1CAN061601, Request for Alternative from Volumetric/Surface Examination Frequency - Requirements of ASME Code Case N-729-1 - Arkansas Nuclear One, Unit 1

Request for Alternative from Volumetric/Surface Examination Frequency - Requirements of ASME Code Case N-729-1 - Arkansas Nuclear One, Unit 1

ML16173A297
Person / Time
Site:	Arkansas Nuclear
Issue date:	06/17/2016
From:	Pyle S Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
1CAN061601, N-729-1
Download: ML16173A297 (46)

Text

{{#Wiki_filter:* ~Entergy Entergy Operations, Inc. 1448 S .R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Regulatory Assurance Arkansas Nuclear One 1CAN061601 June 17, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington , DC 20555

SUBJECT:

Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1 Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51

REFERENCES:

1. Letter from Entergy to U.S. NRC, Regarding Relief Request Number AN01-ISl-024, "Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1 ,"

dated April 28, 2014 (ML14118A477)

2. Letter from U.S. NRC to Entergy, "Arkansas Nuclear One, Unit 1 -

Request for Alternative AN01-ISl-024 from Volumetric/Surface Examination Frequency Requirements of the American Society of Mechanical Engineers Code Case N-729-1 (TAC No. MF4022)," dated December 23, 2014(ML14330A207)

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(z)(1 ), Entergy Operations, Inc. (Entergy) hereby requests NRC approval of the attached lnservice Inspection (ISi) Request for Alternative for Arkansas Nuclear One, Unit 1 (AN0-1). The request is associated with the volumetric/surface examination frequency requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, 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/surface examination be performed within one inspection interval (nominally 10 calendar years) of its inservice date for a replacement reactor vessel closure head (RVCH). The AN0-1 replacement RVCH was placed in service in December 2005 and would nominally require volumetric/surface examination by December 2015. Via Reference 1, Entergy requested a one-time deferral of the inspection by approximately 2.5 years until the refueling outage scheduled to commence in April of 2018. The NRC subsequently approved this alternative, via Reference 2, and pursuant to the former 10 CFR 50.55a(a)(3)(i), now 10 CFR 50.55a(z)(1 ).

1CAN061601 Page 2 of 3 In 2014, the Electric Power Research Institute (EPRI) published a technical report entitled "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375)" that provides justification for extending the volumetric/surface examination interval from 1O years to 20 years. Entergy believes that the conclusions reached in this technical report are appropriate and applicable to establish an extended examination frequency for AN0-1. The one-time deferral previously requested by Entergy in Reference 1 was intended to provide sufficient time for the NRG to review and accept the conclusions reached in MRP-375, as well as the time to make appropriate ASME Code changes. The ASME Code has already adopted a volumetric/surface examination of two inspection intervals (nominally 20 years) in Table 1, Item 84.40 of ASME Code Case N-729-5. However, it is now expected that additional time beyond the currently authorized deferral will be necessary for NRG to have all the information needed to consider the conclusions reached in MRP-375. Specifically, the detailed review of crack growth rate data by the international group of experts cited in Reference 2 is currently expected not to be complete until mid to late 2017 . Thus, Entergy is requesting that the previously approved one-time deferral of the frequency requirements of Table 1 of ASME Code Case N-729-1, Item 84.40 be extended until the refueling outage scheduled to commence in April of 2021. This is approximately 5.5 years beyond the nominal 1O years required by ASME Code Case N-729-1 , and two (2) AN0-1 refueling cycles or approximately 3 years beyond the previously approved deferral period. The justification for this Alternative request is provided in Attachment 1 to this letter. Attachment 2 provides additional information on the available relevant crack growth rate data supporting . This Request for Alternative concludes that the improved performance of the replacement head materials versus the Alloy 600/82/182 materials of the original RVCHs justifies the inspection interval extension and that this extension provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1 ). As documented in Section 6 of Attachment 1, the NRG has approved similar extensions for other replacement RVCHs of more than 5 years beyond the nominal 10-year interval required by ASME Code Case N-729-1 in order to align with scheduled refueling outages. The NRG has approved extensions of 6 years for this purpose. Entergy requests the total one-time deferral of approximately 5.5 years for AN0-1 for the purpose of aligning with the scheduled AN0-1 refueling outages. EPRI Technical Report MRP-375 is a non-proprietary document that is available through the EPRI Website. This submittal contains no regulatory commitments. The last refueling outage to comply with the current alternative approved via Reference 2 is scheduled to commence in April of 2018. In order to provide planning for this outage, Entergy requests approval of the proposed Request for Alternative by July 31, 2017 .

1CAN061601 Page 3 of 3 If you have any questions or require additional information , please contact me. Sincerely, Attachments:

1. Request for Alternative AN01 -ISl-026

2. Dominion Engineering, Inc. Technical Note TN-5696-00-02, "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," Revision 0, March 2015.

cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington , TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. 0. Box 310 London , AR 72847 U. S. Nuclear Regulatory Commission Attn : Mr. Stephen Koenick MS 0-8B1A One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Attachment 1 to 1CAN061601 Request for Alternative AN01-ISl-026

1CAN061601 Page 1of9 REQUEST FOR ALTERNATIVE AN01-ISl-026 Request for Relief in Accordance with 10 CFR 50.55a(z)(1) Inspection of Reactor Vessel Closure Head Nozzles in Accordance with ASME Code Case N-729-1 as Conditioned by 10CFR50.55a Components I Numbers: Reactor Vessel Closure Head Penetration Nozzles 0-1 through 0-69 fabricated with Alloy 690 penetration tubes and Alloy 52/152 partial-penetration welds. Code Classes: American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) , Class 1 Code

References:

ASME Section XI , Division 1, Code Case N-729-1 , as conditioned by 10 CFR 50.55a(g)(6)(ii)(D) Examination Category: Table 1 of ASME Code Case N-729-1 , Item No. , 84.40 Description : Examination Categories for Class 1 Pressurized Water Reactor (PWR) Reactor Vessel Upper Head Inspection Interval Arkansas Nuclear One, Unit 1 (AN0-1 ) I Fourth and Fifth 10-Year lnservice Applicability: Inspection (ISi) Intervals (May 31 , 2008 through May 30, 2027)

1. APPLICABLE CODE REQUIREMENTS:

The Code of Federal Regulations (CFR) 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 [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 Eon 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.

1CAN061601 Page 2 of 9 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 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 every ASME Section XI 10-year ISi interval (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.

2. REASON FOR REQUEST:

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 Reactor Vessel Closure Head (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 (Reference 2). Using plant and laboratory data, Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) Report MRP-375 (Reference 7) 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. The NRC has previously approved a one-time deferral of the inspection by approximately 2.5 years for AN0-1 until the refueling outage scheduled to commence in April of 2018 (Reference 3). This deferral was intended to provide sufficient time for the NRC to review and accept the conclusions reached in MRP-375 (Reference 7) . However, it is now expected that additional time beyond the currently authorized deferral will be necessary for NRC to have all the information needed to consider the conclusions reached in MRP-375 (Reference 7) . Specifically, the detailed review of crack growth rate data by the international group of experts cited in Reference 3 is currently expected not to be complete until mid to late 2017. Thus, Entergy is requesting that the previously approved one-time deferral of the frequency requirements of Table 1 of ASME Code Case N-729-1 , Item 84.40 be extended until the refueling outage scheduled to commence in April of 2021 . This is approximately 5.5 years beyond the nominal 10 years required by ASME Code Case N-729-1, and two (2) AN0-1 refueling cycles or approximately 3 years beyond the previously approved deferral period . As described below, the requested alternative will maintain an acceptable level of quality and safety.

1CAN061601 Page 3 of 9

3. PROPOSED ALTERNATIVE:

Entergy is requesting relief from the exam frequency requirements of Code Case N-729-1 (Reference 1), Item 84.40 for performing volumetric and/or surface exams of the AN0-1 RVCH penetrations . Specifically, this would allow volumetric or surface examinations currently scheduled for the April 2018 refueling outage in accordance with a prior relief request (Reference 4) to be extended to the April 2021 refueling outage (approximately 5.5 years beyond the nominal 10 years required by ASME Code Case N-729-1 in order to align with a scheduled refueling outage). This request applies to the Item 84.40 inspection frequencies only.

4. BASIS FOR ALTERNATIVE:

As discussed in the original ASME technical basis document (Reference 2), the inspection frequency of ASME Code Case N-729-1 (Reference 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 (Reference 5) , which was summarized in the safety assessment for RVCHs in MRP-110 (Reference 6). 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 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 with respect to the performance of Alloy 690 and Alloy 52/152 (Reference 2). Further evaluations were performed to demonstrate the acceptability of extending the inspection intervals for Code Case N-729-1 , Item 84.40 components and documented in MRP-375 (Reference 7). 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, factors of improvement (FOi) studies using laboratory data, deterministic study results, and probabilistic study results. Per MRP-375 (Reference 7) , much of the laboratory data indicated a 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 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 FOi. 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 exa mined per current requirements.

1CAN061601 Page 4 of 9 Service Experience As documented in MRP-375 (Reference 7) , 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 13 of the 41 replacement RVCHs currently operating in the U.S. fleet. This data supports a FOi of at least 5 to 20 to detectable PWSCCs when compared to service experience of Alloy 600 in similar applications. In addition , based on communications with Duke Energy for Oconee Units 1, 2, and 3 and Exelon Generation for Three Mile Island Unit 1 (TMl-1 ), these units received head replacements in the 2003 to 2004 timeframe. These four units have received volumetric head examination in accordance with ASME Code Case N-729-1 . These examinations did not reveal any recordable indications. The replacement RVCHs for the Oconee units were manufactured by B&W Canada. The TMl -1 replacement RVCH and the AN0-1 replacement RVCH were fabricated by the same manufacturer (AREVA) and used Alloy 690 nozzle material produced by the same material supplier (Valinox) . Being B&W plant designs, these units would have similar head configurations and design operating conditions to that of AN0-1. Entergy believes that these examination results additionally support the low likelihood of the potential to experience PWSCC for the AN0-1 RVCH for the extension period . FOi for Crack Initiation Alloy 690 is highly resistant to PWSCC due to its approximate 30% chromium content. Per MRP-115 (Reference 8), 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 FOi 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 7) using multiple data. As discussed in MRP-375, laboratory and plant data demonstrate a FOi 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. FOi for Crack Growth MRP-375 (Reference 7) also assessed laboratory PWSCC crack growth rate data for the purpose of assessing FOi values for growth. Data analyzed to develop a conservative FOi include laboratory specimens with substantial levels of cold work. It is important to note that much of the data used to support Alloy 690 CG Rs 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.

1CAN061601 Page 5 of 9 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 (Reference 9) describing the material variability in CGR for Alloy 600. Most of the laboratory comparisons were bounded by a FOi of 20 , and all were bounded by a factor of 10. Most data support a FOi of much larger th an 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 (Reference 8)) 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 re-inspection years (RIY) = 2.25 constraint on the volumetric or surface reexamination interval of ASME Code Case N-729-1 correspond to an interval of 2.0 Effective Full Power Years (EFPYs). A prior relief request (Reference 4) was approved by NRC (Reference 3) for a nominal interval of 12.5 calendar years, and this request included a calculation demonstrating that an interval of 12.5 years at AN0-1 corresponds to an FOi of 7.7 . The calculation shows that the value of the corresponding FOi is directly proportional to the duration of the alternative interval. The requested Alternative constitutes a factor of 1.24 longer inspection interval (15.5/12.5 = 1.24) , resulting in the same relative increase to the implied FOi (7.7x1 .24 = 9.5). Thus, a nominal interval of 15.5 calendar years for the AN0-1 replacement head implies a FOi of 9.5 versus the standard interval for heads with Alloy 600 nozzles. It is emphasized that the FOi of 9.5 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 . Attachment 2 provides further support for the requested alternative inspection interval based on the available laboratory PWSCC crack growth rate data and the FOi approach. The attachment provides responses to the requests for additional information that the NRC has transmitted to other licensees in the context of similar relief requests (see Section 6, Precedent) . Attachment 2 describes the materials tested for data points within a factor of 12 below the MRP-55 [9] and MRP-115 [8] crack growth rate curves for the 75th percentile of material variability. Attachment 2 also compares these test materials to the specific nozzle and weld materials used in the AN0-1 replacement head . 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 AN0-1 replacement head. Previous Examinations of the AN0-1 Replacement Head A preservice volumetric examination of the replacement RVCH partial-penetration welded nozzles was performed prior to head installation at AN0-1 . There were no recordable indications identified during the preservice volumetric examinations of the nozzle tube in the area of the J-groove welds. 1 1 Furthermore , the NRC concluded that the AN0-1 replacement RVCH met its design requirements as documented in an Inspection Report dated February 13, 2006 [10] using Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection." The inspectors reviewed numerous design and manufacturing documents including the certified material test reports, heat treatment records, welding processes, as well as the preservice volumetric examinations. No findings of significance were identified regarding the replacement RVCH .

1CAN061601 Page 6 of 9 Bare metal VEs were performed of the AN0-1 replacement RVCH in spring 2010 and in spring 2015 in accordance with ASME Code Case N-729-1 , Table 1, Item 84.30. These visual examinations were performed by VT-2 qualified examiners on the outer surface of the RVCH including the annulus area of the penetration nozzles. These examinations 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 (Reference 7) 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 °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 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 temperatu re 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° 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 that being proposed . It is important to note that the operating temperature of the AN0-1 RVCH is conservatively taken to be 613 °F (Reference 4) and is within the bounds of the assumptions . Deterministic calculations performed in MRP-375 (Reference 7) 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 (Reference 11) technical basis report for inspection requirements for heads 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 U.S. heads. The outputs of the probabil istic 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:

1CAN061601 Page 7 of 9

1) An effect on nuclear safety substantially within the acceptance criterion applied in the MRP-117 (Reference 12) technical basis for Alloy 600 heads, and

2) 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 once every second interval (nominally 20 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 substantially reduced effect on nuclear safety when compared to a head 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 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 inspections of the TMl-1 and Oconee 1, 2, and 3 heads in 2012 and 2013. Furthermore, the visual examinations and acceptance criteria as required by Item 84 .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 (Reference 7), 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.

5. DURATION OF PROPOSED ALTERNATIVE:

The proposed Alternative is requested for the remainder of the fourth and fifth ISi intervals because utilizing the proposed examination frequency will require the examination to be performed in the fifth interval.

6. PRECEDENT:

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 prior AN0-1 request for relief and some subsequent requests,

1CAN061601 Page 8 of 9 including the associated status at the time of submittal of this request, are shown below. Alternative intervals greater than 15 years have previously been granted in order to align with scheduled refueling outages. The approved Calvert Cliffs Units 1 & 2 alternative (noted in the table below) permitted an inspection interval not to exceed 16 years in order to align with scheduled refueling outages, and the St. Lucie 2 submittal under NRG review requests an interval of 15.5 years to align with scheduled refueling outages. NRC ADAMS Accession No. Request for Plant Relief Additional RAI NRC Safety Status Request Information Response Evaluation (RAI) Arkansas Nuclear One, ML14118A477 ML14258A020 ML14275A460 ML14330A207 Accepted Unit 1 Beaver Valley, ML14290A140 None None ML14363A409 Accepted Unit 1 Calvert Cliffs, ML15201A067 None None ML15327A367 Accepted Units 1 & 2 Comanche ML15120A038 None None ML15259A004 Accepted Peak 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 Unit 1 St. Lucie, ML14206A939 ML14251A222 ML14273A011 ML14339A163 Accepted Unit 1 St. Lucie, Under NRC ML16076A431 Unit 2 Review

1CAN061601 Page 9 of 9

7.

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. Letter from U.S. NRG to Entergy, "Arkansas Nuclear One, Unit 1 - Request for Alternative AN01-ISl-024 from Volumetric/Surface Examination Frequency Requirements of the American Society of Mechanical Engineers Code Case N-729-1 (TAC No. MF4022)," dated December 23, 2014 . [NRG ADAMS Accession No. ML14330A207]

4. Letter from Entergy to U.S. NRG, Regarding Relief Request Number AN01-ISl -024 ,

      "Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1 ," dated April 28, 2014 . [NRG ADAMS Accession No. ML14118A477]

5. 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 ; NRG ADAMS Accession No. ML041680546]

6. Materials Reliability Program: Reactor Vessel Closure Head Penetration Safety Assessment for U.S. PWR Plants (MRP-110NP) , EPRI , Palo Alto , CA: 2004. 1009807-NP.

[M L041680506]

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

9. 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]

10. NRG Inspection Report for Arkansas Nuclear One - NRG Integrated Inspection Report 05000313/2005010 dated February 13, 2006. [ML060460164]

11 . 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]

12. 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 ; NRG ADAMS Accession No. ML043570129]

Attachment 2 to 1CAN061601 Dominion Engineering, Inc. Technical Note TN-5696-00-02 "Assessment of Laboratory PWSCC Crack Growth Rate Data Complied for Alloys 690, 52, and 152 with Regard to Factors of Improvement (FOi) versus Alloys 600 and 182" Revision 0 March 2015

Dominion fn~ineerin~, Inc - 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 Sun rise Valley Drive, Suite 220 11 Reston. VA 20191 111 PH 703.657.7300 111 FX 703.657.7301

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 0 Original Issue 3{~ rt. rs,..,.fc-J.f G./f. Oh~i<...- A-~)~

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K. J. Fuhr M. Burkard! G. A White G. A White Associate Engineer

• Associ ate 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.

ii

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 Hypothetical 12.0 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 CROM 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

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 LIST OF FIGURES 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 CIEMAT.. ...................... .. .. ...................14 0 Figure 2. Plot of da/dt versus K1for Alloy 690 Data from Heat WP787 .. .. .. .. .......... ... .. ....... 14 0 Figure 3. Plot of da/dt versus K1for Alloy 690 Data from Heat WP142 .. .. .. .. .......... .... .... .. .. 15 0 Figure 4. Plot of da/dt versus K1for 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 CI EMAT ................. .. ...... .. ... .. .. .. .. .................... ... .. .... ... ..... ... .. .. .. .. .. .. ...... .. .. .... .... .. 16 0 Figure 6. Plot of da/dt versus K1for 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, :s; 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, :s: 10% Cold Work, Constant Load or K1 ........................ ...... .. ... 18 0 Figure 10. Plot of da/dt versus K1for Alloy 690 HAZ Data from All Laboratories, :s: 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, :s; 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, :s: 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, :s; 10% Cold Work, Constant Load or K1 .. .. .. .. .... .... .... .. .. .. .20 0 Fig ure 14. Plot of da/dt versus Loading Hold Time (for PPU testing) or Test Segment Duration (for Constant Ki/Load Testing) from Heat WP787 ............ .... .............. .. 21 0 Figure 15. Plot of da/dt versus K1for Alloy 690 Data Produced by ANL and PNNL and Available in Reference [17]; :s: 22% Cold Work .................... ............ .. ................. 22 0 Fig ure 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 K1for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17]; :s; 22% Cold Work ...... ...... .. .. .. .... .......... .... .. .... .. 23 0 Figure 18. Cumulative Distribution Function of Adjusted da/dt Alloy 690 HAZ Data Produced by ANL and PNNL [17]; :s; 22% Cold Work and Constant Load/Ki.. .. .. 23 0 Fig ure 19. Plot of da/dt versus K1 for Alloy 52/152 Data Produced by ANL and PNNL iv

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/Kr ....... .. ..... ........ .............. .... .... ........ ...... ........ .. ....................... ............. ... .... 24 0 Figure 21. Plot of da/dt versus Kr for Alloy 690 Data from All Laboratories, > 10 &~ 20% Cold Work, CROM and Bar Material, Constant Load or Kr Testing .... ...... ... 25 0 Figure 22. Cumulative Distribution Function of Adjusted da/dt Alloy 690 Data from All Labs, ~ 20% Cold Work, CROM and Bar Material, Constant Load or Kr ............ 25 0 Figure 23. Plot of da/dt versus Kr for Alloy 52/152 Data from All Laboratories, > 10 & ~ 20% Cold Work, Constant Load or Kr .... .. .. .. .. ........ ...................................... .. .... .26 0 Figure 24. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data from All Laboratories, ~ 20% Cold Work, Constant Load or Kr .. .. .. .. ........ .. .. .... .... .. .. .. .. .... 26 0 v

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 RIY Re-Inspection Year RV Reactor Vessel RVCH Reactor Pressure Closure Head UNS Unified Numbering System vi

Dominion [n~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 [l] and MRP-115 [2]. In addition, an assessment is made of the available laboratory CGR data for the potential concern of elevated CG Rs 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. 1 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 1 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 ofRIY = 2.25 . The representative head operating temperatures of613°F corresponds to an RIY temperature adjustment factor of 1.38 (versus the reference temperature of600°F) using the activation energy of 31 kcal/mo! (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 of20 years would be (1.38)(19 .6) = 27 .0. The FOi implied by this RIY value is (27.0)/(2 .25) = 12.0.

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 [l] and Alloy 182 in MRP-115 [2]. A comparison between the cumulative distributions of the crack growth rates for Alloys 690/521152 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 (75th 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 beyond 20 and all of the data 3 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 75th 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, 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 ofMRP-375 show cumulative distribution functions of the variability in crack growth rate normalized for temperature and crack loading (i.e., stress intensi ty 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

Dominion [n~ineerin~, Inc. TN-5696-00-02, Rev. 0 growth rates being many times higher than the deterministic 75th 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 (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 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 facto r 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 ofMRP-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 9ARB 1, comprised of Alloy 690 plate material, loaded to 37 MPa(m)0 *5 , and tested at 340°C and 15 cc H2/kg H20 [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) I control element drive mechanism (CEDM) nozzle 3

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 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 H2/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 H2 0 , 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 9Tl and 9T2 (loaded to roughly 36 MPa(m) 0*5), and the remaining six data are from specimens 9T5 or 9T6 (loaded to roughly 27 MPa(m) 0 *5). 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 (:'.:: 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-l and C690-LR-2, comprised ofValinox heat number WP142 CRDM nozzle material that were not cold worked and were tested at 21 to 24 MPa(m) 0 *5 , 320°C, and 23 cc H 2/kg H2 0 [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 of MRP-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 EPRI. The following points are within a factor of 12.0 below the MRP-55 deterministic crack growth rate for Alloy 600:

4

Dominion fn~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 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 H2/kg H20 [11]. Five of the points are from specimens CF690-<;R-_1 and CF69?-CR-3 (loaded to roughly 28 to 32 MPa(m) 05 ) [1 n, other pomt is from specimen CF690-CR-4 (loaded to roughly 22 MPa(m) *) [12]. A and the 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 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 at 340°C and 15 cc H2/kg H 20 , and loaded to roughly 37 MPa(m) 05 [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 75th percentile curve, out of a total of 212 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 of CRDM nozzle J-groove welds because the dilution zone where Alloy 5211 52 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 H20 ([15] and [10]). Ten of these points are for specimen Al52-TS-5 at loads of about 28, 32, and 48 MPa(m) 05 [14]. The other four points were obtained at loads of 5

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 27 MPa(m) 0 .s for specimen N152-TS-l and 30 MPa(m) 0 *5 for specimens A152-TS-2 and A152-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 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-1 15 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 l 0 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 6

Dominion [n~ineerin~, Inc. TN-5696-00-02, Rev. 0 equations of MRP-5 5 and MRP-115. The MRP-5 5 and MRP-115 deterministic crack growth rate equations are not bounding equations, but rather reflect the 75lh 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 hour were screened out of the database for Alloys 821182/132.

The greater resistance of Alloys 690/521152 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 600/821182 and compare this against the full variability of the Alloy 690/52/152 data most applicable to plant conditions. The lack of any 7

Dominion fn~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 of MRP-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 4 recently released by NRC [ 17] , 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 4 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. 8

Dominion fn~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

Dominion fn~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 of less than l 0% (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 [l] 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 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 of MRP-3 75 [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 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 ofreplacement 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

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 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 690/521152 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 I AWS ERNiCrFe-7), Alloy 52M (UNS N06054 I AWS ERNiCrFe-7A), 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 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 11

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 avai lable 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 REFEREN CES

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, 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 l ,"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), 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-11 7), 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-110NP), EPRI, Palo Alto, CA: 2004. 1009807-NP.

[ML04 l 680506]

7. Materials Reliability Program: Probabilistic Fracture Mechanics Analysis ~lPWR 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 6901152152 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

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 Presented at: Alloy 6901152152 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. 109-1 25 , 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 NRCIEPRI Meeting, 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 En vironmental 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 6901152152 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 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 In vestigation of Stress Corrosion Cracking in Nickel-Base Alloys, NUREG/CR-7103 , Vol. 2, April 2012.

19. Ma terials 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

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 Data from Individual Heats 1.E-09

                        ~     CIEMAT I

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Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09

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Dominion En~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09 _ o ANL D GE-GRC f;1 MRP-115 i-- Curve/1 I '-

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Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 Data Most Applicable to Plant Conditions 1.E-09 OANL

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Figure 11. Cumulative Distribution Function of Adjusted da/dt for Alloy 690 HAZ Data from All Laboratories, :5 10% Cold Work, Constant Load or Ki 19

Dominion En~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09 - OANL

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          .5                                                                                                                were reported as "no
          -3      0.4  +--~11-------------+------1                                                                                  growth ."

E  ::=====z=.========~ 8 0.3 +----illP= - - - - - - - - - - - --1-- - -- - l Data are adjusted for FOi = 12 temperature (325°C) and 0.2 +!Alll!--- - - - - - - - - - - - - == ==-- ---1 stress intensity factor. 0.1 c::J-- - - - - - - - - - - - . . . . - - - - - - - 1 I Q = 130 kJ/mol K = 30 MPa.Ym 0.0 .__....._.........................""""'"+--"'_.__,,::::::..................1--___,--~:.::::Wu::+====::;::=:::L=::;:::::;::;:~ 1.E-13 1.E-12 1.E-1 1 1.E-10 1.E-09 Crack Growth Rate (mis) 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

Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 Comparison of Partial Period Unloading (PPU) Conditions vs. Constant Load Conditions 1.E-09 Data are adjusted for  := Specimen temperature (325°C) (Q = 130 -

                                                                                                            -    9T1 kJ/mol) and K (30 MPa--/m) -

1.E-10 IPPU Data f- - - - - - - - - - - - - -, 9T2 Const. Load Data ~ Vi' ~ 9T3 E

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                                                                                 -----~-*

I I 1.E-13 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 Ki/Load Tes ting) from Heat WP787 21

Dominion En~ineerin~, Inc. TN-5696-00-02, Rev. 0 Compilation of ANL and PNNL Data 1.E-09 Box and arrow show the ratio between the MRP-55 1 MRP-55 curve and the data point I Curve/1 I 1.E-10

                                                                                  ~                           96 Vi'

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                                          ..r
                                               ~

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                                                                                                                        .T
                                                                                                                          ....                      OANL PPU

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                              .                                         T Data are adjusted for temperature (325°C) .

Q = 130 kJ/mol 1.E-13 T - TTT T T 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa'1m) 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 1.0 ,-------------,.---- ,,.,..-::-;..-~----::;;:;;-------, 0.9 +------------,.-...~------------------l 0.8 --------~~-~------.f------------1

           .g      0.7 +------~------------1------------                                                                                            0-A-NL_C_L--.-1

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§ 0.6 +-----...~----1-------1---------l + PNNL Vi

            ~ 0.5                                                                                                                The data points at 1E-13 are
            .5                                                                                                                       treated as "no growth ,"
            -3     0.4                                                                                                             consistent with MRP-375.

E 8 0.3 Data are adjusted for IFOi = 12 I temperature (325°C) and 0.2 &------=;==~---+-------l stress intensity factor. 0.1 ----~-------#---------. Q=130kJ/mol

                                       ./                                                                                                K = 30 MPa'1m 0.0 .__.....::.~........'""'+---~::;_._..u....y.._ _.___.__.__._.__.__._._1--__.__::c==:i:+=====~

1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 Crack Growth Rate (mis) 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/Ki 22

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 1.E-09 o ANL CL Box and arrow show the OANL PPU ratio between the MRP-55

                                + PNNL                curve and the data point                                            I MRP-55 r I Curve/1 I 1.E-10 Cil
        ].
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                                                   ~
                                         ,,r
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• Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 . T T I 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa..Jm) Figure 17. Plot of da/dt versus Ki for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17]; S 22% Cold Work 1.0 -r----------------:---~~---::::::--------, ,,.., 0.9 -+--- - - - - - - - ---r-,-- -#-- - - - - -- - - - - - - - ----< 0.8 *

                       +-----------~----___,,__                                                            _ _ _ _ _ _ _ _ __,
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i *

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            ~ 0.5                                                                                            The data points at 1E-13 are
            *5                                                                                                   treated as "no growth ,"
            -3     0.4                                                                                          consistent with MRP-375.

E 8 0.3 Data are adjusted for IFOi = 12 I temperature (325°C) and 0*2 ~------=;==:!.__ ___ ..J-_ _ _ _ ___j stress intensity factor. 0.1 _ _ _ ____,,__----~~------; Q = 130 kJ/mol

                                      ./                                                                              K = 30 MPavfm 0.0 _...-=:c...:............._ _ _~::._,_,L...L..i.+--'--'--L......L...1....L..LI.+----'--__:r::==:::q::::====::rl.I 1.E-13                      1.E-12              1.E-11                         1.E-10             1.E-09             1.E-08 Crack Growth Rate (mis)

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

Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09 Box and arrow show the - I MRP-115 i-::: - ratio between the MRP-115 1.E-10 __....--1 Curve/1 I I 5.5 LJ 9.2 I curve and the data point

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                               + PNNL CK (NUREG)                                   *

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                                                                            ~
                                                                            ~
                                                                                    ~
                                                                                    ¥
                                                                                               ~
                                                                                               ¥
                                                                                                                  ~
                                                                                                                  ¥ 10            15          20      25        30              35              40             45           50        55       60 Stress Intensity Factor (MPa-1m)

Figure 19. Plot of da/dt versus K1 for Alloy 52/152 Data Produced by ANL and PNNL and Available in References [17] and [18]; :S 22% Cold Work 1.0 o ANL CL 0.9

• PNNL CK (NUREG) 0 70 --

                                                                                                                                                 /
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                                                                                                     ~

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                                                                                                                                      ./

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                                                                                            ,I                          The data points at 1E-13 T
          *z>                                                                                                              were reported as "no
            ~ 04 I                                       growth ."

i .

8 E 0.3 I Data are adjusted for 0.2 ....* IFOI = 121 temperature (325°C) and stress intensity factor. 0.1 I Q = 130 kJ/mol 0.0

                      *t
                       ~

_, / K = 30 MPa..Jm 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (mis) Figure 20. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data Produced by ANL and PNNL ([17] and [18]); :S 22% Cold Work and Constant Load/Ki 24

Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 Data for Less than 20% Cold Work from All Laboratories 1.E-09 rv~c=i~~~~~~~~~~~~~~~~~ CIEMAT MRP-55 ~~::1

                                           ~~~~iiiiii~Cui~e/~1~

1.E-13 -+--'-~-'--+-~~+-'-~-'--+-~~+-'-~-'--+-~~._.._~._._~~._.._~._._~...__.__. 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-Vm) Figure 21. Plot of da/dt versus Ki for Alloy 690 Data from All Laboratories, > 10 & ~ 20% Cold Work, CROM and Bar Material, Constant Load or Ki Testing v AMEC

                         ~ CIEMAT DGE-GRc i---------<;,__,.--- - - --+- - - - - - - - ----1
           § 0.7         + PNNL
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         ~ o.5 1-----x-r------;:::::===+/-::::.~---i The data points at 1E-13
         *.=;                                                                                             were reported as "no
         -3      0.4                                                                                             growth."

8E 0.3 +----_.,.__ _...___ _ _ _~'--------l Data are adjusted for 0.2 -+--- -- I---<~~~,___ _ _ _, __ _ _ _ _---< temperature (325°C) and stress intensity factor. 0.1 Q =130 kJ/mol K= 30 MPa./m 0.0 ~-=-.................~--~:.................L...l..f---'----'"------'-..J.-'--'-'-!-----'-__::r:::==::q=====~ 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 Ki 25

Dominion [n~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09 _ DGE-GRC

                         -                  ~ MRP-115 p                 -

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1. E-13 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-Vm)

Figure 23. Plot of da/dt versus Ki for Alloy 52/152 Data from All Laboratories, > 10 & S 20% Cold Work, Constant Load or Ki 1.0 -,;:::===:;-----------------n-- --=--~-----, OANL /" 0.9 ~ CIEMAT 1----------,--7'rtJ-~----.---------..~--i 0.8 D GE-GRC 1-----=!"l'~---------1-------,...,...,..,,,,.,,,......L...,,-,r--~

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          ~ 0.5     -t--- - - - .,.,_- - - - - - - - - -__,_- -_,                                              The data points at 1E-13
          ._E;                                                                                                   were reported as "no
          -S   0.4 +--     ---.r:--- - - - - - - - - - - ---1                                                        growth ."

E ~=====~=========: 8 0.3 T-...i-1-'-- - - - - - - - ---.==t=-- - - - - i Data are adjusted for FOi = 12 temperature (325°C) and 0.2 +r:11F-------------~------1 1 stress intensity factor. 0.1 t-t------------~------t Q = 130 kJ/mol K = 30 MPav'm 0.0 .__......_.............................._ _.._._;_,,,,,.~..................~_.--~:i::W:::;:+:==:::;::~:::;:::::;::::;::w:::;'.,j 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (mis) Figure 24. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data from All Laboratories, S 20% Cold Work, Constant Load or Ki 26

~Entergy Entergy Operations, Inc. 1448 S .R. 333 Russellville, AR 72802 Tel 479-858-4704 Stephenie L. Pyle Manager, Regulatory Assurance Arkansas Nuclear One 1CAN061601 June 17, 2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington , DC 20555

SUBJECT:

Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1 Arkansas Nuclear One, Unit 1 Docket No. 50-313 License No. DPR-51

REFERENCES:

1. Letter from Entergy to U.S. NRC, Regarding Relief Request Number AN01-ISl-024, "Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1 ,"

dated April 28, 2014 (ML14118A477)

2. Letter from U.S. NRC to Entergy, "Arkansas Nuclear One, Unit 1 -

Request for Alternative AN01-ISl-024 from Volumetric/Surface Examination Frequency Requirements of the American Society of Mechanical Engineers Code Case N-729-1 (TAC No. MF4022)," dated December 23, 2014(ML14330A207)

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(z)(1 ), Entergy Operations, Inc. (Entergy) hereby requests NRC approval of the attached lnservice Inspection (ISi) Request for Alternative for Arkansas Nuclear One, Unit 1 (AN0-1). The request is associated with the volumetric/surface examination frequency requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, 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/surface examination be performed within one inspection interval (nominally 10 calendar years) of its inservice date for a replacement reactor vessel closure head (RVCH). The AN0-1 replacement RVCH was placed in service in December 2005 and would nominally require volumetric/surface examination by December 2015. Via Reference 1, Entergy requested a one-time deferral of the inspection by approximately 2.5 years until the refueling outage scheduled to commence in April of 2018. The NRC subsequently approved this alternative, via Reference 2, and pursuant to the former 10 CFR 50.55a(a)(3)(i), now 10 CFR 50.55a(z)(1 ).

1CAN061601 Page 2 of 3 In 2014, the Electric Power Research Institute (EPRI) published a technical report entitled "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375)" that provides justification for extending the volumetric/surface examination interval from 1O years to 20 years. Entergy believes that the conclusions reached in this technical report are appropriate and applicable to establish an extended examination frequency for AN0-1. The one-time deferral previously requested by Entergy in Reference 1 was intended to provide sufficient time for the NRG to review and accept the conclusions reached in MRP-375, as well as the time to make appropriate ASME Code changes. The ASME Code has already adopted a volumetric/surface examination of two inspection intervals (nominally 20 years) in Table 1, Item 84.40 of ASME Code Case N-729-5. However, it is now expected that additional time beyond the currently authorized deferral will be necessary for NRG to have all the information needed to consider the conclusions reached in MRP-375. Specifically, the detailed review of crack growth rate data by the international group of experts cited in Reference 2 is currently expected not to be complete until mid to late 2017 . Thus, Entergy is requesting that the previously approved one-time deferral of the frequency requirements of Table 1 of ASME Code Case N-729-1, Item 84.40 be extended until the refueling outage scheduled to commence in April of 2021. This is approximately 5.5 years beyond the nominal 1O years required by ASME Code Case N-729-1 , and two (2) AN0-1 refueling cycles or approximately 3 years beyond the previously approved deferral period. The justification for this Alternative request is provided in Attachment 1 to this letter. Attachment 2 provides additional information on the available relevant crack growth rate data supporting . This Request for Alternative concludes that the improved performance of the replacement head materials versus the Alloy 600/82/182 materials of the original RVCHs justifies the inspection interval extension and that this extension provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1 ). As documented in Section 6 of Attachment 1, the NRG has approved similar extensions for other replacement RVCHs of more than 5 years beyond the nominal 10-year interval required by ASME Code Case N-729-1 in order to align with scheduled refueling outages. The NRG has approved extensions of 6 years for this purpose. Entergy requests the total one-time deferral of approximately 5.5 years for AN0-1 for the purpose of aligning with the scheduled AN0-1 refueling outages. EPRI Technical Report MRP-375 is a non-proprietary document that is available through the EPRI Website. This submittal contains no regulatory commitments. The last refueling outage to comply with the current alternative approved via Reference 2 is scheduled to commence in April of 2018. In order to provide planning for this outage, Entergy requests approval of the proposed Request for Alternative by July 31, 2017 .

1CAN061601 Page 3 of 3 If you have any questions or require additional information , please contact me. Sincerely, Attachments:

1. Request for Alternative AN01 -ISl-026

2. Dominion Engineering, Inc. Technical Note TN-5696-00-02, "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," Revision 0, March 2015.

cc: Mr. Marc L. Dapas Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington , TX 76011-4511 NRC Senior Resident Inspector Arkansas Nuclear One P. 0. Box 310 London , AR 72847 U. S. Nuclear Regulatory Commission Attn : Mr. Stephen Koenick MS 0-8B1A One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. Bernard R. Bevill Arkansas Department of Health Radiation Control Section 4815 West Markham Street Slot #30 Little Rock, AR 72205

Attachment 1 to 1CAN061601 Request for Alternative AN01-ISl-026

1CAN061601 Page 1of9 REQUEST FOR ALTERNATIVE AN01-ISl-026 Request for Relief in Accordance with 10 CFR 50.55a(z)(1) Inspection of Reactor Vessel Closure Head Nozzles in Accordance with ASME Code Case N-729-1 as Conditioned by 10CFR50.55a Components I Numbers: Reactor Vessel Closure Head Penetration Nozzles 0-1 through 0-69 fabricated with Alloy 690 penetration tubes and Alloy 52/152 partial-penetration welds. Code Classes: American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code) , Class 1 Code

References:

ASME Section XI , Division 1, Code Case N-729-1 , as conditioned by 10 CFR 50.55a(g)(6)(ii)(D) Examination Category: Table 1 of ASME Code Case N-729-1 , Item No. , 84.40 Description : Examination Categories for Class 1 Pressurized Water Reactor (PWR) Reactor Vessel Upper Head Inspection Interval Arkansas Nuclear One, Unit 1 (AN0-1 ) I Fourth and Fifth 10-Year lnservice Applicability: Inspection (ISi) Intervals (May 31 , 2008 through May 30, 2027)

1. APPLICABLE CODE REQUIREMENTS:

The Code of Federal Regulations (CFR) 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 [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 Eon 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.

1CAN061601 Page 2 of 9 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 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 every ASME Section XI 10-year ISi interval (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.

2. REASON FOR REQUEST:

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 Reactor Vessel Closure Head (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 (Reference 2). Using plant and laboratory data, Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) Report MRP-375 (Reference 7) 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. The NRC has previously approved a one-time deferral of the inspection by approximately 2.5 years for AN0-1 until the refueling outage scheduled to commence in April of 2018 (Reference 3). This deferral was intended to provide sufficient time for the NRC to review and accept the conclusions reached in MRP-375 (Reference 7) . However, it is now expected that additional time beyond the currently authorized deferral will be necessary for NRC to have all the information needed to consider the conclusions reached in MRP-375 (Reference 7) . Specifically, the detailed review of crack growth rate data by the international group of experts cited in Reference 3 is currently expected not to be complete until mid to late 2017. Thus, Entergy is requesting that the previously approved one-time deferral of the frequency requirements of Table 1 of ASME Code Case N-729-1 , Item 84.40 be extended until the refueling outage scheduled to commence in April of 2021 . This is approximately 5.5 years beyond the nominal 10 years required by ASME Code Case N-729-1, and two (2) AN0-1 refueling cycles or approximately 3 years beyond the previously approved deferral period . As described below, the requested alternative will maintain an acceptable level of quality and safety.

1CAN061601 Page 3 of 9

3. PROPOSED ALTERNATIVE:

Entergy is requesting relief from the exam frequency requirements of Code Case N-729-1 (Reference 1), Item 84.40 for performing volumetric and/or surface exams of the AN0-1 RVCH penetrations . Specifically, this would allow volumetric or surface examinations currently scheduled for the April 2018 refueling outage in accordance with a prior relief request (Reference 4) to be extended to the April 2021 refueling outage (approximately 5.5 years beyond the nominal 10 years required by ASME Code Case N-729-1 in order to align with a scheduled refueling outage). This request applies to the Item 84.40 inspection frequencies only.

4. BASIS FOR ALTERNATIVE:

As discussed in the original ASME technical basis document (Reference 2), the inspection frequency of ASME Code Case N-729-1 (Reference 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 (Reference 5) , which was summarized in the safety assessment for RVCHs in MRP-110 (Reference 6). 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 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 with respect to the performance of Alloy 690 and Alloy 52/152 (Reference 2). Further evaluations were performed to demonstrate the acceptability of extending the inspection intervals for Code Case N-729-1 , Item 84.40 components and documented in MRP-375 (Reference 7). 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, factors of improvement (FOi) studies using laboratory data, deterministic study results, and probabilistic study results. Per MRP-375 (Reference 7) , much of the laboratory data indicated a 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 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 FOi. 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 exa mined per current requirements.

1CAN061601 Page 4 of 9 Service Experience As documented in MRP-375 (Reference 7) , 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 13 of the 41 replacement RVCHs currently operating in the U.S. fleet. This data supports a FOi of at least 5 to 20 to detectable PWSCCs when compared to service experience of Alloy 600 in similar applications. In addition , based on communications with Duke Energy for Oconee Units 1, 2, and 3 and Exelon Generation for Three Mile Island Unit 1 (TMl-1 ), these units received head replacements in the 2003 to 2004 timeframe. These four units have received volumetric head examination in accordance with ASME Code Case N-729-1 . These examinations did not reveal any recordable indications. The replacement RVCHs for the Oconee units were manufactured by B&W Canada. The TMl -1 replacement RVCH and the AN0-1 replacement RVCH were fabricated by the same manufacturer (AREVA) and used Alloy 690 nozzle material produced by the same material supplier (Valinox) . Being B&W plant designs, these units would have similar head configurations and design operating conditions to that of AN0-1. Entergy believes that these examination results additionally support the low likelihood of the potential to experience PWSCC for the AN0-1 RVCH for the extension period . FOi for Crack Initiation Alloy 690 is highly resistant to PWSCC due to its approximate 30% chromium content. Per MRP-115 (Reference 8), 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 FOi 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 7) using multiple data. As discussed in MRP-375, laboratory and plant data demonstrate a FOi 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. FOi for Crack Growth MRP-375 (Reference 7) also assessed laboratory PWSCC crack growth rate data for the purpose of assessing FOi values for growth. Data analyzed to develop a conservative FOi include laboratory specimens with substantial levels of cold work. It is important to note that much of the data used to support Alloy 690 CG Rs 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.

1CAN061601 Page 5 of 9 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 (Reference 9) describing the material variability in CGR for Alloy 600. Most of the laboratory comparisons were bounded by a FOi of 20 , and all were bounded by a factor of 10. Most data support a FOi of much larger th an 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 (Reference 8)) 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 re-inspection years (RIY) = 2.25 constraint on the volumetric or surface reexamination interval of ASME Code Case N-729-1 correspond to an interval of 2.0 Effective Full Power Years (EFPYs). A prior relief request (Reference 4) was approved by NRC (Reference 3) for a nominal interval of 12.5 calendar years, and this request included a calculation demonstrating that an interval of 12.5 years at AN0-1 corresponds to an FOi of 7.7 . The calculation shows that the value of the corresponding FOi is directly proportional to the duration of the alternative interval. The requested Alternative constitutes a factor of 1.24 longer inspection interval (15.5/12.5 = 1.24) , resulting in the same relative increase to the implied FOi (7.7x1 .24 = 9.5). Thus, a nominal interval of 15.5 calendar years for the AN0-1 replacement head implies a FOi of 9.5 versus the standard interval for heads with Alloy 600 nozzles. It is emphasized that the FOi of 9.5 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 . Attachment 2 provides further support for the requested alternative inspection interval based on the available laboratory PWSCC crack growth rate data and the FOi approach. The attachment provides responses to the requests for additional information that the NRC has transmitted to other licensees in the context of similar relief requests (see Section 6, Precedent) . Attachment 2 describes the materials tested for data points within a factor of 12 below the MRP-55 [9] and MRP-115 [8] crack growth rate curves for the 75th percentile of material variability. Attachment 2 also compares these test materials to the specific nozzle and weld materials used in the AN0-1 replacement head . 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 AN0-1 replacement head. Previous Examinations of the AN0-1 Replacement Head A preservice volumetric examination of the replacement RVCH partial-penetration welded nozzles was performed prior to head installation at AN0-1 . There were no recordable indications identified during the preservice volumetric examinations of the nozzle tube in the area of the J-groove welds. 1 1 Furthermore , the NRC concluded that the AN0-1 replacement RVCH met its design requirements as documented in an Inspection Report dated February 13, 2006 [10] using Inspection Procedure 71007, "Reactor Vessel Head Replacement Inspection." The inspectors reviewed numerous design and manufacturing documents including the certified material test reports, heat treatment records, welding processes, as well as the preservice volumetric examinations. No findings of significance were identified regarding the replacement RVCH .

1CAN061601 Page 6 of 9 Bare metal VEs were performed of the AN0-1 replacement RVCH in spring 2010 and in spring 2015 in accordance with ASME Code Case N-729-1 , Table 1, Item 84.30. These visual examinations were performed by VT-2 qualified examiners on the outer surface of the RVCH including the annulus area of the penetration nozzles. These examinations 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 (Reference 7) 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 °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 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 temperatu re 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° 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 that being proposed . It is important to note that the operating temperature of the AN0-1 RVCH is conservatively taken to be 613 °F (Reference 4) and is within the bounds of the assumptions . Deterministic calculations performed in MRP-375 (Reference 7) 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 (Reference 11) technical basis report for inspection requirements for heads 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 U.S. heads. The outputs of the probabil istic 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:

1CAN061601 Page 7 of 9

1) An effect on nuclear safety substantially within the acceptance criterion applied in the MRP-117 (Reference 12) technical basis for Alloy 600 heads, and

2) 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 once every second interval (nominally 20 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 substantially reduced effect on nuclear safety when compared to a head 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 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 inspections of the TMl-1 and Oconee 1, 2, and 3 heads in 2012 and 2013. Furthermore, the visual examinations and acceptance criteria as required by Item 84 .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 (Reference 7), 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.

5. DURATION OF PROPOSED ALTERNATIVE:

The proposed Alternative is requested for the remainder of the fourth and fifth ISi intervals because utilizing the proposed examination frequency will require the examination to be performed in the fifth interval.

6. PRECEDENT:

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 prior AN0-1 request for relief and some subsequent requests,

1CAN061601 Page 8 of 9 including the associated status at the time of submittal of this request, are shown below. Alternative intervals greater than 15 years have previously been granted in order to align with scheduled refueling outages. The approved Calvert Cliffs Units 1 & 2 alternative (noted in the table below) permitted an inspection interval not to exceed 16 years in order to align with scheduled refueling outages, and the St. Lucie 2 submittal under NRG review requests an interval of 15.5 years to align with scheduled refueling outages. NRC ADAMS Accession No. Request for Plant Relief Additional RAI NRC Safety Status Request Information Response Evaluation (RAI) Arkansas Nuclear One, ML14118A477 ML14258A020 ML14275A460 ML14330A207 Accepted Unit 1 Beaver Valley, ML14290A140 None None ML14363A409 Accepted Unit 1 Calvert Cliffs, ML15201A067 None None ML15327A367 Accepted Units 1 & 2 Comanche ML15120A038 None None ML15259A004 Accepted Peak 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 Unit 1 St. Lucie, ML14206A939 ML14251A222 ML14273A011 ML14339A163 Accepted Unit 1 St. Lucie, Under NRC ML16076A431 Unit 2 Review

1CAN061601 Page 9 of 9

7.

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. Letter from U.S. NRG to Entergy, "Arkansas Nuclear One, Unit 1 - Request for Alternative AN01-ISl-024 from Volumetric/Surface Examination Frequency Requirements of the American Society of Mechanical Engineers Code Case N-729-1 (TAC No. MF4022)," dated December 23, 2014 . [NRG ADAMS Accession No. ML14330A207]

4. Letter from Entergy to U.S. NRG, Regarding Relief Request Number AN01-ISl -024 ,

      "Request for Alternative from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1 ," dated April 28, 2014 . [NRG ADAMS Accession No. ML14118A477]

5. 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 ; NRG ADAMS Accession No. ML041680546]

6. Materials Reliability Program: Reactor Vessel Closure Head Penetration Safety Assessment for U.S. PWR Plants (MRP-110NP) , EPRI , Palo Alto , CA: 2004. 1009807-NP.

[M L041680506]

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

9. 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]

10. NRG Inspection Report for Arkansas Nuclear One - NRG Integrated Inspection Report 05000313/2005010 dated February 13, 2006. [ML060460164]

11 . 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]

12. 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 ; NRG ADAMS Accession No. ML043570129]

Attachment 2 to 1CAN061601 Dominion Engineering, Inc. Technical Note TN-5696-00-02 "Assessment of Laboratory PWSCC Crack Growth Rate Data Complied for Alloys 690, 52, and 152 with Regard to Factors of Improvement (FOi) versus Alloys 600 and 182" Revision 0 March 2015

Dominion fn~ineerin~, Inc - 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 Sun rise Valley Drive, Suite 220 11 Reston. VA 20191 111 PH 703.657.7300 111 FX 703.657.7301

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 0 Original Issue 3{~ rt. rs,..,.fc-J.f G./f. Oh~i<...- A-~)~

                                            ?/z'J/15              3/z. ~/'Z.01$        3fz:1 I z..or.r (3",3JV jio\}

K. J. Fuhr M. Burkard! G. A White G. A White Associate Engineer

• Associ ate 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.

ii

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 Hypothetical 12.0 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 CROM 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

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 LIST OF FIGURES 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 CIEMAT.. ...................... .. .. ...................14 0 Figure 2. Plot of da/dt versus K1for Alloy 690 Data from Heat WP787 .. .. .. .. .......... ... .. ....... 14 0 Figure 3. Plot of da/dt versus K1for Alloy 690 Data from Heat WP142 .. .. .. .. .......... .... .... .. .. 15 0 Figure 4. Plot of da/dt versus K1for 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 CI EMAT ................. .. ...... .. ... .. .. .. .. .................... ... .. .... ... ..... ... .. .. .. .. .. .. ...... .. .. .... .... .. 16 0 Figure 6. Plot of da/dt versus K1for 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, :s; 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, :s: 10% Cold Work, Constant Load or K1 ........................ ...... .. ... 18 0 Figure 10. Plot of da/dt versus K1for Alloy 690 HAZ Data from All Laboratories, :s: 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, :s; 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, :s: 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, :s; 10% Cold Work, Constant Load or K1 .. .. .. .. .... .... .... .. .. .. .20 0 Fig ure 14. Plot of da/dt versus Loading Hold Time (for PPU testing) or Test Segment Duration (for Constant Ki/Load Testing) from Heat WP787 ............ .... .............. .. 21 0 Figure 15. Plot of da/dt versus K1for Alloy 690 Data Produced by ANL and PNNL and Available in Reference [17]; :s: 22% Cold Work .................... ............ .. ................. 22 0 Fig ure 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 K1for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17]; :s; 22% Cold Work ...... ...... .. .. .. .... .......... .... .. .... .. 23 0 Figure 18. Cumulative Distribution Function of Adjusted da/dt Alloy 690 HAZ Data Produced by ANL and PNNL [17]; :s; 22% Cold Work and Constant Load/Ki.. .. .. 23 0 Fig ure 19. Plot of da/dt versus K1 for Alloy 52/152 Data Produced by ANL and PNNL iv

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/Kr ....... .. ..... ........ .............. .... .... ........ ...... ........ .. ....................... ............. ... .... 24 0 Figure 21. Plot of da/dt versus Kr for Alloy 690 Data from All Laboratories, > 10 &~ 20% Cold Work, CROM and Bar Material, Constant Load or Kr Testing .... ...... ... 25 0 Figure 22. Cumulative Distribution Function of Adjusted da/dt Alloy 690 Data from All Labs, ~ 20% Cold Work, CROM and Bar Material, Constant Load or Kr ............ 25 0 Figure 23. Plot of da/dt versus Kr for Alloy 52/152 Data from All Laboratories, > 10 & ~ 20% Cold Work, Constant Load or Kr .... .. .. .. .. ........ ...................................... .. .... .26 0 Figure 24. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data from All Laboratories, ~ 20% Cold Work, Constant Load or Kr .. .. .. .. ........ .. .. .... .... .. .. .. .. .... 26 0 v

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 RIY Re-Inspection Year RV Reactor Vessel RVCH Reactor Pressure Closure Head UNS Unified Numbering System vi

Dominion [n~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 [l] and MRP-115 [2]. In addition, an assessment is made of the available laboratory CGR data for the potential concern of elevated CG Rs 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. 1 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 1 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 ofRIY = 2.25 . The representative head operating temperatures of613°F corresponds to an RIY temperature adjustment factor of 1.38 (versus the reference temperature of600°F) using the activation energy of 31 kcal/mo! (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 of20 years would be (1.38)(19 .6) = 27 .0. The FOi implied by this RIY value is (27.0)/(2 .25) = 12.0.

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 [l] and Alloy 182 in MRP-115 [2]. A comparison between the cumulative distributions of the crack growth rates for Alloys 690/521152 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 (75th 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 beyond 20 and all of the data 3 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 75th 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, 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 ofMRP-375 show cumulative distribution functions of the variability in crack growth rate normalized for temperature and crack loading (i.e., stress intensi ty 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

Dominion [n~ineerin~, Inc. TN-5696-00-02, Rev. 0 growth rates being many times higher than the deterministic 75th 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 (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 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 facto r 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 ofMRP-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 9ARB 1, comprised of Alloy 690 plate material, loaded to 37 MPa(m)0 *5 , and tested at 340°C and 15 cc H2/kg H20 [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) I control element drive mechanism (CEDM) nozzle 3

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 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 H2/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 H2 0 , 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 9Tl and 9T2 (loaded to roughly 36 MPa(m) 0*5), and the remaining six data are from specimens 9T5 or 9T6 (loaded to roughly 27 MPa(m) 0 *5). 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 (:'.:: 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-l and C690-LR-2, comprised ofValinox heat number WP142 CRDM nozzle material that were not cold worked and were tested at 21 to 24 MPa(m) 0 *5 , 320°C, and 23 cc H 2/kg H2 0 [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 of MRP-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 EPRI. The following points are within a factor of 12.0 below the MRP-55 deterministic crack growth rate for Alloy 600:

4

Dominion fn~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 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 H2/kg H20 [11]. Five of the points are from specimens CF690-<;R-_1 and CF69?-CR-3 (loaded to roughly 28 to 32 MPa(m) 05 ) [1 n, other pomt is from specimen CF690-CR-4 (loaded to roughly 22 MPa(m) *) [12]. A and the 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 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 at 340°C and 15 cc H2/kg H 20 , and loaded to roughly 37 MPa(m) 05 [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 75th percentile curve, out of a total of 212 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 of CRDM nozzle J-groove welds because the dilution zone where Alloy 5211 52 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 H20 ([15] and [10]). Ten of these points are for specimen Al52-TS-5 at loads of about 28, 32, and 48 MPa(m) 05 [14]. The other four points were obtained at loads of 5

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 27 MPa(m) 0 .s for specimen N152-TS-l and 30 MPa(m) 0 *5 for specimens A152-TS-2 and A152-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 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-1 15 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 l 0 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 6

Dominion [n~ineerin~, Inc. TN-5696-00-02, Rev. 0 equations of MRP-5 5 and MRP-115. The MRP-5 5 and MRP-115 deterministic crack growth rate equations are not bounding equations, but rather reflect the 75lh 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 hour were screened out of the database for Alloys 821182/132.

The greater resistance of Alloys 690/521152 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 600/821182 and compare this against the full variability of the Alloy 690/52/152 data most applicable to plant conditions. The lack of any 7

Dominion fn~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 of MRP-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 4 recently released by NRC [ 17] , 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 4 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. 8

Dominion fn~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

Dominion fn~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 of less than l 0% (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 [l] 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 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 of MRP-3 75 [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 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 ofreplacement 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

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 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 690/521152 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 I AWS ERNiCrFe-7), Alloy 52M (UNS N06054 I AWS ERNiCrFe-7A), 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 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 11

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 avai lable 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 REFEREN CES

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, 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 l ,"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), 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-11 7), 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-110NP), EPRI, Palo Alto, CA: 2004. 1009807-NP.

[ML04 l 680506]

7. Materials Reliability Program: Probabilistic Fracture Mechanics Analysis ~lPWR 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 6901152152 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

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 Presented at: Alloy 6901152152 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. 109-1 25 , 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 NRCIEPRI Meeting, 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 En vironmental 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 6901152152 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 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 In vestigation of Stress Corrosion Cracking in Nickel-Base Alloys, NUREG/CR-7103 , Vol. 2, April 2012.

19. Ma terials 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

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 Data from Individual Heats 1.E-09

                        ~     CIEMAT I

MRP-55 Curve/1 I 1.E-10

        ~

_§.

        -QJ                                  -                                U A
                                                                                                              ~ MRP/12}--
                                   .Jfl' ro a:::                                                                                 -    -     -
        ..c:

1.E-11

                              /
         ~

0

        "'ro u

u

                               ,/

7'- -- ~ PPU data are represented with . open symbols 1.E-12 -*

                        .                                                                           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-im) 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 CIEMAT

                        - + PNNL                                                                          MRP-551 Curve/1 I ti) 1.E-10
        ],
         -QJ ro a:::
        ..c:                  /
                                   .Jfl' A
                                                            ~
                                                             %               ~ 8A-----~
         ~

1.E-11 0

                          *                     ...,_.                            A u

ro

                                                                  ~

n

                                                                                    ~

A

                                                                                      ~

PPU data are represented with f!f' ~ u 1.E-12

                               /,

• open symbols

                                                                   ~

Data are adjusted for I* temperature (325°C). A Q = 130 kJ/mol 1.E-13 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-im) Figure 2. Plot of da/dt versus K1 for Alloy 690 Data from Heat WP787 14

I I Dominion [n~ineerin~, Inc. TN-5696-00-02, Rev. 0 1.E-09

                          - o ANL
                                 + PNNL                                                                     MRP-55 Curve/1 I
        -.§.
         ~

1.E-10

                                            ~

C1l r-

         ~

ro

                                                                                                          -     ~ MRP/12
         ..c:

1.E-11 /

          ~                                        O=

0

         <.!:)

I

                             .                  - ~

J' - - - PPU data are

         .:.:.                              ~      ......

u represented with ro

           .....                      -7'-                    0 u                         /'                                                                         open symbols 1.E-12
                               .                                                                      Data are adjusted for I

temperature (325°C) . Q = 130 kJ/mol 1.E-13 y 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-Vm) Figure 3. Plot of da/dt versus K1 for Alloy 690 Data from Heat WP142 1.E-09

                          ~       OANL MRP-551 Curve/1 I 1.E-10 Vl
         ],                                  ~

C1l

                                                                                      --- - --~

ro .,;"

          ~
         ..c:

1.E-11 / 00 _o

           ~                                                                 ----

0

          <.!:)

u ro I

                                      ,   - ~
                                              -  --- -                      .., Cl                             PPU data are represented with u                        /                                                                           open symbols 1.E-12
                            .                                                                          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-Vm) Figure 4. Plot of da/dt versus K1 for Alloy 690 HAZ Data from Heat WP142 15

Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09

                        ~      CIEMAT I

MRP-551 Curve/1 I 1.E-10

        ~
        .s r
                                          ~
         ....ro CJ.)
                                    ~                                        il LJ.                             I 0:::                                                                                                  -     -,MRP/12
        ..c:

1.E-11 /

         ~

0 c.!:l PPU data are

        .:re.

u represented with ro

          ....                     -/-                                        e:,.

u / open symbols 1.E-12 I

                                                                                     -                     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 (MPav'm) Figure 5. Plot of da/dt versus K1 for Alloy 690 HAZ Data from Plate Material Tested by CIEMAT 1.E-09 _ D GE-GRC _ + PNN~MRP-1 15,..-- Curve/1 I

                                                                                                                ~*

Vi' 1.E-10

        ].
         ....ro CJ.)

0:::

        ..c:
         ~

0 1.E-11

                                       .J MRP/12 r        - --
                                                       - I Fusion line I point
                                                                                          /

Dilution ...... 1 zone point 1 "ll c.!:l I \'\ri /

                                                                                   /
        .:re.

u I '\'\. / ro u 1.E-12 I \X Data are adjusted for temperature (325°C). Q = 130 kJ/mol

• D 1.E-13 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPav'm)

Figure 6. Plot of da/dt versus K1 for Alloy 152 Data from Heat WC83F8 16

Dominion En~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09 _ o ANL D GE-GRC f;1 MRP-115 i-- Curve/1 I '-

                                                                 ~

1. E-10 + PNNL

        ~

5

                                                                  - * ~
                                                                                               ~               fA
                                                          * - - "'- '     -

• r......,

Q)

          ~
                                                       - ~ ~-.;;;,,    '\. '-'                                ;:;  {"\

a:: j MRP/12 [ 06 ~"'-.. 8

        ..r::.
         ~

1.E-11 0 t:J PPU data are

                                                       -Q-{ ,
                                                        ~-

A Black-filled ANL data present ---

        ~

u u

          ~

1. E-12 represented with open symbols growth rates during the environmental pre-crack period and should not be included.

Data are adjusted for temperature (325°C). Q = 130 kJ/mol B

• 1.E-13 *
• 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa.,,lm)

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

Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 Data Most Applicable to Plant Conditions 1.E-09 OANL

• Bettis 1 MRP-551 CIEMAT I Curve/1 I 1.E-10 1= + PNNL Vi' E ~

A---- Q,)

                                    ~

iii a::: AA - ~MRP/12r-

        ..c:
         ;:    1.E-11         /                                           A 0
                                                      )........,-. t.-- -

(,!) u

                                           .,.  -    ~
                                                     ~

A 7"'- u (l:l 1.E-12 / 0

~ ft

                           .                                                  ;                         ~          Data are adjusted for
                                                                                           .A
                                                                                           ~

T T

                         ~

temperature (325°C). 1.E-13 -- -

                                                                              *~

T - Q = 130 kJ/mol 10 15 25 30 35 40 45 5(} 55 60 Stress Intensity Factor (MPa-Vm) Figure 8. Plot of da/dt versus K1 for Alloy 690 Data from All Laboratories, :S 10% Cold Work, Constant Load or K1 o ANL Bettis CIEMAT 1--- --IU-- - --.--- - - - -,,_- - - - -- - - --1

          .§ 0.7         PNNL        1--~....__----#-------+------------1 6

s *L::;

§ VI 0.6 +-- - - -- - - - ---1-- - - - - - - - - - - - - - - ----l

          ~ 0.5 +------iD'------~~----__,..-----<                                                             The data points at 1E-13
          .~                                                                                                    were reported as "no
          ~

iii 0.4 +.--LJ-- - - - - -- - - ----i

                                                                                   .............,~~                    growth."

E a o.3 -------~-------------< Data are adjusted for temperature (325°C) and 0.2 - - -------i ~~~ stress intensity factor. 0.1 _ _ _ _ _..,,__ _ _ _ __ _ _ _ _ _ _- - - l Q =130 kJ/mol

                                 ./

1.E-12 1.E-11 1.E-10 K = 30 MPa-Vm 0.0 ..---=-~ "'+----~....L-L..Ju+-----JL----L---l-JL.-L-L.1...1..f---.i........'.::==:r:q:====r::d..l 1.E-09 1.E-08 1.E-13 Crack Growth Rate (mis) 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

Dominion En~ineerin~. Inc TN-5696-00-02, Rev. 0 1.E-09 o ANL CIEMAT 1 MRP-55 D GE-GRC I Curve/1 I 1.E-10 Vi' = + PNNL

g

       - QJ
                                          ~
                                                                                     - -- --~

( l:l 0::

       ..c:

1.E-11 / -

        ~

0 (.!)

       .::it!

u

                             ,,                ~

(l:l

         .....                            7'-             LI                    .n u                            /

1.E-12 - -

                                .                                         El*      -    ~            Data are adjusted for I
                                                                   *         ~

temperature (325°C) . Q = 130 kJ/mol 1.E-13

• a 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPav'm)

Figure 10. Plot of da/dt versus Ki for Alloy 690 HAZ Data from All Laboratories, :510% Cold Work, Constant Load or Ki 1.0 0 ,,.,, OANL 0 0.9

6. / L CIEMAT 0.8 D GE-GRC D

D

             .Q 0.7                                                                                              + PNNL 5                         6.

§ 0.6 iii D

             ~ 0.5
             *~
               >                      ***                                                        The data points at 1E-13 were reported as "no
             ~ ::I  04
                     .                                                                                     growth ."

8 E 0.3

• Data are adjusted for temperature (325°C) and 0.2 FOi = 12 stress intensity factor.

0.1 Q = 130 kJ/mol

                                         .,,.,                                                         K = 30 MPav'm 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, :5 10% Cold Work, Constant Load or Ki 19

Dominion En~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09 - OANL

                             - CIEMAT Sl MRP-115~                  -

D GE-GRC Curve/1 I 1.E-10 + PNNL Vi'

§. - -

                                                                                       ~

QJ

          <ll                                                        -er-                                                        ()

0::

        ~
                                             ...J MRP/12 f
         ~

1. E-11 0

        <.!)
                             -                                                         t:;

u /:::,.

          <ll u

1.E-12 n ~~ _, *

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

                                                                               . ,,~.                                    T Q = 130 kJ/mol                 .!_1'.J
                                                                                   *-                -i             **

1.E-13 ~ ~--T

                                                                                                  -"-    ~*

T 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa.Ym) Figure 12. Plot of da/dt versus Ki for Alloy 52/152 Data from All Laboratories, ~ 10% Cold Work, Constant Load or Ki 1.0 -r;:::===,------------------rr---:--=----~-----, OANL 0 efJ o / 0.9 ~ CIEMAT 0.8 o GE-GRC 1 - - - - 1 r " f - - - - - - - - - --1-- - - - -....--...._..---;

                           + PNNL
            § 0.7      t------mr----------~f-------1..l!:.~:....!..1.J

J

g 0.6

           -VI
                       +-- - --#li>-- - - - - - - - - - --#-- - - - - ----#- - ---1
          ~ o.5 +-----~~------------1-------1                                                                           The data points at 1E-13
          .5                                                                                                                were reported as "no
          -3      0.4  +--~11-------------+------1                                                                                  growth ."

E  ::=====z=.========~ 8 0.3 +----illP= - - - - - - - - - - - --1-- - -- - l Data are adjusted for FOi = 12 temperature (325°C) and 0.2 +!Alll!--- - - - - - - - - - - - - == ==-- ---1 stress intensity factor. 0.1 c::J-- - - - - - - - - - - - . . . . - - - - - - - 1 I Q = 130 kJ/mol K = 30 MPa.Ym 0.0 .__....._.........................""""'"+--"'_.__,,::::::..................1--___,--~:.::::Wu::+====::;::=:::L=::;:::::;::;:~ 1.E-13 1.E-12 1.E-1 1 1.E-10 1.E-09 Crack Growth Rate (mis) 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

Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 Comparison of Partial Period Unloading (PPU) Conditions vs. Constant Load Conditions 1.E-09 Data are adjusted for  := Specimen temperature (325°C) (Q = 130 -

                                                                                                            -    9T1 kJ/mol) and K (30 MPa--/m) -

1.E-10 IPPU Data f- - - - - - - - - - - - - -, 9T2 Const. Load Data ~ Vi' ~ 9T3 E

        -QJ                  -~

I I I I 9T4

        ~
         <ti e:::

1.E-11 I ,. ~ I I o. I I 9T5 9T6 3:0 (.!) - I

                                                                                    ~
                                                                                                  -              9T8
                                                                 ~

I

        ~
                                                   --o-                                                          9T9 u              I      /
                                                            ~

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         <ti u

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C\~

                                                        ~

I

                                                                                   \ ',/~              I I                                    I I

I

                                                                                 -----~-*

I I 1.E-13 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 Ki/Load Tes ting) from Heat WP787 21

Dominion En~ineerin~, Inc. TN-5696-00-02, Rev. 0 Compilation of ANL and PNNL Data 1.E-09 Box and arrow show the ratio between the MRP-55 1 MRP-55 curve and the data point I Curve/1 I 1.E-10

                                                                                  ~                           96 Vi'

§. 10 2 9.3 El *

        -Q)

( ll 0::: /I

                                          ..r
                                               ~

I 1110.2 I I II 9.0 l:l

                                                                ~ ~ ~ L/
                                                                                    /
                                                                                                       /

LI I 11 5 1 1 1 MRP11 2r

        ..i:::

1.E-11 10.4 -

         ~                             -I 0
         ....                                                                                         ....                                          o ANL CL
                              ,                  -~-

(.!)

        .:it:.

u u (ll

                                    /

7"'- - - - v 0

                                                                              ~                       *
                                                                                                                        .T
                                                                                                                          ....                      OANL PPU

• PNNL 1.E-12

                              .                                         T Data are adjusted for temperature (325°C) .

Q = 130 kJ/mol 1.E-13 T - TTT T T 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa'1m) 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 1.0 ,-------------,.---- ,,.,..-::-;..-~----::;;:;;-------, 0.9 +------------,.-...~------------------l 0.8 --------~~-~------.f------------1

           .g      0.7 +------~------------1------------                                                                                            0-A-NL_C_L--.-1

l

§ 0.6 +-----...~----1-------1---------l + PNNL Vi

            ~ 0.5                                                                                                                The data points at 1E-13 are
            .5                                                                                                                       treated as "no growth ,"
            -3     0.4                                                                                                             consistent with MRP-375.

E 8 0.3 Data are adjusted for IFOi = 12 I temperature (325°C) and 0.2 &------=;==~---+-------l stress intensity factor. 0.1 ----~-------#---------. Q=130kJ/mol

                                       ./                                                                                                K = 30 MPa'1m 0.0 .__.....::.~........'""'+---~::;_._..u....y.._ _.___.__.__._.__.__._._1--__.__::c==:i:+=====~

1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 Crack Growth Rate (mis) 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/Ki 22

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 1.E-09 o ANL CL Box and arrow show the OANL PPU ratio between the MRP-55

                                + PNNL                curve and the data point                                            I MRP-55 r I Curve/1 I 1.E-10 Cil
        ].
        -OJ
                                                   ~
                                         ,,r
         ~

a::: _ - ~MRP/12r

        ..c:                     /                     I 9.3 k        oO~

3:0 1.E-11

        ~                   i I                       .,_. -                       ...., {"\
        ~

u u

         ~

1.E-12 / T

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

Q = 130 kJ/mol 1.E-13 . T T I 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa..Jm) Figure 17. Plot of da/dt versus Ki for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17]; S 22% Cold Work 1.0 -r----------------:---~~---::::::--------, ,,.., 0.9 -+--- - - - - - - - ---r-,-- -#-- - - - - -- - - - - - - - ----< 0.8 *

                       +-----------~----___,,__                                                            _ _ _ _ _ _ _ _ __,
            .g 0.7 +----..~------+--------1---------r-o-A_N_L_CL-....1

i *

§ 0.6 +--- - - - - - - --1-- - - - -- -- - - - - - ---1 + PNNL Vi *

            ~ 0.5                                                                                            The data points at 1E-13 are
            *5                                                                                                   treated as "no growth ,"
            -3     0.4                                                                                          consistent with MRP-375.

E 8 0.3 Data are adjusted for IFOi = 12 I temperature (325°C) and 0*2 ~------=;==:!.__ ___ ..J-_ _ _ _ ___j stress intensity factor. 0.1 _ _ _ ____,,__----~~------; Q = 130 kJ/mol

                                      ./                                                                              K = 30 MPavfm 0.0 _...-=:c...:............._ _ _~::._,_,L...L..i.+--'--'--L......L...1....L..LI.+----'--__:r::==:::q::::====::rl.I 1.E-13                      1.E-12              1.E-11                         1.E-10             1.E-09             1.E-08 Crack Growth Rate (mis)

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

Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09 Box and arrow show the - I MRP-115 i-::: - ratio between the MRP-115 1.E-10 __....--1 Curve/1 I I 5.5 LJ 9.2 I curve and the data point

        ~                                                             -o ..._

_§_ -r' ------~ Q)

                                                                               - - . J "\,J
                                                                                 ~
                                                                                                    -                    l l
                                                             -~
                                              .J MRP/12 r                               -
          ~

O:'.

        ..c:
         ~

1.E-11 0 (.!) o ANL CL u OANL PPU u

          ~

1.E-12

                               + PNNL CK (NUREG)                                   *

• Data are adjusted for .... - .... - ~ -

1.E-13 temperature (325°C). Q =130 kJ/mol * **

                                                                            ~
                                                                            ~
                                                                                    ~
                                                                                    ¥
                                                                                               ~
                                                                                               ¥
                                                                                                                  ~
                                                                                                                  ¥ 10            15          20      25        30              35              40             45           50        55       60 Stress Intensity Factor (MPa-1m)

Figure 19. Plot of da/dt versus K1 for Alloy 52/152 Data Produced by ANL and PNNL and Available in References [17] and [18]; :S 22% Cold Work 1.0 o ANL CL 0.9

• PNNL CK (NUREG) 0 70 --

                                                                                                                                                 /
                                                                                                           .I
                                                                                                     ~

0.8 0 0 I I

            § 0.7                                    *

• I IMRP-115 L IFOI = 1)

          *z

J

• I I

                                                                                                                                      ./

§ 0.6

(;) I I

          ~ 0.5                            ~
                                                                                            ,I                          The data points at 1E-13 T
          *z>                                                                                                              were reported as "no
            ~ 04 I                                       growth ."

i .

8 E 0.3 I Data are adjusted for 0.2 ....* IFOI = 121 temperature (325°C) and stress intensity factor. 0.1 I Q = 130 kJ/mol 0.0

                      *t
                       ~

_, / K = 30 MPa..Jm 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (mis) Figure 20. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data Produced by ANL and PNNL ([17] and [18]); :S 22% Cold Work and Constant Load/Ki 24

Dominion fn~ineerin~, Inc TN-5696-00-02, Rev. 0 Data for Less than 20% Cold Work from All Laboratories 1.E-09 rv~c=i~~~~~~~~~~~~~~~~~ CIEMAT MRP-55 ~~::1

                                           ~~~~iiiiii~Cui~e/~1~

1.E-13 -+--'-~-'--+-~~+-'-~-'--+-~~+-'-~-'--+-~~._.._~._._~~._.._~._._~...__.__. 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-Vm) Figure 21. Plot of da/dt versus Ki for Alloy 690 Data from All Laboratories, > 10 & ~ 20% Cold Work, CROM and Bar Material, Constant Load or Ki Testing v AMEC

                         ~ CIEMAT DGE-GRc i---------<;,__,.--- - - --+- - - - - - - - ----1
           § 0.7         + PNNL
         *.=;

J

§ 0.6 -+--- - - - - -__.,.,._ _ _ _ _ __,,____ _ _ _ _ _ _ _ _ ___,

Vi

         ~ o.5 1-----x-r------;:::::===+/-::::.~---i The data points at 1E-13
         *.=;                                                                                             were reported as "no
         -3      0.4                                                                                             growth."

8E 0.3 +----_.,.__ _...___ _ _ _~'--------l Data are adjusted for 0.2 -+--- -- I---<~~~,___ _ _ _, __ _ _ _ _---< temperature (325°C) and stress intensity factor. 0.1 Q =130 kJ/mol K= 30 MPa./m 0.0 ~-=-.................~--~:.................L...l..f---'----'"------'-..J.-'--'-'-!-----'-__::r:::==::q=====~ 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 Ki 25

Dominion [n~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-09 _ DGE-GRC

                         -                  ~ MRP-115 p                 -

Curve/1 I

                         - + PNNL 1.E-10 Ii)
        ],                                                                                          -~r1 a.>
         -ro c:::
        ..c::

1.E-11 ~

         ~

0 (.!)

        ~

u IQ El u 1.E-1 2 - D

• 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-Vm)

Figure 23. Plot of da/dt versus Ki for Alloy 52/152 Data from All Laboratories, > 10 & S 20% Cold Work, Constant Load or Ki 1.0 -,;:::===:;-----------------n-- --=--~-----, OANL /" 0.9 ~ CIEMAT 1----------,--7'rtJ-~----.---------..~--i 0.8 D GE-GRC 1-----=!"l'~---------1-------,...,...,..,,,,.,,,......L...,,-,r--~

                        + PNNL
           § 0.7 +========-----;:t;E1=--- - - - - - - ---:if-- - - - -Utlli.::Jl.J
          *~

I

§ 0.6 +------:ai~------------1--------+----l VJ

          ~ 0.5     -t--- - - - .,.,_- - - - - - - - - -__,_- -_,                                              The data points at 1E-13
          ._E;                                                                                                   were reported as "no
          -S   0.4 +--     ---.r:--- - - - - - - - - - - ---1                                                        growth ."

E ~=====~=========: 8 0.3 T-...i-1-'-- - - - - - - - ---.==t=-- - - - - i Data are adjusted for FOi = 12 temperature (325°C) and 0.2 +r:11F-------------~------1 1 stress intensity factor. 0.1 t-t------------~------t Q = 130 kJ/mol K = 30 MPav'm 0.0 .__......_.............................._ _.._._;_,,,,,.~..................~_.--~:i::W:::;:+:==:::;::~:::;:::::;::::;::w:::;'.,j 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (mis) Figure 24. Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data from All Laboratories, S 20% Cold Work, Constant Load or Ki 26}}