LR-N15-0036, Request for Relief from Alloy 690 PWR Reactor Vessel Head Inspection Interval, Fourth (4th) 10-Year Interval

Request for Relief from Alloy 690 PWR Reactor Vessel Head Inspection Interval, Fourth (4th) 10-Year Interval

ML15098A426
Person / Time
Site:	Salem
Issue date:	04/08/2015
From:	Duke P Public Service Enterprise Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LR-N15-0036, S1-I4R-150
Download: ML15098A426 (47)

Text

PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 OPSEG NuclearLLC 10 CFR 50.55a LR-N 15-0036 April 8, 2015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Generating Station, Unit 1 Renewed Facility Operating License No. DPR-70 NRC Docket No. 50-272

Subject:

Request for Relief from Alloy 690 PWR Reactor Vessel Head Inspection Interval, Fourth (4th) 1O-Year Interval In accordance with 10 CFR 50.55a(z), "Codes and standards," PSEG Nuclear LLC (PSEG),

hereby requests NRC approval of proposed Relief Request S1-14R-150 for Salem Unit 1.

PSEG is requesting relief from the reactor vessel closure head requirements of ASME Code Case N-729-1. The proposed alternative would extend the volumetric/surface examination interval for the Salem Unit 1 Alloy 690, 52, and 152 reactor vessel closure head nozzles and partial-penetration welds for approximately 5 years.

PSEG requests approval of the proposed request by April 8, 2016, prior to the next Salem Unit 1 refueling outage (1 R24). Relief Request S1-14R-150 applies to the Salem Unit 1 fourth interval which began on May 20, 2011 and is scheduled to end on May 20, 2021.

The proposed relief request is provided in Attachment 1. Attachment 2 provides further support for the requested alternative inspection interval.

There are no regulatory commitments contained in this letter.

If you have any questions or require additional information, please contact me at 856-339-1466.

Respectfully,

~~=-'? oJ. ~. 0 k __ .A Paul R. Duke, Jr. :_~ - - \

Manager - Licensing

April 8, 2015 10 CFR 50.55a Page 2 LR-N15-0036 Attachments:

1. 10 CFR 50.55a Relief Request S1-I4R-150

2. 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, prepared by Dominion Engineering, Inc.

cc: Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Sanders-Parker, Project Manager, NRC NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Manager IV, NJBNE PSEG Corporate Commitment Tracking Coordinator Salem Commitment Tracking Coordinator

LR-N15-0036 Attachment 1 10 CFR 50.55a Relief Request S1-I4R-150

LR-N15-0036 Salem Nuclear Generating Station, Unit No. 1 Renewed Facility Operating License No. DPR-70 NRC Docket No. 50-272 Relief Request - S1-I4R-150 Proposed Alternative in Accordance with 10 CFR 50.55a(z)(1)

Acceptable Level of Quality and Safety

1. ASME Code Component(s) Affected Code Class: 1 Examination Category: Code Case N-729-1 Item Number: B4.40

Description:

ASME Class 1 Pressurized Water Reactor (PWR) Reactor Vessel Upper Head (Closure Head) (RVCH) nozzles and partial-penetration welds fabricated with primary water stress corrosion cracking (PWSCC)-resistant materials.

Unit/Inspection: Salem Unit 1 / Fourth (4th) 10-Year Interval

2. Applicable Code Edition and Addenda

The Salem Unit 1 Inservice Inspection (ISI) Interval Code of record is the 2004 Edition of ASME Boiler and Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Examinations of the reactor vessel closure head (RVCH) penetrations are performed in accordance with 10 CFR 50.55a(g)(6)(ii)(D), which specifies the use of Code Case N-729-1, with conditions. Salem Unit 1 Fourth (4th) ISI interval began on May 20, 2011 and is scheduled to end on May 20, 2021.

3. Applicable Code Requirement

The Code of Federal Regulations (CFR) 10 CFR 50.55a(g)(6)(ii)(D)(1), requires (in part):

All licensees of pressurized water reactors must 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.

10 CFR 50.55a(g)(6)(ii)(D)(3) conditions ASME Code Case N-729-1 (Reference 1) by stating:

Instead of the specified examination method requirements for volumetric and surface examinations in Note 6 of Table 1 of Code Case N-729-1, the licensee must perform volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Figure 2 of ASME Code Case N-729-1. A demonstrated volumetric or surface leak path assessment through all J-groove welds must be performed. If a surface examination is being substituted for a volumetric examination on a portion of a penetration nozzle that is below the toe of the J-groove weld [Point E on Figure 2 of ASME Code Case N-729-1], the surface examination must be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically.

1

LR-N15-0036 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 the code case. The basic inspection requirements of ASME Code Case N-729-1 for partial-penetration welded Alloy 690 head penetration nozzles are as follows:

Item B4.30 - Direct visual examination (VE) of the entire outer bare metal surface of head for evidence of leakage every third refueling outage or five (5) calendar years, whichever is less.

Item B4.40 - Volumetric and/or surface examination of all nozzles every ASME Section XI 10-year ISI interval, nominally 10 calendar years (provided that flaws attributed to primary water stress corrosion cracking have not previously been identified).

4. Reason for Request

Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D) requires volumetric and/or surface examination of the RVCH penetration nozzles and associated welds no later than nominally 10 calendar years after the head was placed into service. This examination schedule was intended to be conservative and subject to reassessment once additional laboratory data and plant experience on the performance of Alloy 690 and Alloy 52/152 weld metals became available (Reference 2 and Reference 3). Using plant and laboratory data that has since become available, Electric Power Research Institute (EPRI) document Materials Reliability Program (MRP)-375 (Reference 3) was developed to support a technically based volumetric or surface reexamination interval using appropriate analytical tools. This technical basis demonstrates that the reexamination interval can be extended to at least a 20 year interval while maintaining an acceptable level of quality and safety.

Therefore, PSEG is requesting approval of this alternative to allow the use of the ISI examination interval extension of approximately 5 years for the Salem Unit 1 Alloy 690/52/152 reactor vessel closure head penetrations.

In addition to maintaining an acceptable level of quality and safety with this proposed one time 5 year volumetric and/or surface examination frequency extension, Salem Unit 1 will continue to schedule one RVCH visual examination on an interval of every third refueling outage or 5 calendar years, whichever is less, in accordance with Item B4.30 of ASME Code Case N-729-1.

5. Proposed Alternative and Basis for Use The proposed alternative is the one time 5 year volumetric and/or surface examination frequency extension, Salem Unit 1 will continue to schedule one RVCH visual examination on an interval of every third refueling outage or 5 calendar years, whichever is less, in accordance with Item B4.30 of ASME Code Case N-729-1.

The original Salem Unit 1 RVCH, which was manufactured with Alloy 600/82/182 materials, was replaced with a new RVCH, using Alloy 690/52/152 materials, during the refueling outage preceding return to operation in November 2005. In accordance with Table 1 of ASME Code Case N-729-1, Item B4.40, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D)(3),

PSEG will be required to perform a volumetric and/or surface examination of essentially 2

LR-N15-0036 100% of the required volume or equivalent surfaces of the nozzle tube on a nominally 10 year frequency, currently scheduled for spring 2016.

The inspection frequency of ASME Code Case N-729-1 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 4), which is summarized in the safety assessment for RVCHs in MRP-110 (Reference 5). The material improvement factor for primary water stress corrosion cracking (PWSCC) of Alloy 690 and Alloy 52/152 materials over that of mill-annealed Alloys 600 and 182 is shown in MRP-110 to be on the order of 26 or greater.

Further evaluations were performed to demonstrate the resistance of Alloys 690/52/152 to PWSCC under a recent EPRI MRP initiative provided in MRP-375 (Reference 3). This report combines an assessment of the test data and operating experience developed since the technical basis for the 10-year interval of ASME Code Case N-729-1 was developed in 2004 (Reference 2) with deterministic and probabilistic evaluations to assess the improved PWSCC resistance of Alloys 690/52/152 relative to Alloys 600/82/182.

Evaluation of Existing Alloy 690/52/152 Data and Experience by MRP-375 Operating experience, to date, for replacement and repaired components using Alloys 690/52/152 has shown a proven record of resistance to PWSCC during numerous examinations in the approximately 25 years of use in the nuclear industry. This includes steam generators, pressurizers, and RVCHs. In particular, at the completion of the spring 2014 refueling outage season, Alloy 690/52/152 operating experience includes inservice volumetric or surface examinations performed in accordance with ASME Code Case N-729-1 on 13 of the 40 replacement RVCHs currently operating in the U.S. None of these examinations revealed PWSCC cracking, and these examination results further support the low likelihood of the potential for the RVCH to experience PWSCC during the extension period.

One of the replacement heads that was volumetrically examined in accordance with ASME Code Case N-729-1 Item B4.40 was Salem Unit 2. Both Salem Unit 1 and Unit 2 heads were fabricated by the same manufacturer (AREVA) using Alloy 690 nozzle material specifications (SB-167, USN N06690, Thermally Treated Condition), with the design of the Salem heads being essentially identical. As stated above, none of the prior examinations of replacement RVCHs with Alloy 690 nozzles have revealed any indications of PWSCC or service-induced cracking.

In France in 2013, a second 10 year non-destructive examination inspection was performed on one of the first replacement RV heads with Alloy 690/52/152 material (Reference 6).

There were no reports of PWSCC having been detected after approximately 20 years of service.

The evaluation performed in MRP-375 considers a simple Factor of Improvement (FOI) approach applied in a conservative manner to model the increased resistance of Alloys 690 and 52/152 compared to Alloys 600 and 182 at equivalent temperature and stress conditions.1 FOIs were estimated for the material improvements of Alloy 690/52/152 1

Alloy 600 wrought material is the appropriate reference for defining the FOI for Alloy 690 wrought material. As discussed in Section 3.1 of MRP-375, Alloy 182 weld metal is chosen as the reference for defining the FOI for Alloys 3

LR-N15-0036 materials using an extensive database of test data. Results for both crack initiation and crack growth show a substantially improved resistance to PWSCC for Alloy 690 base material and Alloy 52/152 weld materials. Figures 3-2, 3-4, and 3-6 of MRP-375, 2 provide crack growth rate data for Alloy 690/52/152 materials and heat affected zones with curves plotting FOIs of 1, 5, 10, and 20 on a statistical basis that reflects the material variability exhibited in MRP-55 (Reference 7) for Alloy 600 material and in MRP-115 (Reference 8) for Alloy 82/182/132 weld material.3 An FOI of 20 bounds over 70% of the data plotted in each of the three figures, and an FOI of 12 bounds all of the crack growth rate data.4 A comparison between the cumulative distributions of the crack growth rates for Alloys 690/52/152 and Alloys 600/82/182 treats the full variability in both original and replacement alloys. Figures 3-1, 3-3, and 3-5 of MRP-375 compare the full variability of the replacement alloy against a conservative mean growth rate (75th percentile of the heat-to-heat variability) for the original alloy, and thus do not illustrate the full range of material behavior for both the original and replacement alloys. Table 3-6 of MRP-375 provides a summary of FOIs determined on the basis of crack growth rate and crack initiation data. For crack initiation, FOIs reported, although significant, are conservatively small because crack initiation of Alloys 690/52/152 was not observed during testing; instead, the initiation time was assumed to be equivalent to the test duration. Conservatively, credit was not taken for the improved resistance of Alloys 690/52/152 to PWSCC initiation in the main MRP-375 analyses.

Additional Evaluations Performed under MRP-375 MRP-375 applied the FOI results to perform a combination of deterministic and probabilistic evaluations to establish an appropriately conservative inspection interval for Alloy 690 RVCHs. The deterministic technical basis applies industry-standard crack growth calculation procedures to predict time to certain adverse conditions under various conservative assumptions. A probabilistic evaluation is then applied to make predictions for leakage and ejection risk, generally using best-estimate inputs and assumptions, with uncertainties treated using statistical distributions.

The deterministic crack growth evaluation provides a precursor to the probabilistic evaluation to directly illustrate the relationship between the improved PWSCC growth resistance of Alloys 690/52/152 and the time to certain adverse conditions. These evaluations apply conservative crack growth rate predictions and the assumption of an existing flaw (which is replaced with a PWSCC initiation model for probabilistic evaluation).

The evaluations provide a reasonable lower bound on the time to adverse conditions, from 52 and 152 weld metals because Alloy 182 is more susceptible on average to PWSCC initiation and growth than Alloy 82 (due to the higher Cr content of Alloy 82).

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

3 As discussed in Section 3.3 of MRP-375, the laboratory crack growth rate data compiled in MRP-375 represent the values reported by individual researchers, without any adjustment by the authors of MRP-375 other than for temperature and stress intensity factor. The data presented in Figures 3-2, 3-4, and 3-6 of MRP-375 represent essentially the entire set of data points reported by the various laboratories. No screening process was applied to the data on the basis of test characteristics such as minimum required crack extension or minimum required engagement 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% added cold work.

4 One of the weld data points in Figure 3-6 of MPR-375 appears to be within an FOI of 12, but this data point and three others not within an FOI of 12 are irrelevant because they reflect fatigue pre-cracking conditions. The nature of these four points as fatigue pre-cracking segments was clarified subsequent to the publication of MRP-375.

4

LR-N15-0036 which a conservative inspection interval may be recommended. This evaluation draws from various EPRI MRP and industry documents that evaluate, for Alloys 600/82/182, the time from a detectable flaw being created to leakage occurring and from a leaking flaw to the time that net section collapse (nozzle ejection) would be predicted to occur. Applying a conservative crack growth FOI of 20 to circumferential and internal diameter (ID) axial cracking and of 10 to outer diameter (OD) axial cracking for Alloy 690 versus Alloy 600, the results show that more than 20 years is required for leakage to occur and that more than 120 years would be required to reach the critical crack size subsequent to leakage.

The probabilistic model in MRP-375 was developed to predict PWSCC degradation and its associated risks in RVCHs. The model utilized in this probabilistic evaluation is modified from the model presented in Appendix B of MRP-335, Rev. 1 (Reference 9) that evaluated surface stress improvement of RVCHs with Alloy 600 nozzles. The integrated probabilistic model in MRP-375 includes submodels for simulating component and crack stress conditions, PWSCC initiation, PWSCC growth, and flaw examination. The submodels for crack initiation and growth prediction for Alloy 600 reactor pressure vessel head penetration nozzles (RPVHPNs) in MRP-335, Rev. 1 were adapted for RVCHs with Alloy 690 nozzles by applying FOIs to account for the superior PWSCC resistance of Alloys 690/52/152. The average leakage frequency and average ejection frequency were determined using the Monte Carlo simulation model with conservative FOI assumptions. The results show that, using only modest FOIs for Alloys 690/52/152, the potential for developing a safety significant flaw (risk of nozzle ejection) is acceptably small for a volumetric or surface examination period up to 40 years.

The evaluations performed in MRP-375 were prepared to bound all pressurized water reactor (PWR) replacement RVCH designs that are manufactured using Alloy 690 base material and Alloy 52/152 weld materials. The evaluations assume a bounding continuously operating RVCH temperature of 613°F and a relatively large number of RVCH penetrations (89). This number bounds the number present in most replacement heads (including those at Salem Unit 1), but some heads are known to have a modestly larger number of nozzles (e.g., 15% more). The number of penetrations included in the probabilistic model is not a key variable, and the assumed number of penetrations results in a small change in results relative to other sensitivity cases. Thus, the probabilistic calculations of MRP-375 cover all U.S. replacement RVCHs regardless of the precise number of penetrations.

While approval of this request for alternative is not contingent on NRC review and approval of MRP-375, the insights gained in this technical report help substantiate the limited extension duration being requested. In particular, the tabulation of crack growth rate data for Alloys 690/52/152 (Section 3 of MRP-375) and review of inspection experience for Alloy 690/52/152 plant components (Section 2 of MRP-375) are sufficient to demonstrate the acceptability of the limited extension duration being requested. This request is not dependent on the more detailed probabilistic calculations presented in Section 4 of MRP-375.

RVCH Design and Operation The analysis presented in MRP-375 was intended to cover all replacement heads in U.S.

PWRs, including the Salem Unit 1 RVCH. The MRP-375 analyses assume a reactor vessel head operating temperature of 613°F to bound the known reactor vessel head temperatures of all U.S. PWRs currently operating. The RVCH operating temperature for Salem Unit 1 over the operating period from installation of the replacement head until the end of the 5

LR-N15-0036 requested volumetric or surface inspection period is 597.2°F (Reference 10). Thus, the Salem Unit 1 RVCH operating temperature is bounded by the MRP-375 evaluation, which assumes 613°F for the main deterministic and probabilistic calculations.

The Salem Unit 1 RVCH was designed and fabricated using materials and techniques to reduce susceptibility to PWSCC and facilitate prompt detection of potential leakage by visual examination. The RVCH contains fifty-five (55) nozzle penetrations of which fifty-three (53) are used for control rod drive mechanisms (CRDMs), one (1) is used for reactor vessel level instrumentation (RVLIS), and one (1) is a small-diameter vent line penetration near the center of the RVCH. The replacement RVCH was manufactured by Framatome (AREVA) and placed in service in November 2005. The replacement RVCH was manufactured as a single forging, which eliminated several welds including the flange to head weld. The replacement RVCH is fabricated from SA-508, Grade 3, Class 1 low alloy steel and clad with an initial layer of 309 L stainless steel followed by subsequent layers of 308 L stainless steel. The nozzle housing penetrations on the replacement RVCH are fabricated from Inconel SB-167 (Alloy 690) UNS N06690 supplied by Valinox. The nozzle J-groove welds utilized ERNiCrFe-7 (UNS N06052) and/or ENiCrFe-7 (UNS W86152) weld materials.

A pre-service volumetric examination of the Salem Unit 1 replacement RVCH partial-penetration welded nozzles was performed prior to installation. The volumetric examinations included scanning the nozzles to the fullest extent possible, from the end of the nozzle to a minimum of 2 inches above the root of the J-groove weld on the uphill side.

There were no recordable indications identified during the pre-service volumetric examinations of the nozzle tube in the area of the J-groove welds.

Bare metal visual examinations (VE) were performed on the Salem Unit 1 replacement RVCH in 2010 and 2013 in accordance with ASME Code Case N-729-1, Table 1, Item B4.30. The 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. These examinations will be performed again in the upcoming Salem 1R25 refueling outage scheduled to commence in the fall of 2017. During every refueling outage, a separate plant walkdown is performed to visually detect evidence of leakage of plant componentsincluding leakage from the region of the RVCH. For surfaces not obscured by insulation, leakage is visually apparent due to streaking and the precipitation of white crystals of boric acid on the dark surface of reactor components.

All U.S. plants, including Salem Unit 1, have an Alloy 600 management plan for managing PWSCC of reactor components and a boric acid corrosion control (BACC) program for minimizing the potential for consequential corrosion of reactor components. The NEI 03-08 Materials Initiative of the Nuclear Energy Institute (NEI), EPRI MRP-126, Generic Guidance for Alloy 600 Management, requires every U.S. PWR to have an Alloy 600 management program. NRC Generic Letter 88-05 requires that utilities develop and implement programs to identify leaks and take corrective action to prevent recurrence. In addition, under NEI 03-08, the industry document WCAP-15988-NP Rev. 2, Generic Guidance for an Effective Boric Acid Inspection Program for Pressurized Water Reactors, requires that every U.S. PWR have a BACC program that addresses boric acid corrosion due to borated water from any plant system, including those outside of containment. Evidence of leakage during a plant walkdown could indicate the occurrence of PWSCC and is tracked by plant personnel in light of the industry guidance.

6

LR-N15-0036 Minimum FOI Implied by Requested Inspection Period ASME Code Case N-729-1 is based upon conclusions reached (Reference 11) that a reexamination interval between volumetric or surface examinations of one 24-month operating cycle is acceptable for a head with Alloy 600 nozzles and operating at a temperature of 605°F. The inspection period for heads with Alloy 690 nozzles in ASME Code Case N-729-1 is a nominal 10 years, which represents a minimum implied factor of improvement (FOI) of 5 over Alloy 600.

FOI Approach Per the technical basis documents for ASME Code Case N-729-1 for heads with Alloy 600 nozzles (Reference 5, Reference 11, and Reference 12), the effect of differences in operating temperature on the required volumetric or surface reexamination interval for heads with Alloy 600 nozzles can be easily addressed on the basis of the Re-Inspection Years (RIY) parameter. The RIY parameter adjusts the effective full power years (EFPYs) of operation between inspections for the effect of head operating temperature using the thermal activation energy appropriate to PWSCC crack growth. For heads with Alloy 600 nozzles, ASME Code Case N-729-1 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D)(2) limits the interval between subsequent volumetric or surface inspections to RIY = 2.25. The RIY parameter, which is referenced to a head temperature of 600°F, limits the time available for potential crack growth between inspections.

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

(1) Results from Argonne National Laboratory (ANL) indicate that Alloy 690 with 0-26%

cold work has an activation energy between 100 and 165 kJ/mol (24-39 kcal/mol)

(Reference 13). NUREG/CR-7137 (Reference 13) concludes that the activation energy for Alloy 690 is comparable to the standard value for Alloy 600 (130 kJ/mol).

(2) Testing at Pacific Northwest National Laboratory (PNNL) found an activation energy of about 120 kJ/mol (28.7 kcal/mol) for Alloy 690 materials with 17-31% cold work (Reference 14).

(3) Additional PNNL testing determined an activation energy of 123 kJ/mol (29.4 kcal/mol) for Alloy 690 with 31% cold work (Reference 15).

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

Assuming a lower activation energy would actually reduce the bounding minimum FOI required by a U.S. PWR to justify an extended interval.

As discussed in the MRP-117 (Reference 11) technical basis document for heads with Alloy 600 nozzles, effective time for crack growth is the principal basis for setting the appropriate reexamination interval to detect any PWSCC in a timely fashion. U.S. PWR inspection 7

LR-N15-0036 experience for heads with Alloy 600 nozzles has confirmed that the RIY = 2.25 interval results in a suitably conservative inspection program. There have been no reports of nozzle leakage or of safety-significant circumferential cracking in instances when the Alloy 600 nozzles in a head were first examined by non-visual inservice non-destructive examination (Reference 16 and Reference 17) when the RIY = 2.25 re-inspection interval is followed.

Minimum FOI Implied by Requested Inspection Period PSEG has assessed the minimum Alloy 690/52/152 FOI that supports the requested Salem Unit 1 extension period for comparison with the laboratory crack growth rate data presented in MRP-375. An extension of the examination interval to 15 years would imply a factor of 7.5 for Alloys 690/52/152 relative to Alloys 600 and 182 for the proposed period between volumetric or surface examinations for a head operated at a temperature of 605°F. To calculate the minimum implied FOI for the RVCH operating temperature of 597°F, the RIY parameter for the requested examination interval is compared with the ASME Code Case N-729-1 interval for Alloy 600 nozzles of RIY = 2.25.

The representative Salem Unit 1 RVCH operating temperature of 597°F corresponds to an RIY temperature adjustment factor of 0.932 (versus the reference temperature of 600°F) using the activation energy of 31 kcal/mol (130 kJ/mol) for crack growth from ASME Code Case N-729-1. As discussed previously, it is appropriate to apply this standard activation energy for modeling crack growth of Alloy 690/52/152 plant components. Conservatively assuming that the EFPYs of operation accumulated at Salem Unit 1 since RVCH replacement is equal to the calendar years since replacement, the RIY for the requested extended interval of (0.932)(15) = 13.98. The FOI implied by this RIY value for Salem Unit 1 is (13.98)/(2.25) = 6.21.

Considering the statistical compilation of data provided in Figures 3-2, 3-4, and 3-6 of EPRI MRP-375, this factor of improvement is conservatively less than the FOI of 10 that bounds the crack growth rate data presented. Furthermore, as discussed in Section 2 and Section 3 of EPRI MRP-375, PWR plant experience and laboratory testing have demonstrated a large improvement in resistance to PWSCC initiation of Alloys 690/52/152 in comparison to that for Alloys 600/82/182. Hence, the demonstrated improvements in PWSCC initiation and growth confirm the acceptability of the limited requested period of extension on a conservative basis.

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 7, Precedents). Attachment 2 describes the materials tested for data points within a factor of 12 below the MRP-55 (Reference 7) and MRP-115 (Reference 8) crack growth rate curves for the 75th percentile of material variability. It shows that essentially all the Alloy 690 and the Alloy 52/152 data for constant load or constant K conditions are bounded by a factor of 6.2 below the 75th percentile curves in MRP-55 and in MRP-115, respectively.

Furthermore, all the Alloy 690/52/152 data are bounded by a factor greater than 6.2 relative to the MRP-55 and MRP-115 crack growth rates on a statistical basis that accounts for material variability in the original Alloy 600/82/182 materials, as well as in the replacement Alloy 690/52/152 materials. Moreover, per Attachment 2, the crack growth rate data do not show any susceptibility concerns specific to the nozzle or weld materials of the Salem Unit 1 replacement head.

8

LR-N15-0036 Conclusions The Alloy 690 nozzle base materials and Alloy 52/152 weld materials used in the Salem Unit 1 replacement RVCH provide for a clearly superior reactor coolant system pressure boundary for which the potential for PWSCC has been shown to be minute by analysis and by years of positive industry experience. This is further supported by direct visual examinations that have been performed on the Salem Unit 1 replacement RVCH and the lack of PWSCC detected in the spring 2014 volumetric examination of the Alloy 690 nozzles on the Salem Unit 2 RVCH which was fabricated by the same manufacturer using the same nozzle material.

The 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 on a statistical basis by the laboratory data compiled in EPRI MRP-375. 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 limited requested period of extension. Therefore, the Salem Unit 1 RVCH FOI corresponding to the requested extension in the volumetric/surface examination interval provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(z)(1).

6. Duration of Proposed Alternative The proposed Alternative is requested for the duration up to and including the Salem Unit 1 refueling outage 1R27 that is scheduled to commence in the fall of 2020 and that will occur in the Fourth (4th) ten-year ISI inspection interval which began on May 20, 2011 and is scheduled to end on May 20, 2021.

7. Precedents Multiple plants requested an alternative from the frequency of ASME Code Case N-729-1 for volumetric or surface examinations of heads with Alloy 690 nozzles as listed below:

NRC ADAMS Accession No.

Plant Request for Status Relief Request Additional RAI Response Information (RAI)

Arkansas Approved Nuclear One, ML14118A477 ML14258A020 ML14275A460 ML14330A207 Unit 1 Beaver Approved ML14290A140 Valley, Unit 1 ML14363A409 H.B.

Approved Robinson, ML14251A014 ML14294A587 ML14325A693 ML15021A354 Unit 2 St. Lucie, Approved ML14206A939 ML14251A222 ML14273A011 Unit 1 ML14339A163 9

LR-N15-0036

8. References

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

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

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

4. Materials Reliability Program: Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors (MRP-111), EPRI, Palo Alto, CA, U.S. Department of Energy, Washington, DC: 2004. 1009801. [freely available at www.epri.com; NRC ADAMS Accession No. ML041680546]

5. Materials Reliability Program: Reactor Vessel Closure Head Penetration Safety Assessment for U.S. PWR Plants (MRP-110NP), EPRI, Palo Alto, CA: 2004. 1009807-NP. [ML041680506]

6. ASN, Arrt pour maintenance et rechargement en combustible du réacteur n° 3, available at http://www.asn.fr/Controler/Actualites-du-controle/Arret-de-reacteurs-de-centrales-nucleaires/Arret-pour-maintenance-et-rechargement-en-combustible-du-reacteur-n-38.

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

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

8. Materials Reliability Program 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: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement (MRP-335, Revision 1), EPRI, Palo Alto, CA: 2013. 3002000073. [freely available at www.epri.com]

10. Salem Design Calculation: S-C-RC-MDC-1928 Revision 3 Determination of Effective Degradation Years (EDY) for Salem Units 1 and 2

11. Materials Reliability Program: Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PWR Plants (MRP-117), EPRI, Palo Alto, CA: 2004. 1007830.

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

12. 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]

13. U.S. NRC, Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld in Simulated PWR Environment - 2009, NUREG/CR-7137, ANL-10/36, published June 2012. [ML12199A415]

10

LR-N15-0036

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

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

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

16. EPRI MRP Letter 2011-034, Tcold RV Closure Head Nozzle Inspection Impact Assessment, dated December 21, 2011. [ML12009A042]

17. G. White, V. Moroney, and C. Harrington, PWR Reactor Vessel Top Head Alloy 600 CRDM Nozzle Inspection Experience, presented at EPRI International BWR and PWR Material Reliability Conference, National Harbor, Maryland, July 19, 2012.

11

LR-N15-0036 Attachment 2 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 Prepared by Dominion Engineering, Inc.

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

3420 Hillview Avenue Palo Alto, CA 94303-1338 12100 Sunrise Valley Drive, Suite 220 II Reston, VA 20191 II PH 703.657.7300 II FX 703.657.7301

Dominion fn~ineerin~,lnc. 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{.f4 11, ($"",/':_;'+ G,,4-. Oh~~ A. ())~)\V

?/Z1/l) 3/2. '$/'t OI :> 3/7,;1/ t.o ,or' (j,3/23 )"l,fJ\)

K. J. Fuhr M. Burkardt O. A White G. A White Associate Engineer ,Associate Engineer Principal Engineer Principal Engineer The last revision number to reflect any changes for each section of the teclmical 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.

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 CRDM Nozzle and Bar Material Product Forms ....................................... 8 0 2.5 Conclusion ............................................................................................................9 0 3 POTENTIAL IMPLICATIONS OF SPECIFIC CATEGORIES OF NOZZLE AND WELD MATERIALS ........................................................................................................................9 0 3.1 Potential Similarities for Laboratory Specimen Material Exhibiting a Deterministic Factor Less than 12.0 ...................................................................... 9 0 3.2 Potential Implications .......................................................................................... 10 0 4 REFERENCES .................................................................................................................... 12 0 iii

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 (KI) for Alloy 690 Data from Plate Material Tested by CIEMAT.. ............................................. 14 0 Figure 2. Plot of da/dt versus KI for Alloy 690 Data from Heat WP787 .............................. 14 0 Figure 3. Plot of da/dt versus KI for Alloy 690 Data from Heat WP142 .............................. 15 0 Figure 4. Plot of da/dt versus Ki for Alloy 690 HAl Data from Heat WP142 ...................... 15 0 Figure 5. Plot of da/dt versus KI for Alloy 690 HAl Data from Plate Material Tested by CIEMAT .............................................................................................................. 16 0 Figure 6. Plot of da/dt versus KI for Alloy 152 Data from Heat WC83F8 ............................ 16 0 Figure 7. Plot of da/dt versus KI for Alloy 152 Data from Heat WC04F6 ............................ 17 0 Figure 8. Plot of da/dt versus Ki for Alloy 690 Data from All Laboratories, :::; 10% Cold Work, Constant Load or Ki .................................................................................. 18 0 Figure 9. Cumulative Distribution Function of Adjusted da/dt for Alloy 690 Data from All Laboratories,:::; 10% Cold Work, Constant Load or KI ................................... 18 0 Figure 10. Plot of da/dt versus Ki for Alloy 690 HAl Data from All Laboratories, :::; 10%

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

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

Dominion [n~ineerin~, Inc. TN-5696-00-02, Rev, 0 Last Rev.

Page Mod.

and Available in References [17] and [18]; s 22% Cold Work ...... """",,,,,,,,,,,,,,24 o Figure 20, Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data Produced by ANL and PNNL ([17] and [18]); :::; 22% Cold Work and Constant Load/KI ""'''''''''''''''''''''''''''''''''''''''''''''''''''''''"""""""""",,,,,,,,,,,,,,,,,,,,,,,,.. ,24 0 Figure 21, Plot of da/dt versus KI for Alloy 690 Data from All Laboratories, > 10 & s 20% Cold Work, CRDM and Bar Material, Constant Load or KI Testing"""""",25 0 Figure 22, Cumulative Distribution Function of Adjusted da/dt Alloy 690 Data from All Labs, :::; 20% Cold Work, CRDM and Bar Material, Constant Load or Ki ,,,,,,,,,,,,25 0 Figure 23, Plot of da/dt versus KI for Alloy 52/152 Data from All Laboratories, > 10 & :::;

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

Dominion fn~ineerin~r Inc. TN-5S9S-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 ofImprovement 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 [1] and MRP-115 [2]. In addition, an assessment is made of the available laboratory CGR data for the potential concern of elevated CGRs for specific categories of nozzle and weld materials.

Per ASME Code Case N-729-1 [3], the volumtric inspection interval for Alloy 600 RV head nozzles is based on operating time adjusted for operating temperature using the temperature sensitivity for PWSCC crack growth. The normalized operating time between inspections, called the Re-Inspection Years (RIY) parameter, represents the potential for crack growth between successive volumtric examinations. Thus, the FOI for Alloys 690/521152 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 of20 years for Alloy 690 RV head nozzles operating at 613 OF. 1 A temperature of 613 OF is expected to bound the head operating temperature for the U.S. pressurized water reactor (PWR) fleet.

As discussed in Section 3 of Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) report MRP-375 [2], a conservative approach was taken in MRP-375 to develop the factor of improvement (FOI) values describing the primary water stress corrosion cracking (PWSCC) crack growth rates applicable to Alloy 690 reactor vessel (RV) top head penetration nozzles. The crack growth rate data points presented in Figures 3-1, 3-3, and 3-5 ofMRP-375 represent the values reported by individual researchers, without any adjustment by the authors of MRP-375 other than to normalize for the effect of temperature. The data in these figures represent essentially all of the Alloys 690, 52, and 152 data points reported by the various 1 To calculate the implied FOI for the bounding RV top head operating temperature of 613°F, the re-inspection year (RIY) parameter for a requested examination interval of 20 years is compared with the N -729-1 interval for Alloy 600 nozzles ofRlY = 2.25. The representative head operating temperatures of 613°F corresponds to an RIY temperature adjustmcnt factor of 1.38 (versus the reference temperature of 600°F) using the activation energy of 31 kcal/mol (130 kJ/mol) for crack growth of ASME Code Case N-729-1. Conservatively assuming that the effective full power years (EFPY) of operation accumulated since RV top head replacement is equal to 98% of the calendar years since replacement, the RIY for a requested extended period of20 years would be (1.38)(19.6) =

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

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

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

[2]. A comparison between the cumulative distributions of the crack growth rates for Alloys 690/521152 and Alloys 600/821182 treats the full variability in both original and replacement alloys, rather than comparing the variability of the replacement alloy against a conservative mean th (75 percentile) growth rate for the original alloys. By considering the cumulative distributions, a fuller perspective of the improved resistance of Alloys 690/52/152 emerges where over 70% of the data in each of Figures 3-2, 3-4, and 3-6 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 75 th percentile values ofthe 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 690/521152 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 2 Figures 3-2, 3-4, and 3-6 ofMRP-375 show cumulative distribution functions of the variability in crack growth rate normalized for temperature and crack loading (Le., stress intensity factor). Each ordinate value in the plots shows the fraction of data falling below the corresponding normalized crack growth rate. Thus, the cumulative distribution function has the benefit of illustrating the variability in crack growth rate data for a standard set of conditions.

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

2

Dominion fn~ineerin~r Inc. TN-5696-00-02, Rev. 0 growth rates being many times higher than the deterministic 75 th 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 th than a factor of 12.0 below the deterministic (75 percentile) MRP-55 and MRP-115 equations.

2 DISCUSSION OF DATA POINTS FROM MRP-375 [2]

2.1 Data Points Above a Hypothetical 12.0 Factor of Improvement Line in Figure 3-1, 3-3, and 3-5 of MRP-375

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

There are 16 points within a factor of 12.0 below the MRP-55 75 th 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 9ARBl, comprised of Alloy 690 plate material, loaded to 37 MPa(m)05, and tested at 340°C and 15 cc H2/kg H 2 0

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

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

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

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

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

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

Dominion fn~ineerln~, 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)O,5-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 H20, 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)O,5), and the remaining six data are from specimens 9T5 or 9T6 (loaded to roughly 27 MPa(m)O.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 of Valinox heat number WP142 CRDM nozzle material that were not cold worked and were tested at 21 to 24 MPa(m)O.5, 320°C, and 23 cc H2/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 75 th percentile curve, out of a total of 34 points shown in Figure 3-3 ofMRP-375. All but one of the eight data points are for PPU testing, and all but two appear to have had very little to no intergranular engagement.

Six of the points are from ANL testing of specimens comprised of Valin ox 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-CR-1 and CF690-CR-3 (loaded to roughly 28 to 32 MPa(m)O.5) [lll' and the other point is from specimen CF690-CR-4 (loaded to roughly 22 MPa(m)o ) [12]. A plot of crack growth rate versus K for all the Alloy 690 HAZ data from MRP-375 for heat WP142 is provided here as Figure 4. As discussed below, PPU conditions-under which five of these six points were obtained-may result in accelerated crack growth relative to plant conditions.

The remaining two points are from CIEMAT testing of specimens 19ARHI and 19ARH2, comprised of welded Alloy 690 plate material, tested at 340°C and 15 cc 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 of MRP-375. Figure 3-5 shows the complete set of data points compiled by the PWSCC Expert Panel organized by EPRI at the time MRP-375 was completed for Alloy 52 and 152 weld metal specimens. The following points are within a factor of 12.0 below the MRP-115 deterministic crack growth rate for Alloy 182:

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

• One of these points is from PNNL testing of the dilution zone of a dissimilar metal weld between 152M (Special Metals heat WC83F8) and carbon steel, tested at 360°C and 25 cc H2/kg H20 [13]. This material condition is not applicable to the wetted surfaces of CRDM nozzle J -groove welds because the dilution zone where Alloy 521152 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 ofMRP-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 A152-TS-5 at loads of about 28,32, and 48 MPa(m)O.5 [14]. The other four points were obtained at loads of 5

Dominion [n~ineerin~r Inc. TN-5696-00-02, Rev. 0 27 MPa(m)05 for specimen N152-TS-l and 30 MPa(m)05 for specimens A152-TS-2 and AI52-TS-4. The Alloy 152 specimens all came from welded plate material. A plot of crack growth rate versus K for the Alloy 152 data from MRP-375 for heat WC04F6 is provided here as Figurc 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 NI52-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 ofMRP-375 took an inclusive approach to statistical assessment of the compiled data. A conservative approach was applied in which both constant load data and data under PPU conditions were plotted together. In addition, weld data reflecting various levels of weld dilution adjacent to lower chromium materials was included in the data for Alloys 521152. An assessment of the crack growth rate data points most applicable to plant conditions is presented in Figure 8 through Figure 13. The assessment shows very few points located within a factor of 12.0 below the deterministic MRP-55 and MRP-115 lines, with such points only slightly above the line representing a factor of 12.0:

• Figure 8 for Alloy 690 with Added Cold Work Less than 10%.

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

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

• Figure 10 for Alloy 690 HAZ.

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

Figure 11 shows that the data are bounded by an FOI of more than 12 relative to Alloy 600 data on a statistical basis.

• Figure 12 for Alloys 521152.

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 fn~ineerin~r Inc. TN-5696-00-02, Rev. 0 equations ofMRP-55 and MRP-115. The MRP-55 and MRP-115 deterministic crack growth rate equations are not bounding equations, but rather reflect the 75 th percentile of material variability. Thus, the perspective provided in Figure 9, Figure 11, and Figure 13 is most relevant to drawing conclusions regarding FOI values applicable to inspection intervals for heads fabricated using Alloy 690,52, and 152 materials.

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

• As demonstrated and discussed in MRP-115, certain PPU conditions will act to accelerate the crack growth rate. PPU conditions, which include a periodic partial reduction in load, are often used in testing to transition from initial fatigue conditions toward constant load conditions with the crack in a state most representative of stress corrosion cracks if they had initiated in plant components over long periods of time. The periodic load reductions and accompanying load increases may rupture localized crack ligaments along the crack front, facilitating transition of the crack to an intergranular morphology. In MRP-115, data with hold times less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> were screened out of the database for Alloys 821182/132.

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

• The Alloy 521152 weld metal data shown in Figure 3-5 and Figure 3-6 ofMRP-375 include data reflecting a range of weld dilution levels. The data presented in Figure 12 and Figure 13 exclude the weld dilution data points because of the limited number of data points available, the variability in results, and the limited area of continuous weld dilution for potential flaws to grow through. The weld dilution data are not reflective of the full chromium content of Alloy 521152 weld metal.

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

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

The limited number of remaining points in Figure 8 and Figure 12 that lie within a factor of 12.0 below the deterministic MRP-55 and MRP-l15 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 690/521152 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 600/821182 crack growth rates supports a reexamination interval longer than the requested interval corresponding to an FOI of 12.0. The volumetric or surface inspection interval for heads with Alloy 600 nozzles reflects consideration of crack growth rates on a statistical basis, with crack growth rates often higher than that given by the deterministic equations ofMRP-55 and MRP-lI5.

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 690/521152 that have been generated by ANL and PNNL, so the data specific to these national laboratories have also been evaluated separately. Based on the compilation of ANL and PNNL crack growth rate data recently released by NRC [17]4, the results are shown in Figure 15 through Figure 20. These data reflect Alloy 690 test specimens with up to 22% added cold work. The data in Reference

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

As shown in Figure 15, Figure 17, and Figure 19, only 10 of the total 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 521152 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 tn~lneerin~, 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) 1 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 COR performance of Alloys 690/52/152. Thus, the available laboratory COR 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 COR data for the potential concern of elevated CORs 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~r Inc. TN-5696-00-02, Rev. 0

• Figure 3-1 of MRP-375 [2]. The only Alloy 690 CRDM material for which crack growth rate data were available at added cold work ofless than 10% (the threshold for inclusion in Figure 3-1 ofMRP-375) was supplied by Valin ox Nucleaire. The few data using CRDM material from other suppliers were obtained at cold works of 20% or higher and were not included in the assessment. The data do not indicate any correlation between material supplier and susceptibility to crack growth rate. Fourteen of the Alloy 690 crack growth data points within a factor of 12.0 below the MRP-55 [1] deterministic crack growth rate in Figure 3-1 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 ofPPU for eight of these 14 data), this similarity in no way indicates any specific concern for elevated PWSCC susceptibility of the head nozzle material provided by anyone supplier.

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

• Figure 3-5 ofMRP-375 [2]. There are no relevant similarities between (a) the Alloy 52 and 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 of replacement RV heads. Furthermore, the weld specimens used in the crack growth rate testing were machined from test welds in flat plates, not from actual I-groove welds. Thus, the test weld specimens should not be associated with particular fabrication categories of replacement heads.

3.2 Potentjallmplicatjons 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 tn~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 anyone 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 ofMRP-375 reflects a factor of improvement of 12 or greater. In addition, other factors being equal, one would expect a greater range of crack growth rates for a material heat for which a greater number of data points was produced. Some of the scatter likely reflects experimental uncertainty as opposed to true material variability.

Experimental uncertainty is more of a factor for the data for Alloys 690/521152 than for Alloys 600/821182/132 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 ofPWSCC initiation is necessary to produce a flaw that may grow via PWSCC.

Laboratory and plant experience show that Alloys 690/521152 are substantially more resistant to PWSCC initiation than Alloys 600/821182 [2]. PWSCC has not been shown to be an active degradation mode for Alloys 690/521152 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 1A WS ERNiCrFe-7), Alloy 52M (UNS N06054 1AWS ERNiCrFe-7A), Alloy 52MSS (UNS N06055 1AWS ERNiCrFe-13), Alloy 52i (AWS ERNiCrFe-15), Alloy 152 (UNS W86152 1 AWS ENiCrFe-7), and Alloy 152M (UNS W86152 1A WS ENiCrFe-7). Considering the overall set of available crack growth rate data for the various variants of Alloy 52 and 152, there is no basis for concluding at this time any significant difference in the average behavior between the Alloy 52 and Alloy 152 variants in use at u.s. PWR RV heads with Alloy 690 nozzles.

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

Dominion fn~ineerin~r Inc. TN-5696-00-02, Rev. 0 heads with Alloy 600 nozzles, for which PWSCC is an active degradation mode, materials and fabrication categories of heads with relatively high incidence of PWSCC are inspected in accordance with the same requirements as other heads.

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

4 REFERENCES

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

Revision 1, EPRI, Palo Alto, CA: 2002.1006695. [freely available at www.eprLcom]

2. Materials Reliability Program Crack Growth Ratesfor 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.eprLcom]

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

4. "Materials Reliability Program: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375), EPRI, Palo Alto, CA: 2014.3002002441. [freely available at www.epri.com]

5. Materials Reliability Program: Inspection Planfor Reactor Vessel Closure Head Penetrations in Us. PWR Plants (MRP-l17), EPRI, Palo Alto, CA: 2004. 1007830. [freely available at www.eprLcom; NRC ADAMS Accession No. ML043570129]

6. Materials Reliability Program: Reactor Vessel Closure Head Penetration Safety Assessmentfor us. PWR Plants (MRP-llONP), EPRI, Palo Alto, CA: 2004. 1009807-NP.

[ML041680506]

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

2004. 1007834. [ML041680489]

8. D. G6mez-Briceno, J. Lapena, M. S. Garcia, 1. Castro, F. Perosanz, and K. Ahluwalia, "Crack Growth Rate of Alloy 690 /152 HAZ," Presented at: Alloy 6901152152 Research Collaboration Meeting, Tampa, FL, December 1-2,2010.

9. D. G6mez-Briceno, J. Lapena, M. S. Garcia, 1. Castro, F. Perosanz, 1. 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. 109-125,2011.

12. B. Alexandreanu, Y. Chen, K. Natesan and B. Shack, "Update on SCC CGR Tests on Alloys 690/521152 at ANL - June 2011," Presented at: US NRCIEPRI Meeting, June 6-7, 2011. [ML111661946]

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

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

179-196,2011.

15. B. Alexandreanu, Y. Chen, K. Natesan and B. Shack, "Cyclic and SCC Behavior of Alloy 152 Weld in a PWR Environment," Presented at: Alloy 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 Investigation of Stress Corrosion Cracking in Nickel-Base Alloys, NUREG/CR-71 03, Vol. 2, April 2012.

19. Materials Reliability Program: Material Production and Component Fabrication and Installation Practicesfor 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-13 ~-'-'--'-!-'-'-'---'-+-'-'--'-'--t--'-~-+--'--'-'-....I.....j--'---'--'-'-i-'--"L.....J.......J--r-1::;::::::L:::L:i=::i::i::~f=W::w=j 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-lm)

Figure 1. Plot of Crack Growth Rate (da/dt) versus Stress Intensity Factor (KI) for Alloy 690 Data from Plate Material Tested by CIEMAT

1. E-13 -'-'--'-'-'-I-'c......L......L-'-t-'-'-...l.-...1-+-'-.I-l---J'-t---'--'-'---'-+-'-'--'-'-t--'-L....1-.I~:::.::::::;::::y::::;::;::w::::i=i=i::::w::j 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-lm)

Figure 2. Plot of da/dt versus KI for Alloy 690 Data from Heat WP787 14

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

1. E-13 -l-c--'---'--'---+--'---'--'--'-t-'--'-'-'-t--'-'-'---'--4.....~'-'-'-+-L.......l.-JL...l..-J---'--J-'--'-+-'::w::::;:::;::::;::::L::;:::;:=i=:i:::::L:i:~

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

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

1. E-13 +.c-.I-l-.....J---t--'--'-'--'-!-.L...!-I...-L.f-'--'--'---'--t-'-'--'--'---t--'--'-'--'-+J......L...,;I...-L.f~::w::::t=L=i:::i:::i=t=L::w:::::c:j 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPavim)

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

Dominion fnvineerin~r Inc TN-5696-00-02, Rev. 0

1. E-13 ~.L.......I.....-+--'--'-'---Y-..l....L-'---'-t-'-'---'--'-+-'-.L.......I.....-+--'--'-'---Y--'--'-'---'-t.1::L:::;::;:~:c::::;:::;:::;::::;::::;::;::;::j 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPav'm)

Figure 5. Plot of da/dt versus KI for Alloy 690 HAZ Data from Plate Material Tested by CIEMAT 1.E-10

~

~

i 1.E-11

~

u 5 1,E-12 -j:;;;;;========

- Data are adjusted for

-, temperature (325°C).

Q = 130 kJ/mol

1. E-13 +-'-..J.......L..-'-+..l-I-..J........I...-+--'-...J....L-'-+-'-1.--'--'--J-L......J-L--'-I--'--'--'--'--I-'~L......l..-I-'--'-'--L+-L....J....J.......L.f-L-.J,...L..o""-l 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPav'm)

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

Dominion tn~ineerin~, Inc TN-5696-00-02, Rev. 0 1.E-10 ~ +PNNL E

~

- 1.E-11

~

to!) ~. PPU data are Black*filled ANL data present tlC'C .... represen ted WI'th I*****************************************************....................................................1 growth rates during the

.... open s mbols environmental pre-crack period u 1.E-12 lr~~:~;'dt;l~~~~~g~~k:~a~nd~s~ho~u~ld~n~ot§b~e 1=* Data are adjusted for ffiin~clL~ld~ed~,~.~. .~~_~~_

_. temperature (325°C).

Q = 130 kJ/mol

1. E-13 +--'--'---'--'-+-'-'--'-'--t--'---L....J.--'---t-'--'--'--'--"F-'-'--'---'--l---'----'--'-+-'-'-'---'-t---'-'---'-"-l--'---'--'---'--t-'-........."'"-I 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa.ym)

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

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

4IPANL

4IP Bettis"-~~"~'-'-,*",*,~~,",~,~"""""""'""'~"'~~m'"""'_~~"""-~"I MRP-551~~~'~

A CIEMAT 1 Curve/1 I 1,E-10 I  :~",!"PN~L . ~;;;~~~~:5?;'~:~::;:::i;;:=~X=:=::§:.;§f:~;;:;~:~,~",,,,,:..:~~=:::::::::;::;~~,,;:~§;;;;~:S~:=:=

""';9 MRP/12 r-'

• • .. '.w .""'UAW..W.WN.wN"'"MW ~ m' ""uN"" u,"" ..."~_"w,, .,," *.,....WN<,,,.v. ......,..,.. ""'"

1-,~.~,,'~-/-:7,e,;;."""~~~'-'""""~'~~'''"-""'''~~',_""'",,"',,'~,,'~,',"'''_m"~'_,,~

~ /'~-~''--'-"~-i.-~_",'m,AAi--= __

~ ""A 1, E-13 10

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

Figure 8. Plot of da/dt versus KI for Alloy 690 Data from All Laboratories, :::; 10% Cold Work, Constant Load or KI 4IPANL 4IP Bettis A CIEMAT 1------:""iJ%.---""'"""'7-----~r-----------1

§ 0,7 .. PNNL l--:/i,.rl,;X-<'" " - - - - - - - : / - - - - - - - - - f - - - - - - - - - - - - - l

§ *~

g 0.6 +---~-----7I-l------I~------------1 t;)

~ 0.5 ------J/,......--,--;::;:::::;::;:I:;:;;:;------I The data points at 1E-13

> were reported as "no

~ 0.4 +...- - - , ---1--- growth,"

E B 0.3 Il!i!--------.#----------f------,I Data are adjusted for temperature (325°C) and 0.2 .----L~:....gJ stress intensity factor.

0,1 II----r---- Q = 130 kJ/mol

.", K = 30 MPa,fm 0.0 '--==-.:::........"'""""i'--~~1_1.f_------'---'---'--'--'--1..J....L_I_-'--====F====rlj 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1,E-08 Crack Growth Rate (m/s)

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

Dominion fn~ineerin~r Inc. TN-5696-00-02, Rev. 0 1.E-12 Data are adjusted for temperature (325°C).

Q =130 kJ/mol

1. E-13 -\-.--'--'-'---t--'--1-l.-'-+--'-'--'-'-+-'-.L-I.....J~--'--'-4-'-'--'-'-1-'-'---'--'~::i:::i::::L::j::::i::i:::;:::::;:::t=l=i~

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

Figure 10. Plot of da/dt versus KI for Alloy 690 HAl Data from All Laboratories, :S 10% Cold Work, Constant Load or KI 1.0 II! ,.....-

0.9 II! /' ~ II!ANL

/

tW I-A / ACIEMAT 0.8 III

/

l-II! / KlGE-GRC III

§ 0.7

.'§

§ 0.6

... I I

/

.PNNL I-Vi III I

/

~ 0.5 L The data points at 1E-13 I MRP-55 I I

- were reported as "no

'p

-5 I 0.4 v (FOI = 1) growth."

E

<3 0.3 / / Data are adjusted for 0.2

• FOI

/

12 / temperature (325°C) and 0.1 1m

./

I / -

stress intensity factor.

Q = 130 kJ/mol 0.0

_/ / K = 30 MPa.ym 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, :S 10% Cold Work, Constant Load or KI .

19

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

]r: : :l~ ~" ]"~ "'~" "~"~" " ~" "-.~' ~ ":;.-~-

1.E-09 :; @!)ANL  :'2':; ; ;: ~:0.-..

'-~" ',,~~-"'~" "'~" " '~m~":'~";" " ~" : '"~'c, ~" " ~"~" .§" .:; ; ;~.: ~ ".".-..~.~. . . .~. :;.;;-~. . ; :~:; ; ; ~.;~,.,~""~m'.m.~'.~_.~,

mm ....... m....... _

.... ; ;:.; .;.~

~.~. . . .~

. . . . ~.~~

. .: '-. .§:~'-" "~"~ ;':~:;~" ~,.'*w_'m~

.... "~ "§": :'-"'~" -" ";

_ _"""_mWN'~~' .. ".~~,_.'w .....

._w_."_,. .~"_..m~"'.'._

~*""'~~,,~'_" __ ~M'~~"_ "'''~~'~_'''''~~m'~'~~~''' ~~M"~ . . . . . .'*,.,.~ . . "~,,v~~

&\ CIEMAT "~. "....._, __,____,,,,'__ .... ___~

1.E-10 i

~ l.E-11

<.:l

~

u e 1.E-12 t:F=~="""""""'===

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

Q =130 kJ/mol

1. E-13 +-,--'-'--'--I--'-'-...1.--L.-I-'---'--'--'-I-~--'--'~~i'tiDfl!l--'--'~4-L-1-L-1-t--'--.L..-L-"'-I-.l....-L.-'---'-~",-,--'---l 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa~m)

Figure 12. Plot of daldt versus KI for Alloy 52/152 Data from All Laboratories, :S 10% Cold Work, Constant Load or KI 1.0 ,--;=====;--------------=----QPJ--,-~----~---_.

@!)ANL ~ @!) /

0.9 &\CIEMAT I---------c-&\...--~ @ ! ) 7

[I] GE-GRC 1-_--.:!!l!f...::I1'=---_I!A_ _ _ _ _ _ _ _ _--I-_ _ _ _-.,..,"':'=-::~",..,.

0,8 +PNNL /

§ 0,7 +=======

.§

@ 0.6 + - - - - - - - - - - - - - - - / - - - - - - - - - / - - - - - 4 t;)

~ 0.5 + - - - - - - d i f § ' " - . - - - - - - - - - - The data points at 1E-13 i

.2::

0.4 +--~r-----------_f_--_I:======~gr=owt==h.=H====~

were reported as "no 8 0.3 I-:I""'-------*-----;::==~--- L Data are adjusted for IFOI = 121 temperature (325°C) and 0.2 +millL......-~----------~:::.*:::::::::::::::!.......---1 stress intensity factor, 0.1 1 l l \ I I - - - - - - - - . - - - - - - - - r - I - - - - - - - I Q = 130 kJ/~ol K = 30 MPa'im 0, 0 ~---'--........................._r__-~...;:::::... ............,.._........-.d.:':w:::i:i:j====::::;:::::;:::::::;:::c::::;::w:ij l.E-13 l.E-12 1.E-11 1.E-10 1,E-09 Crack Growth Rate (m/s)

Figure 13. Cumulative Distribution Function of Adjusted daldt for Alloy 52/152 Data from All Laboratories, :S 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 "E"'=-'~=:;:::=~:::;;;,=:~=;;;,'~"'='~='~=~-=';;;;=:~%='~-="",:;=~:;~=",,,,;,=,,,;.~=;~=,~~.=,;;;;;,,=~,=~,=,-~~,,:=

':::~"r,=======;:;:;;:::;::::::F===n

=.......

_ _~'W'.., .....w.

Data are adjusted for .,~ Specimen

=w.WmN"~"m~_"Nm'W"""N"'_~'M''''=''''''''''''.w., .~

=:2::"AA~::~=-_~~'~~m~:::=::=:~~~:::::::~:==:::=~:::::~.,,~, ~':::::N~

1~,,"*,**"m'_'''''~__WM'__'_~'~~'~'~"",,,,,,,,,,,,,1 temperature (325°C) (0 = 130 9T1

""''-''- kJ/mol) and K (30 MPa'-'m) ~9T2 1.E-10 i ""'*"" 9T3

~

e

' 1.E-11

<..'j

~

e

(..)

u 1.E-12 10 100 1000 10000 Hold Time (Hours)

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

Dominion fn~lneerin~, Inc TN-5696-00-02, Rev. 0 Compilation of ANL and PNNL Data 1.E-10

~

~

~ 1.E-11

,,- (i) ANL CL

"'~.

~".

~

<.!:)

...-"'" v U .... OANL PPU .

@ ~

!S!

(,,) /' ~ .PNNL 1.E-12 ~

Data are adjusted for

r=~=;: temperature (325°C).

4fe'!

.~ Q = 130 kJ/mol 1.E-13 .... ....v .....

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

Figure 15. Plot of da/dt versus KI for Alloy 690 Data Produced by ANL and PNNL and Available in Reference [17]; s 22% Cold Work 1.0 -,----------~--_.J5---,,~;;;iiiiiIIF"~-----:::;;;;;;;O-------,

0.9 +----------;;;:~-'-------F_-------______l O.S +--------~~__r-----_+_--------______l

.g 0.7 +--------:;r--~--J'--------+---------r-(i)---:-:AN-L-C-L---H

l

§ 0.6 + - - - -. .- " - - - - { - - - - - - - - I - - - - - - - - - I .PNNL tl t

~ 0.5 1---:--:,--*----r----~~f;;;::1-1 The data points at 1E-13 are E

0.4 ***. *. ****.. *"t....". . **.*_. *"*_* " "-_*,. _../-. . *, . *, * *. . . . "."" . . . ."., . L..J,.:.....jr--~

c~r~~~~~n~s~;~~ ~ ~~;~.

(5 0.3 Data are adjusted for 1FOI = 121 temperature (325°C) and 0.2 ~----..!::::/==i===:L.....----+---------I stress intensity factor.

0.1 :It----~<--------~--------I Q = 130 kJ/mol

./ K= 30 MPa-vm 0,0 ~-==",:w.........o.;_-~~-,-,+_-'---'---'--'--'-l..J...L.j----,---'::c==+=====LI 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 i.E-OS Crack Growth Rate (m/s)

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

Dominion tn~ineerin~, Inc. TN-5696-00-02, Rev. 0

1. E-13 -J-,..-'-'-L....f-'-'---I.-'-t--'--'--'--'-+-'-.L..-.L...J............I-L--'--+--'-'---'-'-1-'-L-l-l~:::w:::::;::;::::L::i:::L:::i::::p:::3~

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

Figure 17. Plot of da/dt versus KI for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17]; :::; 22% Cold Work 1.0 ,.,..-

0.9 -"

,- /"

/

I@I

• / -

0.8

/ /

§ 0.7

• v I t>ANL CL

'g

§ 0,6 1i)

• +

.................._n._u............."..........___........__........._......_.m ...... "." .... ..

" -/".... .................--.-.--......... ..-........

~

. . . . . . . . . . . . _ . _ . _ _ _ _ . . <<O .............. " . . . . . . . . . . . . . . . . _~_ ...

~PNNL -

~ 0,5 ,I The data points at 1E-13 are

> ~

I I (FOI MRP-55 I

=1) treated as "no growth,"

~ 0.4 consistent with MRP-375.

8 E

0,3 / /

/

Data are adjusted for 0.2

~

iFOI =12i_" temperature (325°C) and

/ / stress intensity factor.

0.1 Q =130 kJ/mol 0.0 1,E-13

- //

1,E-12

/

1.E-11 1,E-1 0 K =30 MPav'm 1,E-09 1,E-08 Crack Growth Rate (m/s)

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

Dominion fn~ineerin~r Inc. TN-5696-00-02, Rev. 0 Data are adjusted for MM temperature (325°C).

Q = 130 kJ/mol

1. E-13 ~:i::i::~:::::;::;:~::::c::::::i:~+-,-.1-l-'-t--4>-'-'-~...J.......J-~+--"-'--'-4H-'---'-'--'-+-'--'-.l-l...-t--'--'~

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

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

/

@ANLCL @

0.9 """

0.8

• PNNL CK (NUREG) II)

II)

I

/ /

II)

IMRP-115L

!FOI-1)

§ 0.7 J

"*l~* . .-.. --*-..-***. .* .*. -.. -..****-I-..*-. *. * . * *. . . *-_

'p

g

~

0.6 _ ....n .......... ........... ................................... ...................... ........ .... *********** _ . . . . . _ ** ~ _ . . . . . . . <U ............ _ _ _

• _ _ . . . . _ _ ******** _ **

V;

~ 0.5 ,J The data points at 1E-13

-s~ 0.4 *

. ** l-were reported as "no growth,"

E 8 0.3 I Data are adjusted for

~~~. .______ ~___..~IFOI = 12] temperature (325°C) and 0.2 stress intensity factor, 0.1

< / Q = 130 kJ/mol

< ...-"'/ K = 30 MPav'm 0.0 1.E-13 1.E-12 1. E-11 l.E-10 1.E-09 Crack Growth Rate (m/s)

Figure 20. Cumulative Distribution Function of Adjusted daldt Alloy 52/152 Data Produced by ANL and PNNL ([17] and [18]); :S 22% Cold Work and Constant LoadlKI 24

Dominion fn~ineerin~, Inc. TN-5696-00-02, Rev. 0 Data for Less than 20% Cold Work from All Laboratories 1.E-12 Data are adjusted for temperature (325°C).

Q =130 kJ/mol

1. E-13 -J-.-.1..-.l-'-l-'--'--I.-'--+--'-'-...L....l-+--'-.l-L...J'-f--J---'--'--'-I--'-'---'--'-~'--'--'~::L::i:~::w::::w:::i=L=i~

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, CRDM and Bar Material, Constant Load or KI Testing 1.0 &. ".-

0.9

-.>AMEC t.;.&' / ' ~

O.S il"CIEMAT

<) t/ /

/

~GE-GRC

~vl/

§ 0.7 .PNNL

'"§

§ 0.6 c--- ~

~I

/

/

I Vi tf' ,r The data points at 1E-13

~ 0.5 I' I MRP-55 I were reported as "no t-

~ 0.4 .4. Jl>.

(FOI = 1) growth."

L

~

8 E

0.3 / "----

/

Data are adjusted for t I temperature (325°C) and 0.2 t

/

fOI = 121 stress intensity factor .

0.1 ..,.4. ~' Q = 130 kJ/mol 0.0 *-

j

.,/

7

/ K = 30 MPa.ym 1.E-13 l.E-12 1.E-11 1.E-10 l.E-09 1.E-OS 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 fn~lneerin~, Inc. TN-5696-00-02, Rev. 0 1.E-10

~ - IIlil

~

1. E-13 -t-'-...l-l...-'-t---'--'-...J...-L+-'-...J...-L-'-t...l.-L-J.....J.-+-'-L....J.--'--t--'-'--'---j-I~~r-'---'-L....f-.l-L.................-t-'--'--Y 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-lm)

Figure 23. Plot of daldt versus KJ for Alloy 52/152 Data from All Laboratories, > 10 & ~ 20% Cold Work, Constant Load or KJ 1.0 -,;===='i,--------------~.....-,_.=----~-----.,

~ANL /

0.9 ~ CIEMAT f------------;;---,~--'I!i1~-__,,___-----___,F__1 0.8 EJGE-GRC r---~~-----------,__r_----~~~~-J

+PNNL

§ 0.7 -l-'======='-----"Ir-----------

'g

§ 0.6 +-----~~---------__+------_I---__I V;

~ 0.5 +--, The data points at 1E-13

> were reported as "no

~ 0.4 -~P"_----------_I_--__I growth."

8 E0.3 +-~ffiP1------~----_r=:::=:;j==,----1 ~====C=======~

Data are adjusted for temperature (325°C) and 0.2 + m ! I I F - - - - - - - - - - - - - - . - E - - - - - - - - - - J stress intensity factor.

0.1 1 I 1 - - - - - - - - - - - / - - r - - - - - - - l I Q = 130 kJ/mol K = 30 MPa'-lm 0.0 ' -___"'-'...................oIoj-_..........:::"",c:..................,...._____,d::w:i::i:i===::::::;::::::::;::::::;:::::;::::w::::d.j 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (m/s)

Figure 24. Cumulative Distribution Function of Adjusted daldt Alloy 52/152 Data from All Laboratories, ~ 20% Cold Work, Constant Load or I<J 26