which will begin 2/11/2018 and end 2/10/2028. Utilizing the proposed alternative will require the examination to be performed in the fifth interval during the PSL1-30 refueling outage which will commence in Spring 2021.

7. PRECEDENTS:

There have been submittals from multiple plants to request an alternative from the frequency of ASME Code Case N-729-1 for volumetric or surface examinations of heads with Alloy 690 nozzles. The first of these was Arkansas Nuclear One, Unit 1, and some subsequent requests including the associated status at the time of submittal of this request are shown below:

NRC ADAMS Accession No.

Request for Additional Relief Information RAI NRC Safety Plant Request (RAI) Response Evaluation Status Arkansas Accepted Nuclear One, ML14118A477 ML14258A020 ML14275A460 ML14330A207 12/23/14 Unit 1 Beaver Valley, Accepted ML14290A140 ML14363A409 Unit 1 1/28/15 Calvert Cliffs ML15009A035 Accepted ML15327A367 Unit 1 & 2 ML15201A067 12/7/15 Comanche Accepted ML15120A038 ML15259A004 Peak Unit 1 10/30/15 D.C. Cook Accepted ML15023A038 ML15156A906 Units 1 & 2 6/11/15 J.M. Farley, ML14280A260 Accepted ML15104A192 Unit 2 ML15111A387 5/5/15 North Anna, Accepted ML14283A044 ML15091A687 Unit 2 6/25/15 Prairie Island, Accepted ML14258A124 ML15030A008 ML15036A252 ML15125A361 Units 1 and 2 6/4/15 H.B. Robinson, Accepted ML14251A014 ML14294A587 ML14325A693 ML15021A354 Unit 2 2/17/15 Accepted Salem Unit 1 ML15098A426 ML15349A956 12/24/15 St. Lucie, Accepted ML14206A939 ML14251A222 ML14273A011 ML14339A163 Unit 1 12/23/14 St. Lucie, Accepted ML16076A431 ML16292A761 Unit 2 11/1/16

L-2017-026 Attachment 1 Page 9 of 9

8. References

1. ASME Code Case N-729-1, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1, Approved March 28, 2006.

2. ASME Section XI, Code Case N-729, Technical Basis Document, dated September 14, 2004.

3. Materials Reliability Program: Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111),

EPRI, Palo Alto, CA, U.S. Department of Energy, Washington, DC: 2004.

1009801. [freely available at www.epri.com; NRC ADAMS Accession No. ML041680546]

4. Materials Reliability Program: Reactor Vessel Closure Head Penetration Safety Assessment for U.S. PWR Plants (MRP-110NP), EPRI, Palo Alto, CA: 2004.

1009807-NP. [ML041680506]

5. Materials Reliability Program: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375), EPRI, Palo Alto, CA: 2014. 3002002441. [freely available at www.epri.com]

6. Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004. 1006696. [freely available at www.epri.com]

7. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55) Revision 1, EPRI, Palo Alto, CA: 2002. 1006695. [freely available at www.epri.com]

8. Materials Reliability Program: Probabilistic Fracture Mechanics Analysis of PWR Reactor Pressure Vessel Top Head Nozzle Cracking (MRP-105 NP),

EPRI, Palo Alto, CA: 2004. 1007834. [ML041680489]

9. Materials Reliability Program: Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PWR Plants (MRP-117), EPRI, Palo Alto, CA: 2004.

1007830. [freely available at www.epri.com; NRC ADAMS Accession No. ML043570129]

L-2017-026 Attachment 2 Page 1 of 4 Calculation of the Minimum Factor of Improvement Needed by Extension of Reactor Vessel Closure Head Volumetric/Surface Inspection Interval to Support Relief Requests Citing MRP-375 1 Introduction This document presents a calculation of the minimum factor of improvement (FOI) on crack growth time that is implied by an extension of the volumetric/surface inspection interval for the Alloy 690 nozzles of a replacement reactor vessel closure head (RVCH) for St. Lucie Unit 1.

This is the minimum FOI value that crack growth rate testing should demonstrate in order to directly support the requested inspection interval. ASME Code Case N-729-1 [1], which has been mandated by 10CFR50.55 a(g)(6)(ii)(D) with conditions, specifies a reexamination interval of no more than one Section XI inspection interval (nominally 10 calendar years). MRP-375 [2]

supports an extension of this interval to two Section XI inspection intervals.

2 Summary of Results This calculation is performed for St. Lucie Unit 1, which has a replacement head manufactured with Alloy 690/52/152 material, has a head operating temperature of 602.6°F, and requests a five and a half (5.5) calendar year examination interval extension beyond the 10 calendar year examination requirement of Code Case N-729-1. This analysis shows that the minimum FOI on crack growth time implied by a 15.5-year reexamination interval for St. Lucie Unit 1 (a 5.5 year extension relative to Code Case N-729-1) is 7.35.

3 Analysis ASME Code Case N-729-1 [1] addresses the effect of differences in operating temperature on the required volumetric/surface reexamination interval for heads with Alloy 600 nozzles on the basis of the Reinspection Years (RIY) parameter. The RIY parameter adjusts the effective full power years (EFPYs) of operation between inspections for the effect of head operating temperature using the thermal activation energy appropriate to PWSCC crack growth. For heads with Alloy 600 nozzles, Code Case N 729-1 as conditioned by 10CFR50.55a limits the interval between subsequent volumetric/surface inspections to RIY = 2.25. The RIY parameter, which is referenced to a head temperature of 600°F, limits the time available for potential crack growth between inspections. As discussed in the MRP-117 [3] technical basis document for heads with Alloy 600 nozzles, effective time for crack growth is the principal basis for setting the appropriate reexamination interval to detect any PWSCC in a timely fashion.

L-2017-026 Attachment 2 Page 2 of 4 The RIY parameter for heads with Alloy 600 nozzles is adjusted to the reference head temperature using an activation energy of 130 kJ/mole (31 kcal/mole) [1]. Based on the available laboratory data, the same activation energy is applicable to model the temperature sensitivity of growth of a hypothetical PWSCC flaw in the Alloy 690/52/152 material of the replacement RVCH. Key laboratory crack growth rate testing data for Alloy 690 wrought material investigating the effect of temperature are as follows:

(1) Results from ANL indicate that Alloy 690 with 0-26% cold work has an activation energy between 100 and 165 kJ/mol (24-39 kcal/mol) [4]. NUREG/CR-7137 [4] concludes that the activation energy for Alloy 690 is comparable to the standard value for Alloy 600 (130 kJ/mole).

(2) Testing at PNNL found an activation energy of about 120 kJ/mol (28.7 kcal/mole) for Alloy 690 materials with 17-31% cold work [5].

(3) Additional PNNL testing determined an activation energy of 123 kJ/mole (29.4 kcal/mole) for Alloy 690 with 31% cold work [6].

These data show that it is reasonable to assume the same crack growth thermal activation energy as was determined for Alloys 600/82/182 (namely 130 kJ/mol (31 kcal/mol)) for modeling growth of hypothetical PWSCC flaws in Alloy 690/52/152 PWR plant components.

3.1 RIY Parameter Describing the Potential for Crack Propagation The RIY parameter, which quantifies the potential for crack propagation between successive volumetric/surface examinations, is defined by ASME Code Case N-729-1 [1] as follows:

n2 Qg 1 1

RIY = EFPY j exp [3-1]

j = n1 R Thead , j Tref where:

RIY = Reinspection Years, normalized to a reference temperature of 1059.67°R (588.71°K or 600°F)

EFPYj = effective full power years accumulated during time period j Qg = activation energy for crack growth (31 kcal/mole)

R = universal gas constant (1.103x10-3 kcal/mol-°R)

Thead,j = absolute 100% power head temperature during time period j (°R

=°F+459.67)

Tref = absolute reference temperature (1059.67°R)

L-2017-026 Attachment 2 Page 3 of 4 n = number of the time periods with distinct 100% power head temperature 1 since initial head operation n1 = number of the first time period with distinct 100% power head temperature1 since time of most recent volumetric/surface NDE (or replacement) n2 = number of the most recent time period with distinct 100% power head temperature1 For conservatism, one interval using the highest head temperature was used. The RIY expression simplifies to the following assuming a single representative head temperature over the period between successive examinations:

Q 1 1 RIY = EFPY exp g [3-2]

R Thead Tref Conservatively assuming that the EFPYs of operation accumulated at St. Lucie Unit 1 since RVCH replacement is equal to the calendar years since replacement, the RIY for the requested extended period at St. Lucie Unit 1 is calculated as follows:

31 1 1

(15.5 EFPY)exp = (15.5)(1.067)

1.103x103 602.6+459.67 600+459.67 16.54 [3-3]

3.2 Factor of Improvement (FOI) Implied by RIY The FOI implied by this RIY value for St. Lucie Unit 1 (relative to the limiting RIY for heads with Alloy 600 nozzles) is calculated as the following ratio:

690,52,152 (15.5)(1.067) 16.54

=

600,82,182

= 2.25

=

2.25

= 7.35 [3-4]

This FOI value may be compared to laboratory PWSCC crack growth rate data for Alloys 690/52/152 when they are considered relative to standard statistical distributions describing the variability in the crack growth rate for Alloy 600 [7] and Alloy 182 [8]. Alloy 600 wrought 1 Head temperature at 100% power may have been changed during the life of the plant due to design changes, power uprates, etc., and the summation is over the number of distinct periods since the last volumetric/surface NDE.

L-2017-026 Attachment 2 Page 4 of 4 material is the appropriate reference for defining the FOI for Alloy 690 wrought material. As discussed in Section 3.1 of MRP-375, Alloy 182 weld metal is chosen as the reference for defining the FOI for Alloys 52 and 152 weld metals because Alloy 182 is more susceptible on average to PWSCC initiation and growth than Alloy 82 (due to the higher Cr content of Alloy 82).

Note that the temperature factor calculated above (1.067) is relatively modest such that the FOI result is relatively insensitive to the assumed activation energy. For example, an activation energy of 40 kcal/mole, the calculated FOI would be 7.49 instead of 7.35.

4 References

1. ASME Code Case N-729-1, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1, Approved March 28, 2006.

2. Materials Reliability Program: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375). EPRI, Palo Alto, CA:

2014. 3002002441. [freely downloadable at www.epri.com]

3. Materials Reliability Program Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PWR Plants (MRP-117), EPRI, Palo Alto, CA: 2004. 1007830. [freely downloadable at www.epri.com]

4. U.S. NRC, Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld in Simulated PWR Environment - 2009, NUREG/CR-7137, ANL-10/36, published June 2012.

[NRC ADAMS Accession No. ML12199A415]

5. Materials Reliability Program: Resistance of Alloys 690, 152, and 52 to Primary Water Stress Corrosion Cracking (MRP-237, Rev.2): Summary of Findings Between 2008 and 2012 from Completed and Ongoing Test Programs, EPRI, Palo Alto, CA: 2013.

3002000190. [freely downloadable at www.epri.com]

6. M. B. Toloczko, M. J. Olszta, and S. M. Bruemmer, One Dimensional Cold Rolling Effects on Stress Corrosion Crack Growth in Alloy 690 Tubing and Plate Materials, 15th International Conference on Environmental Degradation of Materials in Nuclear Power Systems - Water Reactors, TMS (The Minerals, Metals & Materials Society), 2011.

7. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55)

Revision 1, EPRI, Palo Alto, CA: 2002. 1006695. [freely downloadable at www.epri.com]

8. Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004. 1006696. [freely downloadable at www.epri.com]

EC 288005 RO Attachment 3 PSL-ENG-SESJ-17-002 RO ____

Dominion fn~ineeriny 32;;.;to tal pages TECHNICAL NOTE Assessment of Laboratory PWSCC Crack Growth Rate Data Compiled for Alloys 690, 52, and 152 with Regard to Factors of Improvement (FOI) versus Alloys 600 and 182 TN-5696-00-02 Revision 0 March 2015 Principal Investigators G. White K. Fuhr Prepared for Electric Power Research Institute, Inc.

3420 Hillview Avenue Palo Alto, CA 94303-1338 i

12100 Sunrise Valley Drive, Suite 220

• Reston, VA 20191

• PH 703.657.7300

• FX 703.657.7301

EC 288005 RO Attachment 3 PSL-ENG-SESJ-17 -002 RO 32 total pages Dominion fn~ineerin~. Inc. TN-5696-00-02, Rev. 0 RECORD OF REVISIONS Prepared by Checked by Reviewed by Approved by Rev. Description Date Date Date Date 3(~ (l.B-1<-J.+ G.A-. CJh~K--- C}. Qv.'\V 0 Original Issue

?/z-:>/IS 3/z_~/ozot$ I 3/z:t ZJ> fS (5,3 }z-1 }~o\f K. J. Fuhr M. Burkardt G. A. White G. A. White Associate Engineer , Associate Engineer Principal Engineer Principal Engineer The last revision number to reflect any changes for each section of the technical note is shown in the Table of Contents. The last revision numbers to reflect any changes for tables and figures are shown in the List of Tables and the List of Figures. Changes made in the latest revision, except for Rev. 0 and revisions which change the technical note in its entirety, are indicated by a double line in the right hand margin as shown here.

EC 288005 RO -Attachment 3 PSL-ENG-SESJ- 17-002 RO 32 total _pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 CONTENTS Last Rev.

Page Mod.

1 INTRODUCTION .............. .................. ....... ..... .... ...... .... ..... ..... ... .. ....... ............... .... ............ ... 1 0 2 DISCUSSION OF DATA POINTS FROM MRP-375 [2] ................................... .... .............. .....3 0 2.1 Data Points Above a Hypothetical12.0 Factor of Improvement Line in Figure 3-1, 3-3, and 3-5 of MRP-375 ...... ..... ............................... ....... .. .. ................ 3 0 2.2 Data Most Directly Applicable to Plant Conditions .................................................6 0 2.3 Data Specific to Argonne National Laboratory (ANL) and Pacific Northwest National Laboratory (PNNL) .. ............ .............. ................. ...... ..... ......... 8 0 2.4 Data for Alloy 690 Wrought Material Including Added Cold Work up to 20% for CRDM Nozzle and Bar Material Product Forms ....................................... 8 0 2.5 Conclusion ....... ............... ............................... ........................................ ............... 9 0 3 POTENTIAL IMPLICATIONS OF SPECIFIC CATEGORIES OF NOZZLE AND WELD MATERIALS .... .. ......... ... ... .......... .... ..... ........... ... .. .... .............. .... ..... .... .... .... .. .......... .... ..... .. .. 9 0 3.1 Potential Similarities for Laboratory Specimen Material Exhibiting a Deterministic Factor Less than 12.0 .. .. ........ .... ...... ...... .......... ......... ...... ...... ........... 9 0 3.2 Potential Implications ........ ............ .... ...... ......... ....... ................................. ...........10 0 4 REFERENCES .. .... .... .. ... ........ .. .. ... ... ... .. .... .... ............. .. .. ... .. ....................... ...... .. ... .... .......... 12 0 iii

EC 288005 RO Attachment 3 PSL-ENG-SESJ- 17-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 LIST OF F IGURES Last Rev.

Page Mod.

Figure 1. Plot of Crack Growth Rate (da/dt} versus Stress Intensity Factor (K1) for Alloy 690 Data from Plate Material Tested by CIEMA T.. .. ...... ............ ................ .... .....14 0 Figure 2. Plot of da/dt versus K1 for Alloy 690 Data from Heat WP787 .............................. 14 0 Figure 3. Plot of da/dt versus K1 for Alloy 690 Data from Heat WP142 .............................. 15 0 Figure 4. Plot of da/dt versus K1 for Alloy 690 HAZ Data from Heat WP142 ...................... 15 0 Figure 5. Plot of da/dt versus K1 for Alloy 690 HAZ Data from Plate Material Tested by CIEMAT ...... ... ...... ......... .. ..... ................... ....... ..................................................... 16 0 Figure 6. Plot of da/dt versus K1 for Alloy 152 Data from Heat WC83F8 .. ........ .................. 16 0 Figure 7. Plot of da/dt versus K1 for Alloy 152 Data from Heat WC04F6 .......... .................. 17 0 Figure 8. Plot of da/dt versus K1 for Alloy 690 Data from All Laboratories,~ 10% Cold Work, Constant Load or K1 .................................................................................. 18 0 Figure 9. Cumulative Distribution Function of Adjusted da/dt for Alloy 690 Data from All Laboratories, ~ 10% Cold Work, Constant Load or K1 .. .. .. ...................... ....... 18 0 Figure 10. Plot of da/dt versus K1 for Alloy 690 HAZ Data from All Laboratories,~ 10%

Cold Work, Constant Load or K1 .. .......... ........ ........ .............. .... .. .........................19 0 Figure 11. Cumulative Distribution Function of Adjusted da/dt for Alloy 690 HAZ Data from All Laboratories, ~ 10% Cold Work, Constant Load or K1 ...........................19 0 Figure 12. Plot of da/dt versus K1 for Alloy 52/152 Data from All Laboratories, ~ 10%

Cold Work, Constant Load or K1 ................ ............ ........................ ...... ...... ...... ... 20 0 Figure 13. Cumulative Distribution Function of Adjusted da/dt for Alloy 52/152 Data from All Laboratories, ~ 10% Cold Work, Constant Load or K1 ...... .............. ....... 20 0 Figure 14. Plot of da/dt versus Loading Hold Time (for PPU testing) or Test Segment Duration (for Constant K1/Load Testing) from Heat WP787 .............................. .. 21 0 Figure 15. Plot of da/dt versus K1 for Alloy 690 Data Produced by ANL and PNNL and Available in Reference [17]; ~ 22% Cold Work .............. ..................................... 22 0 Figure 16. Cumulative Distribution Function of Adjusted da/dt Alloy 690 Data Produced by ANL and PNNL in References [17]; ~ 22% Cold Work and Constant Load/ K1 ... ................................................................................... ....... ............ ...... 22 0 Figure 17. Plot of da/dt versus K1 for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17]; ~ 22% Cold Work ...... .. .. ...... ............ ................ 23 0 Figure 18. Cumulative Distribution Function of Adjusted da/dt Alloy 690 HAZ Data Produced by ANL and PNNL [17]; ~ 22% Cold Work and Constant Load/KJ ...... 23 0 Figure 19. Plot of da/dt versus K1 for Alloy 52/152 Data Produced by ANL and PNNL iv

EC 288005 RO Attachment 3 PSL-ENG-SESJ-17-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 Last Rev.

Page Mod.

and Available in References [17] and [18];::;; 22% Cold Work ............................ 24 0 Figure 20. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data Produced by ANL and PNNL ([17] and [18]); ::;; 22% Cold Work and Constant Load/K1...... .... ..................... ........... ................................................... .................. 24 0 Figure 21 . Plot of da/dt versus K1 for Alloy 690 Data from All Laboratories, > 10 & ::;;

20% Cold Work, CRDM and Bar Material, Constant Load or K1 Testing ............. 25 0 Figure 22. Cumulative Distribution Function of Adjusted da/dt Alloy 690 Data from All Labs,::;; 20% Cold Work, CRDM and Bar Material, Constant Load or K1 ............ 25 0 Figure 23. Plot of da/dt versus K1 for Alloy 52/152 Data from All Laboratories, > 10 & ::;;

20% Cold Work, Constant Load or K1 .. ........................................ ........ ............... 26 0 Figure 24. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data from All Laboratories, ::;; 20% Cold Work, Constant Load or K1 ........................................ 26 0 v

EC 288005 RO .Attachment 3 PSL-ENG-SESJ-17-002 RO 32. total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 ACRONYMS ANL Argonne National Laboratory ASME American Society of Mechanical Engineers AWS American Welding Society BWC Babcock & Wilcox Canada CEDM Control Element Drive Mechanism CGR Crack Growth Rate CIEMAT Centro de Investigaciones Energeticas, Medioambientales y Tecnol6gicas CRDM Control Rod Drive Mechanism CT Compact Tension DEI Dominion Engineering, Inc.

EPRI Electric Power Research Institute FOI Factor of Improvement GE-GRC General Electric Global Research Center GTAW Gas Tungsten Arc Welding HAZ Heat Affected Zone ICI In-Core Instrumentation K Stress Intensity Factor MRP Materials Reliability Program NRC Nuclear Regulatory Commission PNNL Pacific Northwest National Laboratory PPU Partial Periodic Unloading PWR Pressurized Water Reactor PWSCC Primary Water Stress Corrosion Cracking RN Re-Inspection Year RV Reactor Vessel RVCH Reactor Pressure Closure Head UNS Unified Numbering System vi

EC 288005 RO Attachment 3 PSL-ENG-SESJ-17-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 1 INTRODUCTION The purpose of this DEI technical note is to examine laboratory crack growth rate (CGR) data for primary water stress corrosion cracking (PWSCC) compiled for Alloys 690, 52, and 152 to assess factors of improvement (FOI) for these replacement alloys relative to the CGR behavior for Alloys 600 and 182 as documented in MRP-55 [1] and MRP-115 [2]. In addition, an assessment is made of the available laboratory CGR data for the potential concern of elevated CGRs for specific categories of nozzle and weld materials.

Per ASME Code Case N-729-1 [3], the volumtric inspection interval for Alloy 600 RV head nozzles is based on operating time adjusted for operating temperature using the temperature sensitivity for PWSCC crack growth. The normalized operating time between inspections, called theRe-Inspection Years (RIY) parameter, represents the potential for crack growth between successive volumtric examinations. Thus, the FOI for Alloys 690/52/152 exhibited by laboratory CGR data can be used to support appropriate volumetric inspection intervals for RV heads with Alloy 690 nozzles. On the basis of the RIY = 2.25 limit of Code Case N-729-1 for Alloy 600 RV head nozzles, an FOI of 12 corresponds to an inspection interval of 20 years for Alloy 690 R V head nozzles operating at 613 op . 1 A temperature of 613 op is expected to bound the head operating temperature for the U.S. pressurized water reactor (PWR) fleet.

As discussed in Section 3 of Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) report MRP-375 [2], a conservative approach was taken in MRP-375 to develop the factor of improvement (FOI) values describing the primary water stress corrosion cracking (PWSCC) crack growth rates applicable to Alloy 690 reactor vessel (RV) top head penetration nozzles. The crack growth rate data points presented in Figures 3-1 , 3-3, and 3-5 ofMRP-375 represent the values reported by individual researchers, without any adjustment by the authors of MRP-375 other than to normalize for the effect of temperature. The data in these figures represent essentially all of the Alloys 690, 52, and 152 data points reported by the various 1

To calculate the implied FOI for the bounding RV top head operating temperature of 613°F, the re-inspection year (RlY) parameter for a requested examination interval of20 years is compared with the N-729-1 interval for Alloy 600 nozzles ofRIY = 2.25. The representative head operating temperatures of 613 °F corresponds to an R1Y temperature adjustment factor of 1.38 (versus the reference temperature of 600°F) using the activation energy of 31 kcaVmol (130 k:J/mol) for crack groW1h of ASME Code Case N-729-1. Conservatively assuming that the effective full power years (EFPY) of operation accumulated since RV top head replacement is equal to 98% of the calendar years since replacement, the RlY for a requested extended period of20 years would be (1.38)(19.6) =

27.0. The FOI implied by this RlY value is (27.0)/(2.25) = 12.0.

EC 288005 RO Attachment 3 PSL-ENG-SESJ-17-002 RO 32 total pages Dominion fn~ineerin~. Inc TN-5696-00-02, Rev. 0 laboratories. No screening process was applied to the data on the basis of test characteristics such as minimum required crack extension or minimum required extent of transition along the crack front to intergranular cracking. Instead, an inclusive process was applied to conservatively assess the factors of improvement apparent in the data for specimens with less than 10 percent added cold work.

The approach was conservative in that no effort was made to screen out data points reflecting tests that are not applicable to plant conditions. Instead, the data were treated on a statistical basis in Figures 3-2, 3-4, and 3-6 ofMRP-375, 2 and compared to the crack growth rate variability due to material variability for Alloy 600 in MRP-55 [1] and Alloy 182 in MRP-115

[2]. A comparison between the cumulative distributions of the crack growth-rates for Alloys 690/52/152 and Alloys 600/82/182 treats the full variability in both original and replacement alloys, rather than comparing the variability of the replacement alloy against a conservative mean (75 1b percentile) growth rate for the original alloys. By considering the cumulative distributions, a fuller perspective of the improved resistance of Alloys 690/52/152 emerges where over 70% of the data in each of Figures 3-2,3-4, and 3-6 ofMRP-375 indicate a factor of improvement 3

beyond 20 and all of the data correspond to a factor of improvement of 12 or greater.

It is emphasized that the deterministic MRP-55 and MRP-115 crack growth rate equations were 1

developed not to describe bounding crack growth rate behavior but rather reflect 75 b percentile values of the variability in crack growth rate due to material variability. Twenty-five percent of the material heats (MRP-55) and test welds (MRP-115) assessed in these reports on average showed crack growth rates exceeding the deterministic equation values. Thus, the most appropriate FOI comparisons are made on a statistical basis (e.g., Figures 3-2, 3-4, and 3-6 of MRP-375). Comparing the crack growth rate for Alloys 690/52/152 versus the deterministic crack growth rate lines in Figures 3-1 , 3-3, and 3-5 ofMRP-375 represents an unnecessary compounding of conservatisms. Essentially none of the data presented lies within a statistical FOI of 12 below the MRP-55 and MRP-115 distributions of material variability. The technical basis for the inspection requirements for heads with Alloy 600 nozzles ([5], [6], [7]) are based on the full range of crack growth rate behavior, including heat-to-heat (weld-to-weld) and within-heat (within-weld) material variability factors. Thus, theRe-Inspection Year (RIY) = 2.25 inspection interval developed for heads with Alloy 600 nozzles reflects the possibility of crack 2

Figures 3-2, 3-4, and 3-6 ofMRP-375 show cumulative distribution functions of the variability in crack growth rate normalized for temperature and crack loading (i.e., stress intensity factor). Each ordinate value in the plots shows the fraction of data falling below the corresponding normalized crack growth rate. Thus, the cumulative distribution function has the benefit of illustrating the variability in crack growth rate data for a standard set of conditions.

3 Excluding data points that reflect fatigue pre-cracking conditions and are not relevant to PWSCC.

2

EC 288005 RO Attachment3 PSL-ENG-SESJ-17-002 RO 32 total pages Dominion [n~ineerin~. Inc. TN-5696-00-02, Rev. 0 growth rates being many times higher than the deterministic 75'h percentile values per MRP-55 and MR.P-115. Nevertheless, as described below, the large majority of the data points for the conditions directly relevant to plant conditions (e.g., constant load conditions) are located more than a factor of 12.0 below the deterministic (75th percentile) MRP-55 and MRP-115 equations.

2 DISCUSSION OF DATA POINTS FROM MRP-375 [2) 2.1 Data Points Above a Hypothetical 12.0 Factor of Improvement Line in Figure 3-1, 3-3, and 3-5 of MRP-375

• Figure 3-1 ofMRP-375. Figure 3-1 shows the complete set of data points compiled by the PWSCC Expert Panel organized by EPRI at the time MRP-375 was completed for Alloy 690 specimens with less than 10% added cold work. The following points are within a factor of 12.0 below the MRP-55 deterministic crack growth rate for Alloy 600:

There are 16 points within a factor of 12.0 below the MRP-55 75th percentile curve, out of a total of75 points shown in Figure 3-1 ofMR.P-375.

These data represent test segments from six distinct Alloy 690 compact tension (CT) specimens that were tested by Centro de Investigaciones Energeticas, Medioambientales y Tecnol6gicas (CIEMAT) and two that were tested by Argonne National Laboratory (ANL).

Two of the points tested by CIEMAT are from specimen 9ARB1, comprised of Alloy 690 plate material, loaded to 3 7 MPa(m) 0*5 , and tested at 340°C and 15 cc H 2/kg H 20

[8]. Both of these data are for the first half of segments that exhibited a crack growth rate that was an order of magnitude lower in the second half of the segment. A plot of crack growth rate versus crack-tip stress intensity factor (K) for the Alloy 690 data from MRP-375 for plate material tested by CIEMAT is provided here as Figure 1.

These two points have minimal implications for the requested inspection interval extension for several reasons:

As illustrated in Figure 1 and subsequent figures using open symbols, one of the two points was generated under partial periodic unloading (PPU) conditions.

As discussed below in Section 2.2, PPU conditions may result in accelerated crack growth rates that are not directly representative of plant conditions, especially for the case of alloys with relatively high resistance to environmental cracking like Alloy 690.

U.S . PWRs operate with a dissolved hydrogen concentration per EPRI guidelines in the range of25-50 cc/kg for Mode 1 operation. Testing at 15 cc/kg results in accelerated crack growth rates versus that for normal primary water due to the proximity of the Ni-NiO equilibrium line [2].

Specimens fabricated from Alloy 690 plate material are not as relevant to plant RV top head penetration nozzles as specimens fabricated from control rod drive mechanism (CRDM) I control element drive mechanism (CEDM) nozzle 3

EC 288005 RO Attachment 3 PSL-ENG-SESJ-17-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 material. CRDM and CEDM nozzles in U.S. PWRs are fabricated from extruded pipe or bar stock material. Note that term CRDM nozzle is used henceforth to refer to both CRDM and CEDM nozzles (CEDM is the terminology used by plants designed by Combustion Engineering).

The wide variability in crack growth rate within even the same testing segment indicates that significant experimental variability exists. Thus, there is a substantial possibility that a limited number of elevated growth rate data points do not reflect the true characteristic behavior of the material tested.

The remaining 11 CIEMAT points are from specimens comprised ofValinox WP787 CRDM nozzle material that was cold worked by a 20% tensile elongation (9.1%

thickness reduction) [9]. One datum was for specimen 9T3-tested at 310°C, 22 cc 05 H 2/kg H20, and 39 MPa(m) * - but was from the test period immediately following a reduction in temperature from 360°C to 310°C [9]. The next period of constant load growth had a factor of 10 lower CGR. The other 10 data are for testing at 325°C and 35 cc H2/kg H 20, and seven of these points are for PPU testing (which may accelerate growth beyond what would be expected for in-service components). Four of the data 05 are for specimens 9Tl and 9T2 (loaded to roughly 36 MPa(m) ), and the remaining 5

six data are from specimens 9T5 or 9T6 (loaded to roughly 27 MPa(m) 0* ). The results for 9Tl and 9T2 are contained in Reference [9]; the final data for 9T5 and 9T6 are contained in EPRI MRP-340, but have not been openly published. As discussed later in Section 2.4, the addition of cold work may result in a material that is substantially more susceptible than the as-received material. The extent of transition along the crack front to intergranular cracking for these data was extremely low (::0 10%) for the ten points from specimens tested at constant temperature. A plot of crack growth rate versus K for the Alloy 690 data from MRP-375 for heat WP787 is provided here as Figure 2. As in Figure 1, there is significant growth rate variability within the data for the same heat of material. The median for the CIEMAT specimens is more than a factor of 12 below the MRP-55 curve. Additionally, the Pacific Northwest National Laboratory (PNNL) data indicate that the specific laboratory that produces the data can significantly influence the reported growth rate, such that there is a substantial possibility that a small number of reported data points with relatively high crack growth rates from a single laboratory are not characteristic of the true susceptibility of a specific heat of Alloy 690 material.

The three ANL data points are for CT specimens C690-CR-1 and C690-LR-2, comprised of Valinox heat number WP 142 CRDM nozzle material that were not cold worked and were tested at 21 to 24 MPa(m) 05 , 320°C, and 23 cc Hi kg H20 [10].

The intergranular engagement for these specimens was extremely low (almost entirely trans granular). A plot of crack growth rate versus K for the Alloy 690 data from MRP-375 for heat WP142 is provided here as Figure 3. As in Figure 2, PNNL data indicate that the specific laboratory that produces the data can significantly influence the reported growth rate.

• Figure 3-3 of MRP-3 75. Figure 3-3 shows the complete set of data points compiled for Alloy 690 heat affected zone (HAZ) specimens at the time MRP-375 was completed by the PWSCC Expert Panel that was organized by EPRI. The following points are within a factor of 12.0 below the MRP-55 deterministic crack growth rate for Alloy 600:

4

EC 288005 RO Attachmen\3 PSL-ENG-SESJ-17-002 RO 32totalpages Dominion [n~ineerin~. Inc TN-5696-00-02, Rev. 0 There are eight points within a factor of 12.0 below the MRP-55 75th percentile curve, out of a total of34 points shown in Figure 3-3 ofMRP-375. All but one of the eight data points are for PPU testing, and all but two appear to have had very little to no intergranular engagement.

Six of the points are from ANL testing of specimens comprised ofValinox CRDM nozzle material beat WP!42 and Alloy !52 filler (Special Metals beat WC43E9),

tested at 320°C and 23 cc H2/kg H20 [11]. Five of the points are from specimens CF690-CR-1 and CF690-CR-3 (loaded to roughly 28 to 32 MPa(m)0 *') [I Jl* and the otherpomttS from spec~men CF690-CR-4 (loaded to roughly 22 MPa(m) *) [12). A plot of crack growth rate versus K for all the Alloy 690 HAZ data from MRP-375 for beat WP142 is provided here as Figure 4. As discussed below, PPU conditions-under which five of these six points were obtained-may result in accelerated crack growth relative to plant conditions.

The remaining two points are from CIEMAT testing of specimens 19ARH1 and 19ARH2, comprised of welded Alloy 690 plate material, tested at340'C and 15 cc H2/kg H20 , and loaded to roughly 37 MPa(m)"-' [8]. A plot of crack growth rate versus K fortbe Alloy 690 HAZ data from MRP-375 for plate material tested by CIEMAT is shown in Figure 5. As discussed later, the orders of magnitude difference between these two PPU points and the constant load testing for this HAZ is indicative of the substantial accelerating effect that PPU testing can have beyond what would be expected in service environments.

Figure 3-5 of AfR.P-375. Figure 3-5 shows the complete set of data points compiled by the PWSCC Expert Panel organized by EPRI at the time MRP-375 was completed for Alloy 52 and 152 weld metal specimens. The following points are within a factor of 12.0 below the MR.P- 115 deterministic crack growth rate for Alloy 182:

There are 19 points within a factor of 12.0 below the MRP- 115 75th percentile curve, out of a total of212 points shown in Figure 3-5 ofMRP-375. Five of these points are not relevant to PWR conditions and should not be considered further, as discussed in the following bullets.

One of these points is from PNNL testing of the dilution zone of a dissimilar metal weld between 152M (Special Metals heat WC83F8) and carbon steel, tested at 360°C and 25 cc H2/kg H20 [13]. This material condition is not applicable to the wetted surfaces ofCRDM nozzle J-groove welds because the dilution zone where Alloy 52/ 152 contacts the low-alloy steel RV bead is below the stainless steel cladding. A plot of crack growth rate versus K for the Alloy

!52 data from MRP-375 for beat WC83F8 is provided here as Figure 6.

Four of the remaining points, including the point closest to the MRY- 115 curve, are for environmental fatigue pre-cracking test segments [14]. The status of these four data points, which are shown in black in Figure 7, as being fatigue pre-cracking test segments irrelevant to PWSCC conditions was clarified subsequent to publication ofMRP-375.

The remaining 14 data points represent four specimens from Alloy 152 weld material (Special Metals beat WC04F6) that were tested by ANI.. at 320'C and 23 cc H,lkg H20 ([15] and [10]). Ten of these points are for specimen Al52-TS -5 at loads of about 28, 32, and 48 MPa(m) 0*' [14]. The other four points were obtained at loads of

EC 288005 RO Attachment 3 PSL-ENG-SESJ-17-002 RO 32 total pages Dominion [n~ineerin~. Inc. TN-5696-00-02, Rev. 0 27 MPa(m) 0 *5 for specimen Nl52-TS-l and 30 MPa(m) 0 *5 for specimens Al52-TS-2 and Al52-TS-4. The Alloy 152 specimens all came from welded plate material. A plot of crack growth rate versus K for the Alloy 152 data from MRP-375 for heat WC04F6 is provided here as Figure 7. All but three of these points were for PPU conditions, which may result in accelerated crack growth rates that are not directly representative of plant conditions. Figure 7 shows a very large variability in the crack growth rate reported by different laboratories for this heat of Alloy 152 weld material.

Roughly one third the ANL data (specimen Nl52-TS- l), all of the General Electric Global Research Center (GE-GRC) data, and all the PNNL data for this heat are for specimens from a single weld made by ANL [16], illustrating the role of experimental variability. A small number of elevated data points for a weld produced by a single laboratory may not be representative of the true material susceptibility.

2.2 Data Most D;rectly Applicable to Plant CondWons As described above, Section 3 ofMRP-375 took an inclusive approach to statistical assessment of the compiled data. A conservative approach was applied in which both constant load data and data under PPU conditions were plotted together. In addition, weld data reflecting various levels of weld dilution adjacent to lower chromium materials was included in the data for Alloys 52/ 152. An assessment of the crack growth rate data points most applicable to plant conditions is presented in Figure 8 through Figure 13 . The assessment shows very few points located within a factor of 12.0 below the deterministic MRP-55 and MRP-115lines, with such points only slightly above the line representing a factor of 12.0:

Figure 8 for Alloy 690 with Added Cold Work Less than I 0%.

Only seven of the 55 points are within a factor of 12.0 below the MRP-55 deterministic crack growth rate for Alloy 600.

Figure 9 shows that the data are bounded by an FOI of more than 12 relative to Alloy 600 data on a statistical basis.

• Figure 10 for Alloy 690 HAZ.

Only one of the 24 points is within a factor of 12.0 below the MRP-55 deterministic crack growth rate for Alloy 600.

Figure II shows that the data are bounded by an FOI of more than 12 relative to Alloy 600 data on a statistical basis.

• Figure 12 for Alloys 52/152.

Only three of83 points are within a factor of 12.0 below the MRP-115 deterministic crack growth rate for Alloy 182.

Figure 13 shows that the data are bounded by an FOI of more than 12 relative to Alloy 182 data on a statistical basis.

As discussed above, the technical basis for heads with Alloy 600 nozzles assumes the substantial possibility of crack growth rates substantially greater than that predicted by the deterministic 6

EC 288005 RO Attachment 3 PSL-ENG-SESJ-17-002 RO 32 total pages Dominion fn~ineerin~. Inc TN-5696-00-02, Rev. 0 equations ofMRP-55 and MRP-115 . The MRP-55 and MRP-115 deterministic crack growth rate equations are not bounding equations, but rather reflect the 75'h percentile of material variability. Thus, the perspective provided in Figure 9, Figure 11, and Figure 13 is most relevant to drawing conclusions regarding FOI values applicable to inspection intervals for heads fabricated using Alloy 690, 52, and 152 materials.

The data presented in Figure 8 through Figure 13 were included on the basis of the following considerations:

• As demonstrated and discussed in MRP-115, certain PPU conditions will act to accelerate the crack growth rate. PPU conditions, which include a periodic partial reduction in load, are often used in testing to transition from initial fatigue conditions toward constant load conditions with the crack in a state most representative of stress corrosion cracks if they had initiated in plant components over long periods of time. The periodic load reductions and accompanying load increases may rupture localized crack ligaments along the crack front, facilitating transition of the crack to an intergranular morphology. In MRP-115, data with hold times Jess than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> were screened out of the database for Alloys 82/182/ 132.

The greater resistance of Alloys 690/52/ 152 to cracking is expected to result in a greater sensitivity of the crack growth rate to partial periodic unloading conditions. Figure 14 and Figure 5, in particular, show that there is an apparent significant bias for the data for Alloy 690 in which the data for partial periodic unloading conditions are substantially higher than for constant load conditions. Thus, the data presented in Figure 8 through Figure 13 have been restricted to the constant load (or constant K) conditions that are most relevant to plant conditions for growth of stress corrosion cracks.

• The Alloy 52/ 152 weld metal data shown in Figure 3-5 and Figure 3-6 ofMRP-375 include data reflecting a range of weld dilution levels. The data presented in Figure 12 and Figure 13 exclude the weld dilution data points because of the limited number of data points available, the variability in results, and the limited area of continuous weld dilution for potential flaws to grow through. The weld dilution data are not reflective of the full chromium content of Alloy 52/ 152 weld metal.

The data presented in Figure 12 and Figure 13 exclude a small number of data points that reflect cracking at the fusion line with carbon or low-alloy steel material. Some of these data reflect cracking in the adjacent carbon or low-alloy steel material that was not post-weld heat treated as would be the case in plant applications.

• The data presented in Figure 12 and Figure 13 eliminate the few data points that in fact reflect fatigue pre-cracking rather than stress corrosion cracking. The status of these data points was clarified subsequent to publication ofMRP-375.

The limited number of remaining points in Figure 8 and Figure 12 that lie within a factor of 12.0 below the deterministic MRP-55 and MRP-115 lines represent the upper end of material and/or

-experimental variability. Figure 9, Figure 11, and Figure 13 consider the variability in crack growth rate among different heats/welds of Alloys 600/82/ 182 and compare this against the full variability of the Alloy 690/52/ 152 data most applicable to plant conditions. The lack of any 7

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion [n~ineerin~. Inc. TN-5696-00-02, Rev. 0 points within a factor of 12 when accounting for variability in Alloy 600/82/182 crack growth rates supports a reexamination interval longer than the requested interval corresponding to an FOI of 12.0. The volumetric or surface inspection interval for heads with Alloy 600 nozzles reflects consideration of crack growth rates on a statistical basis, with crack growth rates often higher than that given by the deterministic equations ofMRP-55 and MRP-115.

2 .3 Data Specific to Argonne National Laboratory (ANL) and Pacific Northwest National Laboratory (PNNL)

The U.S. NRC is most familiar with the crack growth data for Alloys 690/52/ 152 that have been generated by ANL and PNNL, so the data specific to these national laboratories have also been evaluated separately. Based on the compilation of ANL and PNNL crack growth rate data recently released by NRC [17] 4 , the results are shown in Figure 15 through Figure 20. These data reflect Alloy 690 test specimens with up to 22% added cold work. The data in Reference

[17] are consistent with the ANL and PNNL data in the wider database presented in MRP-375.

As shown in Figure 15, Figure 17, and Figure 19, only 10 of the total of86 constant load (or constant K) data points generated by ANL and PNNL are within a factor of 12.0 below the deterministic MRP-55 and MRP- 115 lines. Only one of these points is within a factor less than 9.0 below the deterministic MRP-55 and MRP-115 lines. Furthermore, among the constant load data, only five of the 55 points with less than I 0% cold work are within a deterministic factor of 12.0. Finally, when the statistical variability in material susceptibility is considered for the reference material (Alloys 600 and 182) as well as for the subject replacement alloys, all the data points for constant load conditions show a factor of improvement greater than 12.0. This favorable result is clearly illustrated in Figure 16, Figure 18, and Figure 20.

2.4 Da ta for Alloy 690 Wrought Materia/Including Added Cold Work up to 20% for CRDM Nozzle and Bar Material Product Forms An assessment of the crack growth rate data points for Alloy 690 CRDM nozzle and bar material product forms for cold work levels up to 20% is presented in Figure 21 and Figure 22.

Equivalent plots for Alloy 52/152 material for the purpose of including the limited number (i.e.,

five) of weld metal data points generated for added cold work conditions are shown in Figure 23 4

The data in Reference [16) are augmented by the crack growth rate data for Alloys 521 152 produced by PNNL and previously published in an NRC NUREG contractor report (17]. While these PNNL data are shown graphically in Enclosure 3 of Reference [16], the enclosures of tabular data in this NRC document omitted all of the PNNL data for Alloys 52/152. It is also noted that contrary to the enclosure titles ofReference [16], Enclosure 2 contains the PNNL tabular data, and Enclosure 4 contains the ANL tabular data.

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion [n~ineerin~. Inc. TN-5696-00-02, Rev. 0 and Figure 24. Added cold work for weld metals is not directly relevant to plant material conditions.

For Alloy 690 control rod drive mechanism (CRDM) I control element drive mechanism (CEDM) nozzles and other RV head penetration nozzles, the effective cold-work level in the bulk Alloy 690 base metal is expected to be no greater than roughly 10%. This is based on fabrication practices specific to replacement heads, i.e., material processing and subsequent nozzle installation via welding [19]. Furthermore, the crack growth rate data presented for Alloy 600 in MRP-55 do not include cases of added cold work. Comparing cold worked Alloy 690 data against non-cold worked Alloy 600 data results in a conservatism in the factor of improvement for Alloy 690 material as the cold worked material condition for Alloy 600 would be expected to result in a somewhat increased deterministic crack growth rate for Alloy 600, and thus a greater apparent factor of improvement. Nevertheless, the assessment in Figure 21 through Figure 24 is included in this document to illustrate the effect of higher levels of cold work. These data show the potential for modestly higher crack growth rates for such elevated cold work levels for the material product forms most relevant to RV top head nozzles .

2.5 Conclusion The data presented above support factors of improvement greater than 12 for the CGR performance of Alloys 690/52/152. Thus, the available laboratory CGR data support a volumetric inspection interval of at least 20 years for Alloy 690 RV head nozzles.

3 POTENTIAL IMPLICATIONS OF SPECIFIC CATEGORIES OF NOZZLE AND WELD MATERIALS Section 3 assesses the available laboratory CGR data for the potential concern of elevated CGRs for specific categories of nozzle and weld materials.

3.1 Potential Similarities for Laboratory Specimen Material Exhibiting a Deterministic Factor Less than 12.0 Any similarities between (a) the data points within a factor of 12.0 below the MRP-55/MRP-115 curve in Figure 3-1,3-3, and 3-5 ofMRP-375 and (b) the associated nozzles and weld material used in the RV heads in U.S. PWRs are as follows:

9

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion [n~ineerin~. Inc. TN-5696-00-02, Rev. 0 Figure 3-1 of MRP-375 [2]. The only Alloy 690 CRDM material for which crack growth rate data were available at added cold work ofless than 10% (the threshold for inclusion in Figure 3-1 ofMRP-375) was supplied by Valinox Nucleaire. The few data using CRDM material from other suppliers were obtained at cold works of 20% or higher and were not included in the assessment. The data do not indicate any correlation between material supplier and susceptibility to crack growth rate. Fourteen of the Alloy 690 crack growth data points within a factor of 12.0 below the MRP-55 [I] deterministic crack growth rate in Figure 3-1 ofMRP-375 were produced for specimens of Alloy 690 CRDM nozzle material that was supplied by Valinox Nucieaire. However, for the reasons explained below (e.g.,

the variability among data from different laboratories, the variability among data for a single heat and laboratory, and the use ofPPU for eight of these 14 data), this similarity in no way indicates any specific concern for elevated PWSCC susceptibility of the head nozzle material provided by any one supplier.

Figure 3-3 of MRP-375 [2]. Six of the Alloy 690 HAZ data points above a crack growth rate 12.0 times lower than the MRP-55 deterministic crack growth rate in Figure 3-3 of MRP-375 were also produced for specimens of Alloy 690 CRDM nozzle material that was supplied by Valinox Nucleaire. However, for the reasons explained below, this similarity in no way indicates any specific concern for elevated PWSCC susceptibility of head nozzles produced from Valinox material in comparison to Alloy 690 nozzles from another supplier. It is noted that the welding process used to produce the HAZ in the test specimens is not specific to any particular categories of replacement heads.

Figure 3-5 ofMRP-375 [2]. There are no relevant similarities between (a) the Alloy 52 and !52 data points above a crack growth rate 12.0 times lower than the MRP-11 5 [2]

Alloy 182 deterministic crack growth rate in Figure 3-5 ofMRP-375 and (b) the Alloy 52/ 152 weld material used in any particular categories of replacement heads. The variability among test welds with respect to PWSCC crack growth susceptibility reflects a combination of how the weld was made (welding procedure, weld design, degree of constraint, etc.) and perhaps the material variability in the weld consumable (e.g.,

composition). The test welds used to produce the specimens that showed crack growth rates within a factor of 12.0 below the MRP-115 crack growth rate are not identified with any particular fabricator of replacement RV heads. Furthermore, the weld specimens used in the crack growth rate testing were machined from test welds in flat plates, not from actual J-groove welds. Thus, the test weld specimens should not be associated with particular fabrication categories of replacement heads.

3 .2 Potential Implications The material and welding similarities in no way indicate any specific concern for elevated PWSCC susceptibility of the head nozzles at any U.S. PWR or provided by any supplier in comparison to other heads with Alloy 690 nozzles or Alloy 690 nozzles supplied by any other supplier. It is emphasized that a small number of data points showing relatively high crack growth rates cannot readily be concluded to be characteristic of the true material behavior expected in the field. This conclusion is made considering the following:

10

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0

• The only heats of Alloy 690 CRDM nozzle material that have been used in crack growth rate testing with less than 10% added cold work are supplied by Valinox. Consequently, there is no basis to suggest material from any one supplier is more susceptible than that from another based on the presence or absence of data points within a given factor ofthe deterministic crack growth rate curve from MRP-55.

• The data points showing the highest crack growth rates for the tested Valinox material reflect partial periodic unloading conditions. As discussed above, such conditions tend to result in accelerated crack growth rates that are not representative of plant conditions.

• Most of the crack growth rate data for heats that had points within a factor of 12.0 below the MRP-55 deterministic curve or MRP-115 deterministic curve were substantially lower.

The best-estimate behavior for every heat or test weld of material presented in Figures 3-2, 3-4, and 3-6 ofMRP-375 reflects a factor of improvement of 12 or greater. In addition, other factors being equal, one would expect a greater range of crack growth rates for a material heat for which a greater number of data points was produced. Some of the scatter likely reflects experimental uncertainty as opposed to true material variability.

Experimental uncertainty is more of a factor for the data for Alloys 6901521152 than for Alloys 60018211821132 considering the greater testing challenges associated with the more resistant replacement alloys.

• In some cases, different laboratories have reported large differences in crack growth rate for the same material heat or test weld. This behavior is illustrated in Figure 7 for the Alloy 152 heat WC04F6 and Figure 3 for the Alloy 690 heat WP142. Thus, individual data points showing relatively high crack growth rates might not reflect the true susceptibility of particular categories of nozzle or weld material. Consistent data from multiple laboratories may be needed before one can conclude that a particular category of nozzle or weld material has an elevated susceptibility to PWSCC growth.

• Some type ofPWSCC initiation is necessary to produce a flaw that may grow via PWSCC.

Laboratory and plant experience show that Alloys 6901521152 are substantially more resistant to PWSCC initiation than Alloys 6001821182 [2]. PWSCC has not been shown to be an active degradation mode for Alloys 6901521152 components after use in PWR environments for over 25 years.

• The crack growth rate data compiled in MRP-375 [2] for Alloys 52 and 152 reflect the composition variants applicable to PWR plant applications. Data are included for the following variants: Alloy 52 (UNS N06052 I AWS ERNiCrFe-7), Alloy 52M (UNS N06054 I A WS ERNiCrFe-7 A), Alloy 52MSS (UNS N06055 I A WS ERNiCrFe-13), Alloy 52i (AWS ERNiCrFe-15), Alloy 152 (UNS W86152 I AWS ENiCrFe-7), and Alloy 152M (UNS W86152 I A WS ENiCrFe-7). Considering the overall set of available crack growth rate data for the various variants of Alloy 52 and 152, there is no basis for concluding at this time any significant difference in the average behavior between the Alloy 52 and Alloy 152 variants in use at U.S. PWR RV heads with Alloy 690 nozzles.

In addition, it should be recognized that PWSCC of Alloy 690 R V head penetration nozzles or their Alloy 521152 attachment welds is not an active degradation mode. Thus, it is premature to single out individual materials or fabrication categories of heads with Alloy 690 nozzles for additional scrutiny on the basis of subsets of laboratory crack growth rate data. In the case of 11

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 heads with Alloy 600 nozzles, for which PWSCC is an active degradation mode, materials and fabrication categories of heads with relatively high incidence of PWSCC are inspected in accordance with the same requirements as other heads.

Based on the additional information and discussion provided above, it is concluded that the available crack growth rate data do not indicate any susceptibility concerns specific to the nozzle or weld materials specific to any given replacement head or category of replacement heads.

4 REFERENCES

1. Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primmy Water Stress Con*osion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55)

Revision I, EPRI, Palo Alto, CA: 2002. 1006695. [freely available at www.epri.com]

2. Materials Reliability Program Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) ofAlloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004. 1006696. [freely available at www.epri.com]

3. ASME Code Case N-729-1, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," Approved March 28, 2006.

4. Materials Reliability Program: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-3 75), EPRI, Palo Alto, CA: 2014. 3002002441. [freely available at www.epri.com]

5. Materials Reliability Program: Inspection Plan for Reactor Vessel Closure Head Penetrations in US. PWR Plants (MRP-117), EPRI, Palo Alto, CA: 2004. 1007830. [freely available at www.epri.com; NRC ADAMS Accession No. ML043570129]

6. Materials Reliability Program: Reactor Vessel Closure Head Penetration Safety Assessment for US. PWR Plants (MRP-IIONP), EPRI, Palo Alto, CA: 2004. 1009807-NP.

[ML041680506]

7. Materials Reliability Program: Probabilistic Fracture Mechanics Analysis of PWR Reactor Pressure Vessel Top Head Nozzle Cracking (MRP-105 NP), EPRI, Palo Alto, CA:

2004. 1007834. [ML041680489]

8. D. G6mez-Bricefio, J. Lapefia, M. S. Garcia, L. Castro, F. Perosanz, and K. Ahluwalia, "Crack Growth Rate of Alloy 690 I 152 HAZ," Presented at: Alloy 690/152/52 Research Collaboration Meeting, Tampa, FL, December 1-2, 2010.

9. D. G6mez-Bricefio, J. Lapefia, M.S. Garcia, L. Castro, F. Perosanz, L. Francia, and K.

Ahluwalia, "Update of the EPRI-UNESA-CIEMAT Project CGR Testing of Alloy 690,"

12

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~t Inc. TN-5696-00-02, Rev. 0 Presented at: Alloy 690/152/52 Research Collaboration Meeting, Tampa, FL, November 29-December 3, 2011.

10. Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld in Simulated PWR Environment- 2009, NUREG/CR-713 7, June 2012.

11. B. Alexandreanu, Y. Chen, K. Natesan and B. Shack, "Cyclic and SCC Behavior of Alloy 690 HAZ in a PWR Environment," 15th International Conference on Environmental Degradation, pp. 109-125, 2011.

12. B. Alexandreanu, Y. Chen, K. Natesan and B. Shack, "Update on SCC CGR Tests on Alloys 690/52/152 at ANL- June 2011," Presented at: US NRC/EPRI Meeting, June 6-7, 2011. [ML111661946]

13. M. Toloczko, M. Olszta, N. Overman, and S. Bruemmer, "Stress Corrosion Crack Growth Response For Alloy 152/52 Dissimilar Metal Welds In PWR Primary Water," 16th International Conference on Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, Paper No. 3546, 2013.

14. B. Alexandreanu, Y. Chen, K. Natesan and B. Shack, "SCC Behavior of Alloy 152 Weld in a PWR Environment," 15th International Conference on Environmental Degradation, pp.

179-196, 2011.

15. B. Alexandreanu, Y. Chen, K. Natesan and B. Shack, "Cyclic and SCC Behavior of Alloy 152 Weld in a PWR Environment," Presented at: Alloy 690/152/52 Research Collaboration Meeting, Tampa, FL, November 29-December 3, 2011.

16. M. Toloczko, M. Olszta, N. Overman, and S. Bruemmer, "Observations and Implications oflntergranular Stress Corrosion Crack Growth of Alloy 152 Weld Metals in Simulated PWR Primary Water," 16th International Conference on Environmental Degradation of Materials in Nuclear Power Systems- Water Reactors, Paper No. 3543, 2013.

17. Memo from M. Srinivasan (U.S. NRC-RES) to D. W. Alley (U.S. NRC-NRR),

"Transmittal ofPreliminary Primary Water Stress Corrosion Cracking Data for Alloys 690, 52, and 152," October 30, 2014. [ML14322A587]

18. Pacific Northwest National Laboratory Investigation of Stress Corrosion Cracking in Nickel-Base Alloys, NUREG/CR-7103, Vol. 2, April2012.

19. Materials Reliability Program: Material Production and Component Fabrication and Installation Practices for Alloy 690 Replacement Components in Pressurized Water Reactor Plants (MRP-245), EPRI, Palo Alto, CA: 2008. 1016608.

13

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 Data from Individual Heats 1.E-09 1.E-1 0

• ~ MRP-551 Curve/1 f

~

5

-C1l

<':1 0:::

.t::

1.E-11 .v mr.w

/

/ - ..*

ts.A

- -- -- ~-

0

(.!)

~ '7""'* ~----------*-*---- PPU data are u represented with

<':1 7'- AI:- -

• u 1.E-12 / open symbols Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa,,m)

Figure 1. Plot of Crack Growth Rate (da/dt) versus Stress Intensity Factor (K1) for Alloy 690 Data from Plate Material Tested by CIEMAT 1.E-09 TI~

u"': ~~:g T ~~~~~~~~~~~~~~~~~~ . J MRP-55 5~

1.E-1 0 t:;;~~;;~~;;;;;;~~~~~~~~~==;;~c~u~

-.::::..~::::::::: ...:.....:::::.....:::::::...:::::::::::::::::::..:::::_____::::::::::::::::::::::::::...::.._::__::::::::::::::::::::

rv~e/~1b:

I ~;J

• • • • • • • • • • • • • ~*-***: .* *. . . . * . .* . .* . *.*==~iA * * . if>"&_Ai:_** = * :=,::;;::****;;;;;;;;;*** *~-

~ 1.E-11 ................................... .....................

0

(.!) - A PPU data are t5 I **************** .,,.

- .... '.................................................................................................................................. ,......L7.**A**A ********************************************************** **--******* I represented with

~ 1.E-12 ~~~

~

/ ~,~~~~~~~~~~~ ~~~*t~~~~~~~op~e~n~sy;m;b~o~~ Data are adjusted for

~----===~-----* -- -==--*-_*----*==========*-...------ temperature (325°C).

Q = 130 kJ/mol 1.E-13 -t-.-...___._'--+-'__.__.-'--+..........._............_+--'-._.__.--t-'-...............""'--+............_............_,r-'--'-__.__._~:::;::::;:::.:::::t=.:::::::;:::::;=:::;::::;:::::;::::;:~

10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa,,m)

Figure 2. Plot of da/dt versus K1 for Alloy _690 Data from Heat WP787 14

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 1.E-09l~;J~~~~~~~~~~~~~

l ~~~~~~~~! MRP-55 1~~~~~~~~~~~~iiii~~iiiiiiii~~C~UNiei/~~~~ii 1

1.E-1 0

§'

5

1£~1 1-~=~=~ ~=~=-~ =~ ~=~-~=~~.~~-~-~-~-~-~-~-~ ~=~=~=~=~=~=~~=~=-~=~=~~=~=~=~~=~=~=~~=~=~~=~=~=~~

~=~:*~~ u~-~ ~ ~  :~:~:~-~~

e 1 ~~~~~~~~~~~~~~~~~~~~~~~:::."""~~~~~~- **** ********************o"'"************************************ ************************************************ * * * ** ****** *** ** ***** ** * * * * * * * * * * * *** * * * * * ** * * * * * ***********

~ ~~~~~~ ~~~ ~* * * * * * *** *'<:; I represented with

~ ~~~~,~~~~~~~~~~~~~~~~~~~~o~pe;n~s;ym;b~o;ls~

1.E-12 [: Data are adjusted for l ---- * * * * * * * * * * * * * * ---- - *** **** * * * * * ** * * * * * * * * * * * * * * * * * * ****** ------- temperature (325°C).

• • • • • • • • • • • • • • • • • • • • • • • • • • • -*-*-*********-**-**-*- **************************************-******<)***** ****************-***-*****-****- ************************************--

Q = 130 kJ/mol 1.E-13 +--..---'--'--'-t---'--'-....1.-L-t--L....I.-L-'--f--'--'--'--'-<A~__._.-+-'__._..........,f-'-'--'-'-t-Cc.;:::::;::j::::i::;:::::;::;::::;:::;::::::;:::;::::;::j 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-.im)

Figure 3. Plot of da/dt versus Kt for Alloy 690 Data from Heat WP142

1. E-09lfE ~l:~~~~~~~~~~J MRP-55

~*---~~---**-~***- * -~*- * -~*- * *; ~-c~-~-~-~=~~~ ;;~

1

'Vl 1.E-1 0

~* -~* *-~*- * *;*** *-~*.* .*;.*. *-***-**~;* * * * *;* * * * * * * * *~*-* *****~*-****i**** -~**-***-*~*-*****-~**-**-*~**-***-~*- *-----

t~- -~*- * -~*- * -~*- * -~*- *--~ -~ -* *-~*- ** *-~**-*** *-~**-*** *-~**-* ** *-*~* *-~-** *** -*~ . .;. . . . . . ...;. . . . ......;.......;....

1 i~

0 1.E-11

=- F----aa 0 t5 - '"' PPU data are

~ - - -/- "":. . . . . . . . . . ......................... --~*-****** . - - ------ . . .. represented with

-~---~--~--~--;.;.:;~D~a:ta~a~;:~:~~:J~.:~sy~~e:~~~:r~

1

~ 1 .E- 12 lr-¥- ,. ~~~~--~----~--~----~-~--~--~---~----~-~--~--~- -~- ~- -~- -~- -~- -~- ~- -~- ~- -~- ~- ----~---~---~---~-- ~

~-- :=:.==:*=:=::::=*=::::=.==:=~--==::.-:=.:.:.=.:.:.:.:.:.*:.:.::.:.:.:.:.::.:.:.::.:.::-::.::.::::::*:-::::=.::=.:.-:.: temperature (325° C).

• * ************ *** - * * * * * * * * * * * * *** * **** ----- Q = 130 kJ/mol 1.E-13 .J-.-~'--~-'__.__..........,__,_,__.__._+-'-L....L-l--t-'--'-'--'--t--'--'-............_,r-'--'_._,__p:::;::::::;:::L::f=;:::::::;::::;::::l:::f::::;::::::;:~

10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-.im)

Figure 4. Plot of da/dt versus Kt for Alloy 690 HAZ Data from Heat WP142 15

EC 285797 RO Attachment 3 PSl-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 1.E-13 -J-.....J.......L_.l......j---'--'--'-+...J.....L..""--'--~-'-J........I....J'-+-'--'--'-~""--'-.J......L.._t-'-'--'--"..,_.:::::;:::::;:::::L:::t=:;::::::::L::L::::J:::i::::;:::::;:~

10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa"-'ml Figure 5. Plot of da/dt versus K1 for Alloy 690 HAZ Data from Plate Material Tested by CIEMAT

1. E-13 +-'-....L..-L.....L.-f--'-'........_'-+-J.......L._J........L._+-'-.J......L..~....L....L.....L..-L.-+-'--'---'-I--'--J.......L._-'-+-.J......L....J......I..+-'-....L..-L.....L.-f......._._..l,.,ooj-....j 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa"-'ml Figure 6. Plot of da/dt versus K1 for Alloy 152 Data from Heat WC83F8 16

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~,lnc. TN-5696-00-02, Rev. 0 1.E-09 [::ro~A~N~L~1--~-~**-~---~---~-~---~---~--~-- ~--~---~-~----~-~---~-~----~~-~--~---~-~--~--§--*~---~----~-~---~- ~-~~*** * ~* *~~*--~--~--~---~-~-

DGE-GRC U-10 ~ PNNL

~

.s

~ ~: represented with l- - - - - - - - - -1 u 1.E- 12 +F=o="p=e=n:::!s==m=b=ol=s:::::=!=o-=

Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 +-'--'--"-'-+--'---'----'---+-'--'-J........L.._+-'--'-'-~-'--"-'--'--+-'-'-L..::iL.+--'--'--'--'-+....___.__.___._+-'-_.__.--'--~........_.._.__~

10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa.Ym)

Figure 7. Plot of da/dt versus K1 for Alloy 152 Data from Heat WC04F6 17

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 Data Most Applicable to Plant Conditions 1.E-09 1:~o::A~N~L~-~~~~~--~** ~* *-~* * * *~* * * *~* * * ~* * * *~* * * *~* * * ~** * *~* *** ~**** *~* * ~** * *~* * * *~-** *~** ~~~~~--~----~-~~

.§.

I= oBettis

~ 1.E-10 ::: ~ PNNL

..t. CIEMAT

~=* *::**:-*:*-:**:**: -:*:-*:**-~*-*;-.;-*--* ~* *-*-: ~:.: :-~~~~=:l~~~~~

=: : : : : : : : : : : : : : : : :. .:::::. ~

~- . . -

~~~~;~r:**-:****:- :J. . . .:. .

Q) ~

-;;; /./ - - - -- * - -- ...P!I.*ti . --fMRP/121-0::: / &A A.- --- .-- - - ~

~ 1E~1 ~~~~~~-~~~~~~~~~~~~~~~~~~~~~~~

u ,. ~

~ 1.E~2 1~~~~~~~~~-~~~~-~-~~~~~b~~~~-~ ~~~~~~~~~~~-~-~-~--~ - ~-r*~~~~~~~ Q ..,

Data are adjusted for temperature (325°C).

r-----------~~r----&------1 Q = 130 kJ/mol 1.E-13 +~~~~~~~~~~~~~:~~~~~~~~~

~ ~ ~ """ ~~~~~~~

10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-.lm)

Figure 8. Plot of da/dt versus K1 for Alloy 690 Data from All Laboratories, :S 10% Cold Work, Constant Load or K1 1.0 oANL fj,ff ,..... - /

&~ -f*

0.9 oBettis 0.8 A CIEMAT ----# 7- - - - -

.§ 0.7

• PN NL - ----d!s ,f 5

=5 o.6 "'~

/),&.

,t I 8~

I VI

~ 0.5 I I The data points at 1E-13

.::: I <if I

I I MRP-55 I were reported as "no

~ 0.4 (FOI = 1) growth."

E rs I I 8 0.3 Data are adjusted for 0.2 FOI- 12

/

I temperature (325°C) and stress intensity factor.

0.1 7

/ / Q = 130 kJ/mol

_,., ~ K= 30 MPa..Jm 0.0 1.E-13 1.E-12 1.E-11 1.E-1 0 1.E-09 1.E-08 Crack Growth Rate (m/s)

Figure 9. Cumulative Distribution Function of Adjusted da/dt for Alloy 690 Data from All Laboratories, :S 10% Cold Work, Constant Load or K1 18

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16~002 RO 32 total pages Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0

~ A CIEMAT *---._*--~----~-----*--~-------~-- ...-... j MRP-551*--------

D GE-GRC **- - -** ,,. I Curve/1 r - '

~ PNNL ~~ ~~~~:~~~'  : .:. .. . . ::::. .*.* . .:* .* . ::.:*.*::*.*:::: . *::: * ************** - . . . ........... ........ -

• • • • • • • • • • • • • • • *****************~*******

- /-,_.!'--./" _ _ _ _ _ _ _ _ _ ___ __ _ __ __ _ ~_

35 40 45 50 55 60 Stress Intensity Factor (MPa>Jm)

Figure 10. Plot of da/dt versus K, for Alloy 690 HAZ Data from All Laboratories,~ 10% Cold Work, Constant Load or K1 1.0 ,......-

0.9 121

©

(')

L / 0 ANL 0.8

/2,.

/ / 8. CIEMAT CJGE-GRC -

I

.§ 0.7 u D

D I

I ~ PNNL -

'5

§ 0.6 Lb.

I I I Vi

~ 0.5 ~

~

D l- I The data points at 1E-13

• ~ 04

~

I I MRP-55 I were reported as "no

'3 . (FOI = 1) growth."

~

8 E

0.3 ~ I I Data are adjusted for 0.2

~

~

FOI

/

12 I temperature (325°C) and stress intensity factor.

0.1

[] / / Q = 130 kJ/mol 0.0 .-

/ /

~ K = 30 MPa>Jm 1.E-13 1.E-12 1.E-11 1.E-1 0 1.E-09 1.E-08 Crack Growth Rate (m/s)

Figure 11. Cumulative Distribution Function of Adjusted da/dt for Alloy 690 HAZ Data from All Laboratories, ~ 10% Cold Work, Constant Load or K1 19

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 Dl - - - - --1

• - Data are adjusted for

- temperature (325°C).

Q = 130 kJ/mol 1.E-13 -f---L--'--L--'--t--'----'-'-+-"'---'-"'---'-+-'-....___._~~~JO-'--'--<>=-t--'-..___._-'-+....___._-'--'--+-'---'--L--'--t--'----'-'--l 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-.lm)

Figure 12. Plot of da/dt versus Kr for Alloy 52/152 Data from All Laboratories, ~ 10% Cold Work, Constant Load or K1 1.0 0ANL &(!) /li':j" 0.9 A CIEMAT "' 0

/

~ 11.

~::,11. D.

I I

.!2 0.8 LJGE-GRC

• PNNL

~

.A 1 IMRP-115 IFOI-1)

L 0.7 I "5

~ I

@ 0.6 Vi

~ 0.5 0, I I

I I

7 The data points at 1E-13

~

f 1-were reported as "no 8

E 0.4 0.3 s I growth."

Data are adjusted for 0.2 Jf

{!

[FOI=12[ temperature (325°C) and stress intensity factor.

0.1 I Q = 130 kJ/mol

_,./ K = 30 MPa-Jm 0.0 1.E-13 1.E-12 1.E-11 1.E-1 0 1.E-09 Crack Growth Rate (m/s)

Figure 13. Cumulative Distribution Function of Adjusted da/dt for Alloy 52/152 Data from All Laboratories, ~ 10% Cold Work, Constant Load or K1 20

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 Comparison of Partial Period Unloading (PPU) Conditions vs. Constant Load Conditions 10 100 1000 10000 Hold Time (Hours)

Figure 14. Plot of da/dt versus Loading Hold Time (for PPU testing) or Test Segment Duration (for Constant Kdload Testing) from Heat WP787 21

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Domin ion [n~inmin~. Inc TN-5696-00-02, Rev. 0 Figure 15. Plot of da/dt versus KJ for Alloy 690 Data Produced by ANL and PNNL and Available in Reference [17]; S 22% Cold Work 1.0 .,-------------,...----::----~----=-----,

0.9 +----------r<>'--~'----------.f'----------1 0.8 +--------*~-----....:-----/-------------i Data are adjusted for temperature (325°C) and 0.2 ll-------=~=='------/--------1 stress intensity factor.

0.1 !1-- -----..,._-- -- -/-- - -- - ----i Q = 130 kJ/mol K= 30 MPa..Jm 0.0 .__.....:::_...,..__....,~"'-'-'+-~~...._....,-_...___l,=~~~~=:dl 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 Crack Growth Rate (m/s)

Figure 16. Cumulative Distribution Function of Adjusted da/dt Alloy 690 Data Produced by ANL and PNNL in References [17]; S 22% Cold Work and Constant Load/ K1 22

EC 285797 RO Attachment3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion [n~ineerin~. Inc. TN-5691.>-00-02, Rev. 0 1.E-13 .j_....~..._.....~-'-+~._.__.~~._.__~..._.~..__._~....._....'::;::=t=<====t="=:;:'J 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa ~m)

Figure 17. Plot of da/dt versus!(] for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17); :5 22% Cold Work 1.0 ,.------------:::-=-~-- ./""-:::;;oo------,

0 /- ./

0.9 +---~-----n- u / ~-------- /L---------1

! : : :~-** *-~ -~ - * * *:=~* * * * * * *~ * * * ~*-* ~*-* *-~ *,-~~-~ *~* * *

• _** * * *j* ~*-*-**-****-**-**---

.*:_!!!' 0.5

• I

• • _***** _ *f

• • _-__*_/_ - - ~ ~;~~~L 1

Thedatapointsat1E-13are r

- 1~ fRP-55 treated as "no growth,"

.l§

~

04

+-- - - - - - --t-- I

- --1 FOI = 1 consistent with MRP-375.

8 0.3 j Data are adjusted for

~ temperature (325°C) and

~ *-;;;;/(_____ -;;;;/ stre~=~1~~~~tor.

0.0 -J--cc....::::.............--:o:::::::~r--~~~-~-"===<F===~

1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-OB Crack Growth Rate (m/s)

Figure 18. Cumulative Distribution Function of Adjusted da/dt Alloy 690 HAZ Data Produced by ANL and PNNL [17]; :5 22% Cold Work and Constant Load/Kr 23

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion [n~ineerin~. Inc TN-569&-00-02, Rev. 0

• • OANL CL OANLPPU 'I** * . ...............*....

1.E-12 + PNNL CK (NUREG) *

• Data are adjusted for temperature (325*C).

Q = 130 kJ/mol 1.E-13 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa'-'ml Figure 19. Plot of da/dt versus Kr for Alloy 52/152 Data Produced by ANL and PNNL and Available in References [17] and [18]; => 22% Cold Work 1.0 TF=============,-----------------------0~--~~----------,

09 f l o ~a I

• 7 /

• H* PNNL CK (NUREG) 1 - - - - - - - - - - : . 0 ~--/-T------~ 1"- .....--l

§ ::

• s,g

' ~~~~in 7

0.6 + ***************************c*******************************************-************************-******-*************** /--****-***********************************-*-**** *1

~  ; I c_;_ 0.5 +-------.+----------------------F-------1 The data points at 1E-13

- 1 were reported as 'no

] 0:3 041==:t*:===========t_===J~~~gr~ow~lh~.~

• I Data are adjusted for

• ~~~

temperature (325*C) and 0.2 f--'**- - - - - - - - - - - - - - -l!:'FO,!!I_:--...r.12g__ _--l stress intensity factor.

0.1 ........... . --- . . ------ *****-7 -- -

I Q =130 kJ/mol K= 30MPa'-'m 0.0 l---~-~~t----=-:::::..~ --~--~=::i:i=======il ......

1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growtl1 Rate (m/s)

Figure 20. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data Produced by ANL and PNNL ([17] and [18]); => 22% Cold Work and Constant Load/Kr 24

EC 285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion [n~ineerin~, Inc. TN-5696-00-02, Rev. 0 Data for Less than 20% Cold Work from All Laboratories vAMEC =::._ *

.&. CIEMAT Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 -J-.-..J.......I._-'-+-.I.....L.-J......J....f---'-'--'--'-+-'-....I.....I.....J.....t--'--'--L-J_+--'-.I.....L.-'--+-'L......I.....I-'-+~::::;::::;:=t=L::;::::c::i::i=:;:::::;::~

10 15 20 25 30 ' 35 40 45 50 55 60 Stress Intensity Factor (MPa"'l/m)

Figure 21. Plot of da/dt versus K1 for Alloy 690 Data from All Laboratories, > 10 & ~ 20% Cold Work, CRDM and Bar Material, Constant Load or Ki Testing 1.0 /::,. ,......-

0.9 vAMEC

&.& L ~

0.8 L'. CIEMAT EJGE-GRC --**

o /S/ L_

§ 0.7

g

I 0.6

+ PNNL Do, It T I I

Vi rJV /-/ I

~ 0.5 The data points at 1E-13

.::: &.o I I MRP-551 were reported as "no

~ 0.4 A

lFOI = 1) growth."

~~

8 E

0.3 ~D.

I I Data are adjusted for 0.2 *

~ FOI

/

12 I temperature (325°C) and stress intensity factor.

0.1 A /::,. ,

/ Q = 130 kJ/mol 0.0

• - !// ~ K = 30 MPa"'l/m 1.E-13 1.E-12 1.E-11 1.E-1 0 1.E-09 1.E-08 Crack Growth Rate (m/s)

Figure 22. Cumulative Distribution Function of Adjusted da/dt Alloy 690 Data from All Labs, ~

20% Cold Work, CRDM and Bar Material, Constant Load or K1 25

Ec-285797 RO Attachment 3 PSL-ENG-SESJ-16-002 RO 32 total pages Dominion [n~ineerin~, Inc. TN-5696-00-02, Rev. 0 1.E-13 +-'-"""---'--'--t-'---L--'--'--f---'--'----'-+-'-...............-'-+"""---'--'---L-+-'--'--'--'-f-L-'----'--'-t--'-'-...............-t-'--'---L--'-t--'--'--'-Y 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa~m)

Figure 23. Plot of da/dt versus KJ for Alloy 52/152 Data from All Laboratories, > 10 & ~ 20% Cold Work, Constant Load or K1 1.0 (J ANL #Q ..--

0.9 .2> CIEMAT

~

0 / /

D GE-GRC ~ £]

t:;,&.  !), '-"

I 7 0.8

• PNNL 7' 1 IMRP-115[_

§ 0.7

I

§ 0.6 Vi {

.r I I

I I

7 fFOI 1)

I

~ 0.5

.::: ;v ,I I

I The data points at 1E-13 were reported as "no

~ 0.4 growth."

8 E

0.3 ~ I Data are adjusted for 0.2 J' IFOI=12I temperature (325°C) and stress intensity factor.

tf I Q = 130 kJ/mol 0.1 0.0 i _,./ K = 30 MPa--Jm 1.E-13 1.E-12 1.E-11 1.E-1 0 1.E-09 Crack Growth Rate (m/s)

Figure 24. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data from All Laboratories, ~ 20% Cold Work, Constant Load or K1 26