ML17101A678

Relief Request 55 - Request for Relief from the Frequency Requirements of ASME Code Case N-729-1

ML17101A678
Person / Time
Site:	Palo Verde
Issue date:	04/07/2017
From:	Lacal M Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-07464-MLL/TNW
Download: ML17101A678 (48)

Text

Qaps MARIA L. LACAL Senior Vice President, Nuclear Regulatory & Oversight Palo Verde 102-07464-MLL7TNW Nuclear Generating Station April 7, 2017 P.O. Box 52034 Phoenix, AZ 85072 Mail station 7605 Tel 623.393.6491 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3 Docket Nos. STN 50-528/529/530 Reiief Request 55 - Request for Reiief from the Frequency Requirements of ASME Code Case N-729-1 Pursuant to 10 CFR 50.55a(z)(l), Arizona Public Service Company (APS) requests relief from the frequency requirements of American Society of Mechanical Engineers (ASME) Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D). Table 1, item B4.40 of ASME Code Case N-729-1 requires that a volumetric and^r surface examination be performed within one inspection interval (nominally 10 calendar years) of its inservice date for a replacement reactor vessel closure head (RVCH).

Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) Report MRP-375 was developed to support a technically based volumetric and/or surface re-examination interval using appropriate analytical tools. The technical basis in MRP-375 demonstrates that the re-examination interval can be extended while maintaining an acceptable level of quality and safety.

The enclosure contains Relief Request 55 for the proposed extended inservice inspection (ISI) frequency of 15 years. APS requests approval of this relief request by December 31, 2017, for planning purposes in advance of the PVNGS Unit 2 refueling outage scheduled to start in the Fall of 2018, October 6, 2018.

No new commitments are being made in this submittal.

If you have any questions about this request, please contact Michael D. DiLorenzo, Licensing Section Leader, at (623) 393-3495.

Sincerely, Lacal, Maria DN; cn=Lacal, Maria UZ06149)

UZ06149) Date: 2017.04.0716:10:52

-07W MLL/TNW/MSC/sma

102-07464-MLL/TNW ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Request for Relief from the Frequency Requirements of ASME Code Case N-729-1 Page 2

Enclosure:

Relief Request 55 - Proposed Alternative in Accordance with 10 CFR 50.55a(z)(l) cc: K. M. Kennedy NRC Region IV Regional Administrator S. P. Lingam NRC NRR Project Manager for PVNGS M. M. Watford NRC NRR Project Manager C. A. Peabody NRC Senior Resident Inspector for PVNGS

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10CFR 50.55a(z)(1)

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

1. ASME CODE COMPONENT(S) AFFECTED Component; Reactor Vessel Closure Head(RVCH) Nozzles Code Class: Class 1 Examination ASME Code Case N-729-1, Alternative Examination Category; Requirements for PWR Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1 Code Item B4.40 Number;

Description:

Control Element Drive Mechanism (CEDM) Nozzles RVCH Vent Nozzle Size: 4.050 Inch (Nominal Outside Diameter) 1.050 Inch (Nominal Outside Diameter)

Material: RVCH -SA-508 Grade 3 Class 1 Nozzles - SB-166 N06690 (Alloy 690)

Buttering and Weld Material - ERNiCrFe-7 / ERNiCrFe-7A /

ENiCrFe-7 (Alloy 52/152)

There are 97 CEDM nozzles and 1 vent nozzle welded to the inside surface of the RVCH with partial penetration J-groove welds.

2. APPLICABLE CODE EDITION AND ADDENDA The fourth inservice inspection (ISI) interval code of record for Palo Verde Nuclear Generating Station (PVNGS) will be the 2007 Edition with 2008 Addenda of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Examinations of the RVCH penetrations are performed in accordance with the Code of Federal Regulations (CFR) 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of ASME Code Case N-729-1 (Reference 1), with conditions.

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

3. APPLICABLE CODE REQUIREMENT 10 CFR 50.55a(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-729-1 subject 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, must implement their augmented inservice inspection program by December 31,2008.

10 CFR 50.55a(g)(6)(ii)(D)(3) conditions ASME Code Case N-729-1 (Reference 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 must 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 must be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically.

ASME Code Case N-729-1 (Reference 1), section 2410 specifies that RVCH penetrations (nozzles and partial-penetration welds) shall be examined on a frequency in accordance with Table 1 of this code case. The basic inspection requirements of ASME Code Case N-729-1 (Reference 1), as amended by 10 CFR 50.55a, for partial-penetration welded Alloy 690 RVCH penetration nozzles are as follows:

Volumetric or surface examination of all nozzles every 10-year inspection interval (nominally 10 calendar years) provided that flaws attributed to primary water stress corrosion cracking (PWSCC) have not been identified, and Direct visual examination (VE) of the outer surface of the RVCH for evidence of leakage every third refueling outage or 5 calendar years, whichever is less.

4. REASON FOR REQUEST ASME Code Case N-729-1 (Reference 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

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1) after the RVCH 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, per ASME Section XI, Code Case N-729 (Reference 2). Using plant and laboratory data that has since become available. Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) Report MRP-375 (Reference 3) 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 at least a 20-year interval while maintaining an acceptable level of quality and safety. Similar to the precedence summarized in section 7 of this enclosure, Arizona Public Service Company (APS) is requesting approval of this alternative to allow the use of an extended ISI interval of 15 years for the PVNGS Units 1, 2, and 3 Alloy 690/52/152 RVCH penetrations.

5. PROPOSED ALTERNATIVE AND BASIS FOR USE APS is requesting relief from the examination frequency requirements of ASME Code Case N-729-1 (Reference 1), item B4.40 for performing volumetric and/or surface exams of the PVNGS Units 1, 2, and 3 RVCH penetrations. Specifically, this would allow volumetric or surface examinations to be extended as listed in the table below:

Startup from RVCH Current Exam Outage Exam Outage of Unit Replacement per Code Case N 729-1 Proposed Alternative 1 May 2010 U1R21 (currently U1R25 (currently (from U1R15) scheduled scheduled for spring 2019) for spring 2025) 2 December 2009 U2R21 (currently U2R25 (currently (from U2R15) scheduled scheduled for fall 2018) for fall 2024) 3 November 2010 U3R21 (currently U3R25 (currently (from U3R15) scheduled scheduled for fall 2019) for fall 2025)

The examinations are currently scheduled to occur approximately nine calendar years after RVCH replacement, so the alternative corresponds to a nominal extension of five calendar years beyond the requirements of ASME Code Case N-729-1 (Reference 1), item B4.40. This request applies to the item B4.40 inspection frequencies only.

As discussed in the original ASME technical basis document (Reference 2), the inspection frequency of ASME Code Case N-729-1 (Reference 1) for RVCHs with 3

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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 (Reference 4),

which was summarized in the safety assessment for RVCHs in MRP-110NP (Reference 5). The material improvement factor for 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 (nominal 10 calendar years) was established in 2004 as a conservative approach and was intended to be subject to reassessment upon the availability of additional laboratory data and plant experience on the performance of Alloy 690 and Alloy 52/152 (Reference 2).

Further evaluations were performed and documented in MRP-375 (Reference 3) to demonstrate the acceptability of extending the inspection intervals for ASME Code Case N-729-1 (Reference 1), item B4.40 components. In summary, extending the intervals from once each interval (nominally 10 calendar years) to once every 15 calendar years is based on plant service experience, factor of improvement (FOI) studies using laboratory data, deterministic study results, and probabilistic study results.

Per MRP-375 (Reference 3), much of the laboratory data indicated an FOI of 100 for Alloy 690/52/152 versus Alloy 600/182/82 (for equivalent temperature and stress conditions) in terms of crack growth rates (CGRs). In addition, laboratory and plant data demonstrate an 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 Alloy 690/52/152 metals, the proposed inspection interval is based on conservatively smaller factors of improvement.

Deterministic calculations demonstrate that the alternative volumetric re-examination 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 degrees of circumferential extent) 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 RVCH with Alloy 600 nozzles examined per current requirements.

Service Experience As documented in MRP-375 (Reference 3), 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 consecutive years of service for thousands of Alloy 690 steam generator tubes, and more than 22 consecutive years of service for thick-wall and thin-wall Alloy 690 applications. This substantial operating experience includes service at pressurizer and hot-leg

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1) 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 (Reference 1) on 13 of the 41 replacement RVCHs currently operating in the U.S. fleet. This data supports an FOI in time of at least 5 to 20 to detectable PWSCC when compared to service experience of Alloy 600 in similar applications.

Factors of Improvement (FOI) for Crack Initiation Alloy 690 is highly resistant to PWSCC due to its approximate 30% chromium content. Per MRP-115 (Reference 6), it was noted that Alloy 82 CGR is 2.6 times slower than Alloy 182. There is no strong evidence for a difference in Alloy 52 and 152 CGRs. Therefore data used to develop FOIs 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 FOI approach was applied in a conservative manner in MRP-375 (Reference 3) using multiple data. As discussed in MRP-375 (Reference 3), laboratory and plant data demonstrate an FOI in excess of 20 in terms of the time to PWSCC initiation. Conservatively, credit was not taken for the improved resistance of Alloy 690/52/152 to PWSCC initiation in the MRP-375 (Reference 3) analyses.

Factors of Improvement for Crack Growth MRP-375 (Reference 3) also analyzed laboratory PWSCC CGR data for the purpose of assessing FOI values for growth. Data analyzed to develop a conservative FOI include laboratory specimens with significant 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 (Reference 3) considered the most current worldwide set of available PWSCC CGR data for Alloy 690/52/152 materials.

Figure 3-2 of MRP-375 (Reference 3), compares data from Alloy 690 specimens with less than 10% cold work and the statistical distribution from MRP-55 (Reference

7) describing the material variability in CGR for Alloy 600. Most of the laboratory comparisons were bounded by an FOI of 20, and all were bounded by a factor of 10.

Most data support an 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 (Reference 3, relative to the distribution from MRP-55) and for the Alloy 52/152 weld metal (relative to the distribution from MRP-115, Reference 6) as shown in Figure 3-6 of MRP-375 (Reference 3). Based on the data, it is conservative to assume an FOI of between 10 and 20 for CGRs.

5

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

As discussed in the technical basis documents (References 6 and 7) for RVCHs with Alloy 600 nozzles, effective time for crack growth is the principal basis for setting the appropriate re-examination interval to adequately detect any PWSCC. For RVCHs with Alloy 600 nozzles and Alloy 82/182 attachment welds, ASME Code Case N-729-1 (Reference 1) as conditioned by 10 CFR 50.55a, limits the interval between subsequent volumetric or surface inspections to reinspection years (RIY) = 2.25. The RIY parameter, which is referenced to a RVCH temperature of 600°F, limits the time available for potential crack growth between inspections. United States pressurized water reactor (PWR) inspection experience for RVCHs with Alloy 600 nozzles has confirmed that the RIY = 2.25 interval results in a suitably conservative inspection program.

Including an anticipated 3.4°F RVCH temperature increase due to changing fuel design, the representative RVCH operating temperature at PVNGS Units 1, 2, and 3 is 596.3°F (Reference 8). This RVCH operating temperature would result in an RIY temperature adjustment factor of 0.911 (versus the reference temperature of 600°F) using the activation energy of 31 kcal/mol for crack growth of ASME Code Case N-729-1 (Reference 1). Laboratory PWSCC CGR testing for Alloy 690 wrought material by multiple investigators (References 9, 10, and 11) has shown thermal activation energy values comparable to the standard activation energy applied to model growth of Alloy 600/82/182 (31 kcal/mol or 130 kJ/mol). Thus, it is appropriate to apply this standard activation energy for modeling crack growth of Alloy 690/52/152 plant components. Conservatively assuming that the effective full power years (EFPY) of operation accumulated at the three PVNGS RVCHs since RVCH replacement is equal to the calendar years since replacement, the RIY for the requested extended interval of 15 years would be RIY = (0.911) x (15 years) =

13.67. The FOI implied by this RIY value for PVNGS is FOI = (13.67)/(2.25) = 6.1.

Thus, a nominal interval of 15 calendar years for the PVNGS RVCHs implies an FOI of 6.1 versus the standard interval for RVCHs with Alloy 600 nozzles. It is emphasized that the FOI value of 6.1 implied by the requested extension periods represents a level of reduction in PWSCC CGR versus that for Alloy 600/82/182 that is completely bounded by the laboratory data compiled in EPRI MRP- 375 when accounting for material variability. Given the lack of PWSCC detected to date in any PWR plant applications of Alloy 690/52/152, the simple FOI assessment clearly supports the requested period of extension.

The attachment to the enclosure provides further support for the requested alternative inspection interval based on the available laboratory PWSCC CGR data and the FOI approach. The enclosure provides responses to the requests for additional information (RAI) that the NRC has transmitted to other licensees in the context of similar relief requests (see Section 7, Precedent). The attachment to the enclosure describes the materials tested for data points within a factor of 12 below the MRP-55 (Reference 7) and MRP-115 (Reference 6) CGR curves for the 75*^

percentile of material variability. It is concluded that the available CGR data support

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1) the use of an FOI of at least 12 when setting the volumetric or surface examination interval for RVCHs with Alloy 690 nozzles and Alloy 52/152 attachment welds. The FOI value of 6.1 associated with the proposed alternative for the PVNGS RVCHs is conservatively smaller than an FOI of 12.

Moreover, per the attachment to the enclosure, the CGR data do not show any susceptibility concerns specific to the nozzle or weld materials of the PVNGS RVCHs. There are not any relevant similarities between (a) the data points within a factor of 6.1 below the MRP-55 (Reference 7) and MRP-115 (Reference 6) lines in

• • ^i^%ures of the attachment to the enclosure that deterministically represent the 75*'^ percentile of material variability and (b) the associated nozzles and weld material used in the current RVCHs at the PVNGS units. The PVNGS replacement RVCHs were fabricated by Doosan Heavy Industries using Alloy 690 nozzle material produced by Doosan Heavy Industries per ASME SB-166. The nozzle J-groove welds were produced using Alloy 52 (ASME SFA-5.14 ERNiCrFe-7), Alloy 52 (ERNiCrFe-7A to ASME Code Case 2142-2 requirements), and Alloy 152 (ASME SFA-5.11 ENiCrFe-7) weld material. None of the Alloy 690 or Alloy 690 HAZ data points in the attachment to the enclosure (which are limited to data with at most 20%

added cold work) were produced for specimens of control rod drive mechanism (CRDM) nozzle material that was supplied by Doosan. Also, there are no other similarities that indicate any specific concern for elevated PWSCC susceptibility of the RVCH ndzzles at PVNGS in comparison to other RVCHs with Alloy 690 nozzles.

Furthermore, 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 potentially the material variability in the weld consumable (e.g., composition). The test weld specimens in the attachment to the enclosure should not be associated with particular fabrication categories of RVCHs because the test welds used to produce the CGR data compiled in MRP-375 (Reference 3) are not identified with any particular fabricator of replacement RVCHs.

Previous Examinations of the PVNGS RVCHs All of the replacement RVCH partial-penetration welded nozzles of each PVNGS RVCHs were examined prior to service. During fabrication, there were no indications detected by dye penetrant examination (PT), i.e., PT white. Ultrasonic examination (UT) and eddy current examination (ET) of 100% of the nozzles were performed following hydrostatic testing with no unacceptable indications.

A bare metal visual examination (VE) was performed of the PVNGS replacement RVCHs in 2014 on Unit 1 and Unit 2 and in 2015 on Unit 3 in accordance with ASME Code Case N-729-1 (Reference 1), Table 1, item B4.30. This visual examination was performed by VE 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.

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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 (Reference 3) 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 at 613°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 EFPY. Assuming an FOI of 10 to 20 as previously established for Alloy 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 an FOI of 10 to 20 resulted in a median time of 176 EFPY 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 (Reference 3). All cases were bounding and support an inspection interval greater than is being proposed. It is important to note that the operating temperatures of the PVNGS RVCHs are less than 601 °F and well within the bounds of the assumptions.

Deterministic calculations performed in MRP-375 (Reference 3) 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 Alloy 690 nozzle material 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 (Reference

12) technical basis report for inspection requirements for RVCHs with Alloy 600 nozzles, 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 United States RVCHs. 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 CGR for the replacement nickel-base alloys (with no credit for improved resistance to initiation), the probabilistic results with the alternative inspection regime show:

8

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

1) An effect on nuclear safety substantially within the acceptance criterion applied in the MRP-117 (Reference 13) technical basis for Alloy 600 RVCHs,

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

Conclusion In summary, the basis for extending the intervals from once each interval (nominally 10 calendar years) to once every 15 calendar years is based on plant service experience, FOI studies using laboratory initiation and growth data, deterministic modeling, and probabilistic study results. The results of the analysis show that the alternative proposed frequency results in a significantly reduced effect on nuclear safety when compared to a RVCH with Alloy 600 nozzles and examined per the current requirements. The minimum FOI implied by the requested extension period represents a level of reduction in PWSCC CGR versus that for Alloys 600/82/182 that is completely bounded by the laboratory data compiled in MRP-375 (Reference

3) when accounting for heat-to-heat variability of Alloy 600 and weld-to-weld variability of Alloy 82/182. The proposed revised interval will continue to provide reasonable assurance of structural integrity.

Additional assurance of structural integrity is provided by periodic visual examinations of the PVNGS RVCHs. The visual examinations and acceptance criteria as required by item B4.30 of Table 1 of ASME Code Case N-729-1 (Reference 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 (Reference 3), the visual examination requirement of the outer surface of the RVCH 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 RVCH 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.

6. DURATION OF PROPOSED ALTERNATIVE The proposed alternative is requested for the fourth ISI interval because utilizing the proposed examination frequency will require the examination to be performed in the fourth interval.

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

7. PRECEDENT There have been submittals from multiple plants to request an alternative from the frequency of ASME Code Case N-729-1 (Reference 1) for volumetric or surface examinations of RVCHs with Alloy 690 nozzles. Furthermore, the NRC has developed Safety Evaluations to permit the extension of the interval for volumetric or surface inspections per ASME Code Case N-729-1 (Reference 1). The first of these was Arkansas Nuclear One, Unit 1, and some subsequent requests including the associated RAI and Safety Evaluations (SE) are shown in the table below.

NRC ADAMS Accession No.

Request for Relief Additional NRC Safety Request Information RAI Response Evaluation Plant (RAI) Status Arkansas Nuclear ML14118A477 ML14258A020 ML14275A460 ML14330A207 Accepted One, Unit 1 Arkansas Nuclear ML16173A297 None None ML17018A283 Accepted One, Unit 1 Beaver Valley, ML14290A140 None None ML14363A409 Accepted Unit 1 Beaver Valley, Under Unit1 ML17044A440

- - - NRC Review Calvert Cliffs, ML15201A067 None None ML15327A367 Accepted Units 1 & 2 Comanche Peak, ML15120A038 None None ML15259A004 Accepted Unit 1 D.C. Cook, ML15023A038 None None ML15156A906 Accepted Units 1 & 2 J.M. Farley, ML15111A387 None None ML15104A192 Accepted Unit 2 North Anna, ML14283A044 None None ML15091A687 Accepted Unit 2 Prairie Island, ML14258A124 ML15030A008 ML15036A252 ML15125A361 Accepted Units 1 and 2 H.B. Robinson, ML14251A014 ML14294A587 ML14325A693 ML15021A354 Accepted Unit 2 Salem, ML15098A426 None None ML15349A956 Accepted Uniti St. Lucie, ML14206A939 ML14251A222 ML14273A011 ML14339A163 Accepted Uniti St. Lucie, Under Uniti ML17045A357 NRC Review St. Lucie, ML16076A431 None None ML16292A761 Accepted Unit 2

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

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: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375), EPRI, Palo Alto, CA: 2014

4. 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 [NRC ADAMS Accession No. ML041680546]

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

[NRC ADAMS Accession No. ML041680506]

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

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

8. Westinghouse Electric Company, Upper Head Mean Temperature Evaluation to Support Implementation of NGF for PVNGS Units 1, 2 and 3, Document N001-0205-00147, January 2016

9. 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]

10. 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

11. 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

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

Materials in Nuclear Power Systems - Water Reactors, TMS (The Minerals, Metals & Materials Society), 2011

12. Materials Reliability Program: Probabilistic Fracture Mechanics Analysis ofPWR Reactor Pressure Vessel Top Head Nozzle Cracking (MRP-105 NP), EPRI, Palo Alto, CA: 2004 [NRC ADAMS Accession No. ML041680489]

13. Materials Reliability Program: Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PVVR Plants (MRP-117), EPRI, Palo Alto, CA; 2004 [NRC ADAMS Accession No. ML043570129]

Enclosure Relief Request 55 - Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

Attachment 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 13

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 12100 Sunrise Valley Drive, Suite 220 Reston, VA 20191 PH 703.657.7300 FX 703.657.7301

)omiiiioii 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 0 Original Issue /I, Olu'tc.

K. J. Fuhr M. Biirkardt G. A. Wliite ejiSJS 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.

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1 Introduction............................................................................................ 1 0 2 Discussion ofData PointsfromMRP-375 [2]................................................. 3 0 2.1 Data Points Above a Hypothetical 12.0 Factor of Improvement Line in Figure 3-1, 3-3,and 3-5of 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 AAaterials.................................................................................................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 o

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

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

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

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and Available in References [17] and [18]; < 22% Cold Work..................... 24 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/Ki........................................................................................ 24 0 Figure 21. Plot of da/dt versus K\ for Alloy 690 Data from All Laboratories, > 10 & <

20% Cold Work, CRDM and Bar Material, Constant Load or K\ 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 K\.........25 0 Figure 23. Plot of da/dt versus K\ for Alloy 52/152 Data from All Laboratories, > 10 & ^

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

Dominion Engineering, 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 Tnvestigaciones Energeticas, Medioambientales y Tecnologicas 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 RIY Re-Inspection Year RV Reactor Vessel RVCH Reactor Pressure Closure Head UNS Unified Numbering System

)ominioiiEnpeerin?,lnc 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 faetors 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 the Re-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 RV head nozzles operating at 613°F.' A temperature of 613°F 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 of MRP-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 To calculate the implied FOI for the bounding RV top head operating temperature of 613°F, the re-inspection year (RIY) parameter for a requested examination interval of 20 years is compared with the N-729-1 interval for Alloy 600 nozzles of RIY = 2.25. The representative head operating temperatures of 613°F corresponds to an RIY temperature adjustment factor of 1.38 (versus the reference temperature of 600°F) using the activation energy of 31 kcal/mol (130 kJ/mol) for crack growth 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 RIY for a requested extended period of 20 years would be (1.38)(19.6) =

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

Dominion [npmiD?, 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 of MRP-375,^ 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*' 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 of MRP-375 indicate a factor of improvement 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 developed not to describe bounding crack growth rate behavior but rather reflect 75* 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 FOl 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 of MRP-375 represents an unnecessary compounding of conservatisms. Essentially none of the data presented lies within a statistical FOl 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, the Re-Inspection Year (RIY) = 2.25 inspection interval developed for heads with Alloy 600 nozzles reflects the possibility of crack Figures 3-2,3-4, and 3-6 of MRP-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.

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

Dominion Enpeeiin;, Inc TN-5696-00-02, Rev. 0 growth rates being many times higher than the deterministic 75^ percentile values per MRP-55 and MRP-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 (75*' 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 of MRP-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 75**' percentile curve, out of a total of 75 points shown in Figure 3-1 of MRP-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 Tecnologicas (CIEMAT) and two that were tested by Argonne National Laboratory (ANL).

- Two of the points tested by CIEMAT are from specimen 9ARBI, comprised of Alloy 690 plate material, loaded to 37 MPa(m)° ^ and tested at 340°C and 15 cc H2/kg H2O

[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 of 25-50 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) / control element drive mechanism (CEDM) nozzle

lominion Enpeerin?, 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 of Valinox 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 Wilkg H2O, 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 H2O, 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 are for specimens 9T1 and 9T2 (loaded to roughly 36 MPa(m)°^), and the remaining six data are from specimens 9T5 or 9T6 (loaded to roughly 27 MPa(m)°^). The results for 9T1 and 9T2 are contained in Reference [9]; the final data for 9T5 and 9T6 are contained in EPRl 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 (<

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 WP142 CRDM nozzle material that were not cold worked and were tested at 21 to 24 MPa(m)° ^ 320°C, and 23 cc H2/kg H2O [10].

The intergranular engagement for these specimens was extremely low (almost entirely transgranular). 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 ofMRP-375. 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 EPRl. The following points are within a factor of 12.0 below the MRP-55 deterministic crack growth rate for Alloy 600:

)ominioiiEn;ineerin;,lnc TN-5696-00-02, Rev. 0 There are eight points within a factor of 12.0 below the MRP-55 75^^ percentile curve, out of a total of 34 points shown in Figure 3-3 of MRP-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 of Valinox CRDM nozzle material heat WP142 and Alloy 152 filler (Special Metals heat WC43E9),

tested at 320°C and 23 cc Ha/kg H2O [11]. Five of the points are from specimens CF690-CR-1 and CF690-CR-3 (loaded to roughly 28 to 32 MPa(m)°^) [11], and the other point is from specimen 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 heat WP142 is provided here as Figure 4. As discussed below, PPU conditions under which five of these six points were obtainedmay result in accelerated crack growth relative to plant conditions.

The remaining two points are from CIEMAT testing of specimens 19ARHI and 19ARH2, comprised of welded Alloy 690 plate material, tested at 340°C and 15 cc Ha/kg H2O, and loaded to roughly 37 MPa(m)^ [8]. A plot of crack growth rate versus K for the 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 ofMRP-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 MRP-115 deterministic crack growth rate for Alloy 182; There are 19 points within a factor of 12.0 below the MRP-115 75^ percentile curve, out of a total of 212 points shown in Figure 3-5 of MRP-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 of CRDM nozzle J-groove welds because the dilution zone where Alloy 52/152 contacts the low-alloy steel RV head is below the stainless steel cladding. A plot of crack growth rate versus K for the Alloy 152 data from MRP-375 for heat WC83F8 is provided here as Figure 6.

Four of the remaining points, including the point closest to the MRP-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 of MRP-375.

The remaining 14 data points represent four specimens from Alloy 152 weld material (Special Metals heat WC04F6) that were tested by ANL at 320°C and 23 cc H2/kg H2O ([15] and [10]). Ten of these points are for specimen A152-TS-5 at loads of about 28, 32, and 48 MPa(m)°^ [14]. The other four points were obtained at loads of

Dominion Enpeerin;, Inc TN-5696-00-02, Rev. 0 27 MPa(m) for specimen N152-TS-1 and 30 MPa(m)°^ for specimens A152-TS-2 and AI52-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 N152-TS-1), 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 Directly Applicable to Plant Conditions As described above. Section 3 of MRP-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-115 lines, 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 10%.

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 11 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 of 83 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

Dominion Enpeorin?, Inc TN-5696-00-02, Rev. 0 equations of MRP-55 and MRP-115. The MRP-55 and MRP-115 deterministic crack growth rate equations are not bounding equations, but rather reflect the 75'*^ 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 less 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 of MRP-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 of MRP-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 oiany

Dominion Enpeerin?, 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 of MRP-55 and MRP-115.

2.3 Data Specific to Arqonne 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 [I?]'*, 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 of 86 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 10% 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 Data for Alloy 690 Wrought Material 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 The data in Reference [16] are augmented by the crack growth rate data for Alloys 52/152 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 of Reference [ 16], Enclosure 2 contains the PNNL tabular data, and Enclosure 4 contains the ANL tabular data.

)ominioDEii;ineerin?,lnc 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) / 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 AAaterials 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 (or Laboratory Specimen Material Exhibitins 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 of MRP-375 and (b) the associated nozzles and weld material used in the RV heads in U.S. PWRs are as follows;

)oniiniooEnpeei1n?,lnc 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 of less than 10% (the threshold for inclusion in Figure 3-1 of MRP-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 [1] deterministic crack growth rate in Figure 3-1 of MRP-375 were produced for specimens of Alloy 690 CRDM nozzle material that was supplied by Valinox Nucleaire. 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 of PPU 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 ofMRP-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 of MRP-375 [2]. There are no relevant similarities between (a) the Alloy 52 and 152 data points above a crack growth rate 12.0 times lower than the MRP-115 [2]

Alloy 182 deterministic crack growth rate in Figure 3-5 of MRP-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:

Dominion tnpeerin?, 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 of the 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 of MRP-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 690/52/152 than for Alloys 600/82/182/132 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 of PWSCC initiation is necessary to produce a flaw that may grow via PWSCC.

Laboratory and plant experience show that Alloys 690/52/152 are substantially more resistant to PWSCC initiation than Alloys 600/82/182 [2]. PWSCC has not been shown to be an active degradation mode for Alloys 690/52/152 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 / AWS ERNiCrFe-7), Alloy 52M (UNS N06054 / AWS ERNiCrFe-7A), Alloy 52MSS (UNS N06055 / AWS ERNiCrFe-13), Alloy 52i (AWS ERNiCrFe-15), Alloy 152 (UNS W86152/AWS ENiCrFe-7), and Alloy 152M (UNS W86152 / AWS 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 RV head penetration nozzles or their Alloy 52/152 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

lominion tnpmn?, 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 Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials (MRP-55)

Revision 1, EPRl, Palo Alto, CA: 2002. 1006695. [freely available at ^\'w^\^eDri.com1

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

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-375), EPRl, Palo Alto, CA: 2014. 3002002441. [freely available at www.epri.coml

5. Materials Reliability Program: Inspection Plan for Reactor Vessel Closure Head Penetrations in US. PWR Plants (MRP-117), EPRl, 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 Assessmentfor US. PWR Plants (MRP-1 lONP), EPRl, 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), EPRl, Palo Alto, CA:

2004. 1007834. [ML041680489]

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

9. D. Gomez-Briceno, J. Lapena, M. S. Garcia, L. Castro, F. Perosanz, L. Francia, and K.

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

Dominion Enpeerin?, 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-7137, 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/EPRlMeeting, June 6-7, 2011. [MLl 11661946]

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 ofMaterials 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 of Intergranular Stress Corrosion Crack Growth of Alloy 152 Weld Metals in Simulated PWR Primary Water, 16th International Conference on Environmental Degradation of

-.ti: 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 of Preliminary Primary Water Stress Corrosion Cracking Data for Alloys 690, 52, and 152, October 30, 2014. [ML14322A587]

18. Pacific Northwest National Laboratory Investigation ofStress Corrosion Cracking in Nickel-Base Alloys, NUREG/CR-7103, Vol. 2, April 2012.

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.

Dominion {o^ineerin?, Inc TN-5696-00-02, Rev. 0 Data from Individual Heats ACIEMAT MRP-55 Curve/1

^ MRP/12 h PPU data are represented with 1.E-12 open symbols Data are adjusted for temperature (325°C).

Q = 130kJ/mol Stress Intensity Factor (MPaVm)

Figure 1. Plot of Crack Growth Rate (da/dt) versus Stress Intensity Factor (Ki) for Alloy 690 Data from Plate Material Tested by CIEMAT ACIEMAT PNNL MRP-55 Curve/1 1.E-10 MRP/12 h PPU data are represented with 1.E-12 open symbols Data are adjusted for temperature (325°C).

Q = 130kJ/mol 30 35 40 Stress Intensity Factor (MPaVm)

Figure 2. Plot of da/dt versus K\ for Alloy 690 Data from Heat WP787

Dominion Enpeerin;, Inc TN-5696-00-02, Rev. 0 1.E-09

• ANL PNNL MRP-55 Curve/1 PPL) data are represented with 1.E-12 open symbols Data are adjusted for temperature (325°C).

Q = 130kJ/mol 10 15 20 25 30 35 40 45 Stress Intensity Factor (MPaVm)

Figure 3. Plot of da/dt versus K\ for Alloy 690 Data from Heat WP142 MRP-55 Curve/1

^ MRP/12 h O

PPL) data are represented with 1.E-12 open symbols Data are adjusted for temperature (325°C).

Q=130kJ/mol 1.E-13 Stress Intensity Factor (MPaVm)

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

)ominion Enpeerin;, Inc. TN-5696-00-02, Rev. 0 aCIEMAT MRP-55 Curve/1 A

I MRP/12 PPU data are represented with 1.E-12 open symbols Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 30 35 40 Stress Intensity Factor (MPaVm)

Figure 5. Plot of da/dt versus K\ for Alloy 690 HAZ Data from Plate Material Tested by CIEMAT 1.E-09 GE-GRC PNNL MRP-115 Curve/1

(/)

1 Dilution 2 zone point MRP/12 Fusion line point o

u SE 6 1.E-12 --

Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 25 30 35 40 45 Stress Intensity Factor (MPaVm)

Figure 6. Plot of da/dt versus K\ for Alloy 152 Data from Heat WC83F8

Dominion &?ineerin?, Inc TN-5696-00-02. Rev. 0 GE-GRC MRP-115 Curve/1 1.E-10 -- PNNL J MRP/12 PPU data are Black-filled ANL data present represented with growth rates during the open symbols environmental pre-crack period 1.E-12 and should not be included.

Data are adjusted for temperature (325°C).

Q = 130kJ/mol 1.E-13 Stress Intensity Factor (MPaVm)

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

Dominion Enpoerin?, Inc TN-5696-00-02, Rev. 0 Data Most Applicable to Plant Conditions 1.E-09

• ANL

• Bettis MRP-55 ACIEMAT Curve/1 1.E PNNL H MRP/12 h Data are adjusted for temperature (325°C).

Q = 130kJ/mol Stress Intensity Factor (MPaVm)

Figure 8. Plot of da/dt versus #Ci for Alloy 690 Data from All Laboratories, < 10% Cold Work, Constant Load or K\

• ANL 0.9 --

• Bettis ACIEMAT

• PNNL The data points at IE-13 were reported as no

_____ growth._____

Data are adjusted for FOI = 12 temperature (325°C) and stress intensity factor.

Q = 130kJ/mol K = 30 MPaVm 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (m/s)

Figure 9. Cumulative Distribution Function of Adjusted da/dt for Alloy 690 Data from All Laboratories, < 10% Cold Work, Constant Load or K\

loroinion Enpeerin?, Inc TN-5696-00-02. Rev. 0

• ANL ACIEMAT MRP-55 GE-GRC Curve/1 1.E-10 --

PNNL MRP/12 Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 Stress Intensity Factor (MPaA/m)

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

1.0

• ANL 0.9 ACIEMAT 0.8 GE-GRC i 0.7 PNNL I 0.6

° 0.5 The data points at IE-13

• 0.4 were reported as no

_____ growth."

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

0.1 Q=130 kJ/mol K = 30 MPaVm 0.0 1.E-13 1.E-12 1.E-11 1.E-10 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 Ki

Dominion [npoerin;, Inc TN-5696-00-02, Rev. 0

• ANL ACIEMAT MRP-115 GE-GRC Curve/1 1.E PNNL j MRP/12 r*

1.E-12 Data are adjusted for temperature (325°C).

Q=130 kJ/mol 30 35 40 Stress Intensity Factor (MPaVm)

Figure 12. Plot of da/dt versus Ki for Alloy 52/152 Data from All Laboratories, < 10% Cold Work, Constant Load or K\

• ANL

- ACIEMAT GE-GRC APNNL - ^ MRP-115 C

o (FOI = 1)

__________^_______ /  : /

3

§ b

7^ The data points at 1E-13 were reported as no 3 _____ growth."

E o

3 X Data are adjusted for FOI = 12 temperature (325°C) and

/ stress intensity factor.

Q = 130 kJ/mol

/ K = 30 MPaVm 0.0 ' I '

1.E-13 1.E-12 1.E-11 1.E-10 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 Ki

Dominion {npeerin?, Inc TN-5696-00-02, Rev. 0 Comparison of Partial Period Unloading (PPU) Conditions vs. Constant Load Conditions Data are adjusted for Specimen temperature (325°C) (Q = 130 kJ/mol) andK(30MPaVm)

PPU Data 1.E-10 Const. Load Data 1.E-12 I

100 Hold Time (Hours)

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

Dominion En?ineerin?, Inc TN-5696-00-02, Rev. 0 Compilation of ANL and PNNL Data Box and arrow show the ratio between the MRP-55 curve and the data point MRP-55 Curve/1 1.E-10 MRP/12 1.E -

• ANL CL OANL PPU PNNL Data are adjusted for temperature (325°C).

Q = 130 kJ/mol

^4 '

Stress Intensity Factor (MPaVm)

Figure 15. Plot of da/dt versus K\ for Alloy 690 Data Produced by ANL and PNNL and Available in Reference [17]; < 22% Cold Work

• ANL CL PNNL The data points at IE-13 are MRP-55 treated as no growth, fFOI = 1) consistent with MRP-375.

Data are adjusted for FOI = 12 temperature (325°C) and stress intensity factor.

0.1 Q = 130 kJ/mol K = 30 MPaVm 1.E-13 1.E-11 1.E-10 1.E-09 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]; < 22% Cold Work and Constant Load//Ci

Dominion Enpmn;, Inc TN-5696-00-02, Rev. 0 1.E-09

• ANLCL Box and arrow show the OANL PPU ratio between the MRP-55 PNNL curve and the data point MRP-55 Curve/1 1.E-10 t0 1 MRP/12 h 1.E-11 I

o u

2 o 1.E-12 Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 25 30 35 40 45 Stress Intensity Factor (MPaVm)

Figure 17. Plot of da/dt versus K\ for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17]; < 22% Cold Work

• ANLCL PNNL The data points at 1E-13 are MRP-55 treated as no growth, (FOI = 1) consistent with MRP-375.

Data are adjusted for FOI = 12 temperature (325°C) and stress intensity factor.

0.1 Q = 130 kJ/mol K = 30 MPaVm 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (m/s)

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/Ki

Dominion {npmn?, Inc TN-5696-00-02, Rev. 0 Box and arrow show the MRP-115 ratio between the MRP-115 Curve/1 curve and the data point MRP/12 O

• ANL CL OANL PPU 1.E-12 PNNL CK (NUREG)

Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 Stress Intensity Factor (MPaVm)

Figure 19. Plot of da/dt versus K\ for Alloy 52/152 Data Produced by ANL and PNNL and Available in References [17] and [18]; < 22% Cold Work 1.0

1. ANLCL

^  : ./ /

PNNL CK (NUREG)

.9 0-7 +/-1 r MRP-115 (FOI = 11 to I 0.5

---

1---------------------------

4 The data points at IE-13 4- were reported as no I 0.4 - 4- growth.

E 5 0.3 4- Data are adjusted for 0.2 FOI = 12 temperature (325°C) and

/ stress intensity factor.

0.1 Q = 130kJ/mol K = 30MPaVm 0.0 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth 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 LoadIKi

Dominion Enfinpprin?, Inc tn-5696 oo-o2, Rev. o Data for Less than 20% Cold Work from All Laboratories 1.E-09 ACIEMAT MRP-55 Curve/1 1.E GE-GRC PNNL

^ MRP/12 h 1.E-12 --

Data are adjusted for temperature (325°C).

Q = 130kJ/mol 30 35 40 Stress Intensity Factor (MPaVm)

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

^AMEC

. aciemat

-- BGE-GRC

-- APNNL

//J/ / /

O VJ

• "c 0.6

[___ /

/' /

I I 0.5 ^ ^ / MRP-55 The data points at IE-13

• were reported as no I 0.4 >> -------- (^UI-1) growth.

A /

E 0.3 ^___L___ /i Data are adjusted for

/ /

0.2 -------------- FOI-12 temperature (325°C) and

/-----------------------

_____ ^ '___________/ stress intensity factor.

0.1 Q = 130kJ/mol K = 30 MPaVm 0.0 ..................................................................... ................... "* "il---------------- 1 ............................. ....................^

' 1---------- 'I'll 1.E-13 1.E-12 1.E-11 1.E-10 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 K\

Dominion Enfineerin?, Inc TN-5696^02, Rev. 0 GE-GRC MRP-115 Curve/1 1.E PNNL J MRP/12 P 1.E-12 Data are adjusted for temperature (325°C).

Q = 130kJ/mol Stress Intensity Factor (MPaVm)

Figure 23. Plot of da/dt versus K\ for Alloy 52/152 Data from All Laboratories, > 10 & < 20% Cold Work, Constant Load or K\

• ANL ACIEMAT 0.8 -- GE-GRC PNNL The data points at IE-13 were reported as no

_____ growth.

Data are adjusted for FOI = 12 temperature (325°C) and stress intensity factor.

Q = 130 kJ/mol K = 30 MPaVm 1.E-11 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 fCi