RNP-RA/17-0043, Relief Request (RR)-12 for Relief from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-4

Relief Request (RR)-12 for Relief from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-4

ML17269A016
Person / Time
Site:	Robinson
Issue date:	09/22/2017
From:	Sherman C Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RNP-RA/17-0043
Download: ML17269A016 (47)

Text

(~ DUKE ENERGY" Serial: RNP-RA/17-0043 SEP 2 2 2017 Attn: Document Control Desk United States Nuclear Regulatory Commission Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23 Charles E. Sherman H. B. Robinson Steam Electric Plant Unit 2 Director - Nuc Org Effectiveness Duke Energy 3581 West Entrance Road Hartsville, SC 29550 0: 843 857 1609 F: 843 8571319 Ch11ck.Sherma11@d11ke-e11ergv.co111 10 CFR 50.55a RELIEF REQUEST (RR)-12 FOR RELIEF FROM VOLUMETRIC/SURFACE EXAMINATION FREQUENCY REQUIREMENTS OF ASME CODE CASE N-729-4 Pursuant to 10 CFR 50.55a, Duke Energy Progress, LLC hereby requests relief from performing the required volumetric/surface examinations for the H. B. Robinson Steam Electric Plant (RNP),

Unit No. 2, reactor vessel closure head (RVCH) components identified in ASME Code,Section XI, Code Case N-729-4. The details and justification for this request are enclosed.

Duke Energy Progress, LLC requests approval of the enclosed relief request by June 2018 to support the Fall 2018 refueling outage.

This letter contains no new Regulatory Commitments.

If you have any questions concerning this matter, please contact Mr. Tony Pila, Manager -

Regulatory Affairs at (843) 857-1409.

Sincerely, ~c Charles E. Sherman Director - Nuc Org Effectiveness CES/cac

Attachment:

H. B. Robinson Steam Electric Plant, Unit No. 2, Relief Request (RR)-12 for Relief from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-4

Enclosure:

Dominion Engineering, Inc. - Technical Note (TN-5696-00-02, Rev. 0) cc:

NRC Regional Administrator, NRC Region II Mr. Dennis Galvin, NRC Project Manager, NRR NRC Resident Inspector

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 1of13 H.B. Robinson Steam Electric Plant, Unit No. 2, Relief Request (RR)-12 for Relief from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-4

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 2of13

1.

American Society of Mechanical Engineers (ASME) Code Component(s) Affected The affected components are ASME Class 1 Pressurized Water Reactor (PWR) Reactor Vessel Closure Head (RVCH) nozzles and partial-penetration welds fabricated with primary wat~r stress corrosion cracking (PWSCC) resistant materials. H. B. Robinson Steam Electric Plant (RNP) Unit 2 penetration tubes and vent pipe are fabricated from Alloy 690 with Alloy 52/152 attachment welds.

2.

Applicable Code Edition and Addenda

The 5th inservice inspection (ISi) interval Code of record for RNP Unit 2 is the 2007 Edition with 2008 Addenda of ASME Boiler and Pressure Vessel Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components."

3.

Applicable Code Requirement

§ (0) of the Code of Federal Regulations (CFR) 10CFR50.55a(g)(6)(ii) is the Augmented ISi requirements for Reactor Vessel Head Inspections.

10CFR50.55a(g)(6)(ii)(D)(1), requires (in part):

"Holders of operating licenses or combined licenses for pressurized-water reactors as of or after August 17, 2017 shall implement the requirements of ASME BPV Code Case N-729-4 instead of ASME BPV Code Case N-729-1, subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through ( 4) of this section, by the first refueling outage starting after August 17, 2017."

ASME Code Case N-729-4, -2410 specifies the Inspection Program requirements for the reactor vessel upper head and its penetrations (nozzles and partial-penetration welds).

The basic inspection requirements of Code Case N-729-4, -2410 as contained within Table 1 for partial-penetration welded Alloy 690 head penetration nozzles are as follows:

Item 84.40: Volumetric or surface examination of all nozzles, not to exceed one inspection interval (nominally 10 calendar years) provided that flaws attributed to primary water stress corrosion cracking (PWSCC) have not been identified.

Item 84.30: Direct visual examination (VE) of the outer surface of the head for evidence of leakage every third refueling outage or 5 calendar years, whichever is less.

The Item 84.40 volumetric and/or surface re-examination interval of ASME Code Case N-729-4 is identical to that of Code Case N-729-1, which was mandated by NRC prior to August 17, 2017 by 10CFR50.55a(g)(6)(ii)(D). The previous NRC conditions on N-729-1 and the current NRC conditions on N-729-4 in 10CFR50.55a(g)(6)(ii)(D) do not affect the re-examination interval required for Item 84.40.

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 3of13

4.

Reason for Request

The treatment of Alloy 690 RPV Closure Heads in Code Case versions N-729 through N-729-4 (Ref. 1) was intended to be conservative and subject to reassessment once additional laboratory \\data and plant experience on the performance of Alloy j90 and Alloy 52/152 weld metals become available. Using plant and laboratory data Electric Power Research Institute (EPRI) document Materials Reliability Program (MRP) - 375 was developed to support a technically based volumetric or surface re-examination interval using appropriate analytical tools. This technical basis demonstrates that the re-examination interval can be extended to a 20 year interval length while maintaining an acceptable level of quality and safety.

The NRC has previously approved a one-time deferral of the next volumetric or surface inspection of the RNP Unit 2 head nozzles by approximately 3 years until the refueling outage scheduled to commence in September of 2018 (Ref. 2). This deferral was intended to provide sufficient time for the NRC to review and accept the conclusions reached in MRP-375 (Ref. 3).

However, it is now expected that additional time beyond the currently authorized deferral will be necessary for NRC to have all the information it needs to consider the conclusions reached in MRP-375 (Ref. 3). Specifically, the detailed review of crack growth rate data by the international group of experts cited by NRC in Reference 2 is currently expected not to be complete until late 2017. Duke Energy is requesting approval of a one-time extension beyond the previously approved one-time extension of the nominal 10-year re-examination interval for Item 84.40 of Table 1 of Code Cases N-729-1 and N-729-4. The requested extension is until the refueling outage scheduled to commence in September of 2020.

5.

Proposed Alternative and Basis for Use Proposed Alternative Pursuant to 10CFR 50.55a (z)(1 ), Duke Energy requests an alternative from performing the required volumetric or surface examinations for the RNP RVCH components identified above at the frequency prescribed in ASME Code,Section XI, Code Case N-729-4. Specifically, Duke Energy requests to extend the frequency of the volumetric or surface examination of the RNP RVCH of Table 1, Item 84.40 of ASME Code Case N-729-4 for an additional one (1) operating cycle beyond that approved by the NRC in accordance with RNP Relief Request (RR-11) (Ref. 2). This request would extend the volumetric or surface examination to the 32"d refueling outage, which is scheduled to commence in September of 2020 (within 6 years beyond the nominal 10 years required by ASME Code Case N-729-4 in order to align with a scheduled refueling outage). No alternative examination processes are proposed to those required by ASME Code Case N-729-4, as conditioned by 1 OCFR50.55a(g)(6)(ii)(D). The visual examinations and acceptance criteria as required by Item 84.30 of Table 1 of ASME Code Case N-729-4 are not affected by this request and will continue to be performed on a frequency of every 3rd refueling outage or 5 calendar years, whichever is less.

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 4of13 As documented in Section 7 (Precedents), NRC has approved similar extensions for other replacement RVCHs of more than 5 years beyond the nominal 1 o years required by ASME Code Case N-729-1 in order to align with scheduled refueling outages. NRC has approved extensions of 6 years for this purpose. Duke Energy requests the total time deferral of approximately 5 years (an additional 2 years beyond the previously approved deferral of 3 years, ML15021A354) for RNP for the purpose of aligning with the scheduled RNP Unit 2 refueling outages.

Basis for Use The original RNP 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 that returned to operation in October 2005. In accordance with NRC approval of RNP Relief Request (RR-11), RNP will be required to perform a volumetric and/or surface examination of all nozzles during the 31st refueling outage, which is scheduled to commence in September of 2018.

The basis for the volumetric and/or surface examination frequency for heads with Alloy 690 nozzles of ASME Code Case N-729 through N-729-4 comes, in part, from the analysis performed in EPRI Materials Reliability Program (MRP)-111 (Ref. 4), which was summarized in the safety assessment for RVCHs in EPRI MRP-11 O (Ref. 5). The material improvement factor for Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 690/52/152 materials over that of mill annealed Alloys 600/82/182 was shown by this report to be in 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 EPRI MRP-375 (Ref. 3). This report presents both deterministic and probabilistic evaluations that assess the improved PWSCC resistance of Alloys 690/52/152.

Additional Evaluations Performed under EPRI MRP-375 As documented in EPRI MRP-375 (Ref. 3) operating experience for replacement and repaired components using Alloys 690/52/152 has shown a proven record of resistance to PWSCC during numerous examinations in the 25+ years of its application. This includes steam generators, pressurizers, and RVCHs.

In particular, at the completion of the spring 2017 refueling outage season, Alloy 690/52/152 operating experience includes inservice volumetric or surface examinations performed on 16 of the 40 currently operating plant replacement RVCHs in the US in accordance with the Augmented ISi Requirements for Reactor Vessel Head Inspections.

In France in 2013, a second 10 year nondestructive examination (NDE) inspection was performed on one of the first reactor vessel (RV) heads to be replaced with Alloy 690/52/152 material. 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 Alloy

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 5 of 13 690 compared to Alloy 600 at equivalent temperature and stress conditions. Even though base metal and welding variability of test data exist (i.e. heat affected zones, weld dilution zones, etc.), relative but conservative FOls were estimated for the material improvements of Alloy 690/52/152 materials using an extensive database of test data.

Results for both crack initiation and crack growth conclude a higher resistance to PWSCC for Alloy 690 base material and Alloy 52/152 weld materials. EPRI MRP-375, Figures 3-2, 3-4, and 3-6 provide crack growth data for Alloy 690/52/152 materials and heat affected zones with represented curves plotting FOls of 1, 5, 10, and 20. A FOi of 20 bounds most of the data plotted; however, a FOi of 10 or less bounds all of the data.

EPRI MRP-375, Table 3-6, provides a summary of crack growth rate (CGR) and crack initiation data. For crack initiation, FOls reported, although significant, are conservative because in many cases 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.

Additionally, many of the Alloy 690 crack growth rate tests were performed on specimens with considerable amounts of cold work (up to 40%), which is known to accelerate CGRs to rates that are not representative of cold work levels applicable to reactor vessel head penetrations.

EPRI MRP-375 then performed a combination of deterministic and probabilistic evaluations to establish a reasonable 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 CGR 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 which a conservative inspection interval may be recommended. This evaluation draws from various EPRI MRP and industry documents which 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 inside diameter (ID) axial cracking and of 10 to outside diameter (OD) axial cracking for Alloys 690/52/152 versus Alloys 600 and 182, 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 EPRI 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 EPRI MRP-335, Rev.1 (Ref. 6) that evaluated surface stress improvement of Alloy 600 RVCHs. The integrated probabilistic model in EPRI MRP-375 includes submodels for simulating component and crack stress conditions, PWSCC initiation,

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 6of13 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 Alloy 690 RVCHs by applying FOls to account for its superior PWSCC resistance. The probabilistic calculations are based on a Monte Carlo simulation model including PWSCC initiation, crack growth, and flaw detection via ultrasonic testing. The average leakage frequency and average ejection frequency were determined using conservative FOi assumptions.

The results show that using only modest FOls for Alloy 690/52/152 RVCHs, the potential for developing a safety significant flaw (risk of nozzle ejection) is acceptably small for a volumetric or surface examination period of 20 years.

The evaluations performed in EPRI MRP-375 were prepared to bound all 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).

While Duke Energy is not requesting NRC review and approval of EPRI MRP-375 to approve this request for alternative, the insights gained in this technical report help substantiate the limited extension duration being requested for RNP of one operating cycle (32nd outage) beyond the 31 51 refueling outage established by NRC approval of RNP RR-11, representing a total extension within 6 calendar years of the nominal 10 calendar year interval of ASME Code Case N-729-4. In particular, the tabulation of CGR data for Alloys 690/52/152 (Section 3 of EPRI MRP-375) and review of inspection experience for Alloy 690/52/152 plant components (Section 2 of EPRI 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 EPRI MRP-375.

RNP Unit 2 RVCH Design and Operation The analysis performed by EPRI MRP-375 bounds the design and operation of the RNP replacement RVCH. The RVCH contains forty-seven (47) nozzle penetrations of which forty-five (45) are used for control rod drive mechanisms (CRDMs), and two (2) small diameter penetrations near the center of the RVCH are used for the Reactor Head Vent (RHV) and Reactor Vessel Level Indication System (RVLIS). The Replacement RVCH was manufactured by Mitsubishi and placed in service in October 2005. The replacement RVCH was manufactured as a single forging, which eliminated the center disc and flange circumferential weld in the original RNP RVCH. The replacement RVCH is fabricated from SA-508, Grade 3, Class 1 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 SB-167 (Alloy 690) UNS N06690 and the vent pipe is made from SB-167 (Alloy 690) and SA-312 Type 316. The nozzle J-groove welds utilize ERNiCrFe-7 (UNS N06052) and ENiCrFe-7 (UNS W86152) weld materials.

A preservice volumetric examination of the RNP replacement RVCH J-groove welded CROM, RHV and RVLIS nozzles was performed by Westinghouse prior to installation.

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 7of13 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 ultrasonic examination (UT) responses indicative of planar flaws identified during the volumetric examinations. Additionally, a preservice eddy current examination of the CROM, RHV and RVLIS nozzle welds was performed.

There were no responses indicative of planar flaws identified during the eddy current examinations.

Bare metal visual examinations were performed in 2010 and 2015 in accordance with ASME Code Case N-729-1, Table 1, Item B4.30. These visual examinations were performed by visual examination (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.

The EPRI MRP-375 analyses assume a reactor vessel head operating temperature of 613°F to bound the known RV head temperatures of all PWRs currently operating. The nominal operating hot leg temperature for RNP is 604.1°F. Core bypass flow is expected to reduce the upper head temperature by approximately 4.35°F, which would result in an average RVCH temperature of approximately 599.75°F. Based on this, the RNP RVCH average operating temperature (which is the measure of temperature relevant to potential PWSCC degradation) is bounded by the evaluation results in EPRI MRP-375, which assumes 613°F for its main deterministic and probabilistic calculations.

FOi Implied by Inspection Period Duke Energy has also assessed the representative Alloy 690/52/152 FOi for the requested RNP extension period for comparison with the full set of laboratory CGR data.

The requirements in ASME Code Case N-729-4 are based upon conclusions reached that a head with Alloy 600 nozzles and operating at a temperature of 605°F is safe to operate up to 2 years (one 24 month operating cycle) between volumetric or surface examinations. The same period for Alloy 690 RVCHs in N-729-4 is 10 years which represents a factor of 5 over the Alloy 600 RVCHs. A simple extension of that improvement factor to 15 years would be a factor of 7.5 for the proposed period between volumetric or surface examinations for RNP.

However, the RVCH operating temperature assumed in the technical basis for heads with Alloy 600 nozzles (References 5, 7, & 8) for ASME Code Case N-729-4 was 605°F, compared to an operating temperature of 599.75°F for RNP. Code Case N-729-4 addresses the effect of differences in operating temperature on the required volumetric or surface re-examination interval for heads with Alloy 600 nozzles 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-729-4 as conditioned by 1 OCFR50.55a, 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. As discussed in

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 8 of 13 th~ technical basis documents for heads with Alloy 600 nozzles, effective time for crack growth is the principal basis for setting the appropriate re-examination interval to detect any PWSCC in a timely fashion. U.S. PWR inspection experience for heads with Alloy 600 nozzles has confirmed that the RIV = 2.25 interval results in a suitably conservative inspection program.

There have been no reports of nozzle leakage or of safety significant circumferential cracking subsequent to the time that the Alloy 600 nozzles in a head were first examined by non-visual inservice non-destructive examination for plants conforming to the 2.25 RIV Interval (References 9 & 10).

The representative RNP RVCH operating temperature of 599.75°F would result in an RIV temperature adjustment factor of 0.994 (versus the reference temperature of 600°F) using the activation energy of 31 kcal/mol for crack growth of ASME Code Case N-729-4. Laboratory PWSCC crack growth rate testing for Alloy 690 wrought material by multiple investigators (References 11, 12, & 13) has shown thermal activation energy values comparable to the standard activation energy applied to model growth of Alloys 600/82/182 (31 kcal/mol or 130 kJ/mol). Thus, it is appropriate to apply this standard activation energy for modeling crack growth of Alloy 690/52/152 plant components.

Conservatively assuming that the EFPVs of operation accumulated at RNP since RVCH replacement is equal to the calendar years since replacement, the RIV for the requested extended period at RNP would be (0.994) x (15 years)= 14.91 RIV. The FOi implied by this RIV value for RNP is (14.91)/(2.25) = 6.6 FOi.

Supplemental Assessment of Laboratory PWSCC Crack Growth Rate Data Compiled for Alloys 690, 52, and 152 The enclosure provides further support for the requested alternative inspection interval based on the available laboratory PWSCC crack growth rate data and the FOi approach. The enclosure provides responses to the requests for additional information (RAI) that NRC has transmitted to other licensees in the context of similar relief requests, including the prior RNP relief request. Specifically, the enclosure describes the materials tested for data points within a factor of 12 below the MRP-55 (Ref. 14) and MRP-115 (Ref. 15) crack growth rate curves for the 751h percentile of material variability. It is concluded that the available crack growth rate data support the use of an FOi of at least 12 when setting the volumetric or surface examination interval for heads with Alloy 690 nozzles and Alloy 52/152 attachment welds. The FOi of 6.6 associated with the proposed alternative for the RNP RVCH is conservatively smaller than an FOi of 12.

Moreover, per the enclosure, the crack growth rate data do not show any susceptibility concerns specific to the nozzle or weld materials used in the RNP Unit 2 replacement head. There are not any relevant similarities between (a) the data points within a factor of 6.6 below the MRP-55 (Ref. 14) and MRP-115 (Ref. 15) lines in the figures of the enclosure that deterministically represent the 75th percentile of material variability and (b) the associated nozzles and weld material used in the current RVCH at RNP Unit 2.

The H.B. Robinson Unit 2 replacement RV head was procured from Mitsubishi Heavy Industries, LTD. The Alloy 690 nozzle material used was supplied by Sumitomo Metal Industries. The ASTM/ASME material specification for the nozzle material is SB-167

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 9of13 UNS N06690, and the head was constructed to ASME Section Ill, 1998 Edition, 2000 Addenda. The weld metals used were Unified Numbering System (UNS) N06052 I American Welding Society (AWS) ERNiCrFe-7 (Alloy 52-Gas Tungsten Arc Welding (GTAW)) and UNS W86152 I AWS ENiCrFe-7 (Alloy 152-Shielded Metal Arc Welding (SMAW)). The RV head manufacturer (i.e., welding organization) was Mitsubishi Heavy Industries, LTD. None of the Alloy 690 or Alloy 690 HAZ data points in the enclosure (which are limited to data with at most 20% added cold work) were produced for specimens of CROM nozzle material that was supplied by Sumitomo. Also, there are no other similarities that indicate any specific concern for elevated PWSCC susceptibility of the head nozzles at RNP Unit 2 in comparison to other heads with Alloy 690 nozzles.

Furthermore, the variability among test welds with respect to PWSCC crack growth susceptibility reflects a combination of how the weld was made (welding procedure, weld design, degree of constraint, etc.) and perhaps the material variability in the weld consumable (e.g., composition). The test weld specimens in the enclosure should not be associated with particular fabrication categories of replacement heads because the test welds used to produce the crack growth rate data compiled in MRP-375 are not identified with any particular fabricator of replacement RV heads. Based on the enclosure and the additional information and discussion provided above, it is concluded that the crack growth rate data do not indicate any susceptibility concerns specific to the nozzle or weld materials specific to the H.B. Robinson Unit 2 replacement head.

Conclusions Duke Energy believes that the Alloy 690 nozzle base and Alloy 52/152 weld materials used in the RNP replacement RVCH provide for a clearly superior reactor coolant system pressure boundary where the potential for PWSCC has been shown by analysis and by years of positive industry experience to be remote. This is further supported by visual examination of the RNP RVCH in 201 O and 2015 and the volumetric examinations performed by other Westinghouse designed plants during their nominal 10-year volumetric or surface examination under similar operating conditions and which did not reveal PWSCC.

The minimum FOi implied by the requested extension period represents a level of reduction in PWSCC crack growth rate versus that for Alloys 600/821182 that is completely bounded by the laboratory data compiled in EPRI MRP-375 when accounting for heat-to-heat variability of Alloy 600 and weld-to-weld variability of Alloys 82/182/132. 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. Furthermore, the visual examinations and acceptance criteria as required by Item B4.30 of Table 1 of ASME Code Case N-729-4 are not affected by this request, and these visual examinations will continue to be performed on a frequency of every third refueling outage or 5 calendar years, whichever is less. The requirement for visual examination of the outer surface of the head for evidence of leakage supplements the volumetric and/or surface examination requirement and conservatively addresses the potential concern for boric acid corrosion of the low-alloy steel head due to PWSCC leakage.

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 10 of 13 The proposed revised interval will continue to provide reasonable assurance of structural integrity. For the reasons noted above, it is requested that the NRC authorize this proposed alternative in accordance with 1 OCFR50.55a(z)(1) as the alternative provides an acceptable level of quality and safety.

6.

Duration of Proposed Alternative The proposed alternative is requested for the duration up to and including the 32"d RNP refueling outage that is scheduled to commence in September of 2020.

7.

Precedents There have been submittals from multiple plants to request an alternative from the nominal 10-year interval of ASME Code Cases N-729-1 and N-729-4 for volumetric or surface examinations of heads with Alloy 690 nozzles. The prior RNP Unit 2 request for relief and requests from other plants-including the associated status at the time of submittal of this request-are shown below. Alternative intervals greater than 15 years have previously been granted in order to align with scheduled refueling outages. The approved Calvert Cliffs Units 1 & 2 alternative (noted in the table below) permitted an inspection interval not to exceed 16 years in order to align with scheduled refueling outages. Furthermore, alternatives have been approved at three sites that further extend the inspection interval from an initial approved alternative to a total interval of up to 15.5 years at Arkansas Nuclear One 1, Beaver Valley 1, and St. Lucie 1.

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 11of13 NRC ADAMS Accession No.

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

Arkansas Nuclear One, Unit 1 ML14118A477 ML14258A020 ML14275A460 (2.5-year Extension)

Arkansas Nuclear One, Unit 1 ML16173A297 None None (5.5-vear Extension)

Beaver Valley, Unit 1 ML14290A140 None None (2-year Extension)

Beaver Valley, Unit 1 ML17044A440 None None (5-vear Extension)

Calvert Cliffs, ML15201A067 None None Units 1 & 2 Comanche Peak, ML15120A038 None None Unit 1 D.C. Cook, ML15023A038 None None Units 1 & 2 J.M. Farley, ML15111A387 None None Unit 2 North Anna, ML14283A044 None None Unit2 Prairie Island, ML14258A124 ML15030A008 ML15036A252 Units 1 and 2 Palo Verde, ML17101A678 Units 1, 2, and 3 H.B. Robinson, ML14251A014 ML14294A587 ML14325A693 Unit2

Salem, ML15098A426 None None Unit 1 St. Lucie, Unit 1 ML14206A939 ML14251A222 ML14273A011 (3-year Extension)

St. Lucie, Unit 1 ML17045A357 None None (5.5-year Extension)

St. Lucie, Unit 2 ML16076A431 None None

8.

References NRC Safety Evaluation ML14330A207 ML17018A283 ML14363A409 ML17222A162 ML15327A367 ML15259A004 ML15156A906 ML15104A192 ML15091A687 ML15125A361 ML15021A354 ML15349A956 ML14339A163 ML17219A174 ML16292A761

1.

ASME Code Case N-729-4, "Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds,Section XI, Division 1," Approved June 22, 2012 Status Accepted Accepted Accepted Accepted Accepted Accepted Accepted Accepted Accepted Accepted Under NRC Review Accepted Accepted Accepted Accepted Accepted

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 12 of 13

2.

U.S. NRG, "H. B. Robinson Steam Electric Plant, Unit No. 2 - Relief Request (RR-

11) - Relief from Volumetric/Surface Examination Frequency Requirements of ASME Code Case N-729-1 (TAC No. MF4801)" Safety Evaluation Report, February 2015 (ML15021A354).

3.

EPRI MRP-375, "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles," Report No. 3002002441, February 2014 (freely available at www.epri.com).

4.

EPRI MRP-111, "Resistance to Primary Water Stress Corrosion Cracking of Alloys 690, 52, and 152 in Pressurized Water Reactors," Report No.1009801, March 2004 (ML041680546).

5.

EPRI MRP-110, "Reactor Vessel Closure Head Penetration Safety Assessment for U.S. PWR Plants," Report No.1009807, April 2004 (ML041680506).

6.

EPRI MRP-335 (Rev.1 ), "Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement," Report No. 3002000073, January 2013 (freely available at www.epri.com).

7.

EPRI MRP-117, "Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PWR Plants," Report No. 1007830, December 2004 (ML043570129).

8.

EPRI MRP-105, "Probabilistic Fracture Mechanics Analysis of PWR Reactor Pressure Vessel Top Head Nozzle Cracking," Report No. 1007834, April 2004 (ML041680489).

9.

EPRI MRP-395, "Reevaluation of Technical Basis for Inspection of Alloy 600 PWR Reactor Vessel Top Head Nozzles," Report No. 3002003099, September 2014 (freely available at www.epri.com).

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

11. U.S. NRG, "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).

12. EPRI MRP-237 (Rev.2), "Resistance of Alloys 690, 152, and 52 to Primary Water Stress Corrosion Cracking: Summary of Findings Between 2008 and 2012 from Completed and Ongoing Test Programs," Report No. 3002000190, April 2013 (freely available at www.epri.com).

13. 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.

14. EPRI MRP-55, Revision 1, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Thick-Wall Alloy 600 Materials," Report No.

1006695, November 2002 (freely available at www.epri.com).

United States Nuclear Regulatory Commission Attachment to Serial: RNP-RA/17-0043 Page 13of13

15. EPRI MRP-115, "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds," Report No.

1006696, November 2004 (freely available at www.epri.com).

United States Nuclear Regulatory Commission Enclosure to Serial: RNP-RA/17-0043 33 Pages (including cover sheet)

Dominion Engineering, Inc. -Technical Note (TN-5696-00-02, Rev. 0)

~

Dominion fnvineerin:Y TECHNIC~L 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 Principal Investigators G. White K. Fuhr Prepared for TN-5696-00-02 Revision 0 March 2015 Electric Power Research Institute, Inc.

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

Dominion En~ineerin~, Inc.

TN-5696-00-02, Rev. 0 RE10RD OF REVISIONS Prepared by Checked by Reviewed by Approved by Rev.

Description Date Date Date Date 0

Original Issue 3Cf74 11.11-~-J.f G./f. a"'~~

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K. J. Fuhr M. Burkardt G. A. White G. A. White Associate Engineer

• Associate Engineer Principal Engineer Principal Engineer The last revision number to reflect any changes for each section of the 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.

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Page Mod.

1 INTRODUCTION.................................................................................................................. 1 0

2 DISCUSSION OF DATA POINTS FROM MRP-375 [2].......................................................... 3 0

2.1 Data Points Above a Hypothetical 12.0 Factor of Improvement Line in Figure 3-1, 3-3, and 3-5 of MRP-375..................................................................... 3 0

2.2 Data Most Directly Applicable to Plant Conditions................................................. 6 0

2.3 Data Specific to Argonne National Laboratory (ANL) and Pacific Northwest National Laboratory (PNNL)................................................................. 8 0

2.4 Data for Alloy 690 Wrought Material Including Added Cold Work up to 20% for CROM Nozzle and Bar Material Product Forms....................................... 8 0

2.5 Conclusion............................................................................................................ 9 0

3 POTENTIAL IMPLICATIONS OF SPECIFIC CATEGORIES OF NOZZLE AND WELD MATERIALS........................................................................................................................ 9 0

3.1 Potential Similarities for Laboratory Specimen Material Exhibiting a Deterministic Factor Less than 12.0...................................................................... 9 0

3.2 Potential Implications.......................................................................................... 10 0

4 REFERENCES.................................................................................................................... 12 0

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TN-5696-00-02, Rev. 0 LIST OF FIGURES I Last Rev.

Page Mod.

Figure 1.

Plot of Crack Growth Rate (da/dt) versus Stress Intensity Factor (K1) for Alloy 690 Data from Plate Material Tested by CIEMAT............................................... 14 0

Figure 2.

Plot of da/dt versus K1 for Alloy 690 Data from Heat WP787.............................. 14 0

Figure 3.

Plot of da/dt versus K1 for Alloy 690 Data from Heat WP142.............................. 15 0

Figure 4.

Plot of da/dt versus K1 for Alloy 690 HAZ Data from Heat WP142...................... 15 0

Figure 5.

Plot of da/dt versus K1 for Alloy 690 HAZ Data from Plate Material Tested by CIEMAT.............................................................................................................. 16 0

Figure 6.

Plot of da/dt versus K1 for Alloy 152 Data from Heat WC83F8............................ 16 0

Figure 7.

Plot of da/dt versus K1 for Alloy 152 Data from Heat WC04F6............................ 17 0

Figure 8.

Plot of da/dt versus K1 for Alloy 690 Data from All Laboratories, :5 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, :510% Cold Work, Constant Load or K1................................... 18 0

Figure 10.

Plot of da/dt versus K1 for Alloy 690 HAZ Data from All Laboratories, :510%

Cold Work, Constant Load or K1......................................................................... 19 0

Figure 11.

Cumulative Distribution Function of Adjusted da/dt for Alloy 690 HAZ Data from All Laboratories, S 10% Cold Work, Constant Load or Ki........................... 19 0

Figure 12.

Plot of da/dt versus K1 for Alloy 52/152 Data from All Laboratories, :5 10%

Cold Work, Constant Load or K1......................................................................... 20 0

Figure 13.

Cumulative Distribution Function of Adjusted da/dt for Alloy 52/152 Data from All Laboratories, s 10% Cold Work, Constant Load or K1........................... 20 0

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]; s 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]; s 22% Cold Work and Constant Load/Ki............................................................................................................... 22 0

Figure 17.

Plot of da/dt versus Ki for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference [17]; s 22% Cold Work............................................ 23 0

Figure 18.

Cumulative Distribution Function of Adjusted da/dt Alloy 690 HAZ Data Produced by ANL and PNNL [17]; s 22% Cold Work and Constant Load/Ki...... 23 0

Figure 19.

Plot of da/dt versus Ki for Alloy 52/152 Data Produced by ANL and PNNL iv

Dominion En~ineerin~, Inc.

TN-5696-00-02, Rev. 0 Last Rev.

Page and Available in References [17] and [18]; s 22% Cold Work............................ 24 Mod.

0 Figure 20.

Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data Produced by ANL andlPNNL ([17] and [18]); s 22% Cold Work and Constant Load/Ki............................................................................................................... 24 0

Figure 21.

Plot of da/dt versus K1 for Alloy 690 Data from All Laboratories, > 10 & s 20% Cold Work, CROM 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, s 20% Cold Work, CROM and Bar Material, Constant Load or Ki............ 25 0

Figure 23.

Plot of da/dt versus K1 for Alloy 52/152 Data from All Laboratories,> 10 & s 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, s 20% Cold Work, Constant Load or Ki........................................ 26 0

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Dominion En~ineerin~. Inc.

TN-5696-00-02, Rev. 0 ACRONYMS ANL ASME AWS BWC CEDM CGR CIEMAT CROM CT DEi EPRl FOi GE-GRC GTAW HAZ ICI K

MRP NRC PNNL PPU PWR PWSCC RIY RV RVCH UNS Argonne National Laboratory American Society of Mechanical Engineers American Welding Society Babcock & Wilcox Canada Control Element Drive Mechanism Crack Growth Rate Centro de Investigaciones Energeticas, Medioambientales y Tecnol6gicas Control Rod Drive Mechanism Compact Tension Dominion Engineering, Inc.

Electric Power Research Institute Factor of Improvement General Electric Global Research Center Gas Tungsten Arc Welding Heat Affected Zone In-Core Instrumentation Stress Intensity Factor Materials Reliability Program Nuclear Regulatory Commission Pacific Northwest National Laboratory Partial Periodic Unloading Pressurized Water Reactor Primary Water Stress Corrosion Cracking Re-Inspection Year Reactor Vessel Reactor Pressure Closure Head Unified Numbering System vi

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

INTRODUCTION I

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 CG Rs for specific categories of nozzle and weld materials.

Per ASME Code Case N-729-1 [3], the volumtric inspection interval for Alloy 600 RV head nozzles is based on operating time adjusted for operating temperature using the temperature sensitivity for PWSCC crack growth. The normalized operating time between inspections, called the Re-Inspection Years (RIY) parameter, represents the potential for crack growth between successive volumtric examinations. Thus, the FOI for Alloys 690/521152 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 of 20 years for Alloy 690 RV head nozzles operating at 613°F. 1 A temperature of 613°F is expected to bound the head operating temperature for the U.S. pressurized water reactor (PWR) fleet.

As discussed in Section 3 of Electric Power Research Institute (EPRI) Materials Reliability Program (MRP) report MRP-375 [2], a conservative approach was taken in MRP-375 to develop the factor of improvement (FOI) values describing the primary water stress corrosion cracking (PWSCC) crack growth rates applicable to Alloy 690 reactor vessel (RV) top head penetration nozzles. The crack growth rate data points presented in Figures 3-1, 3-3, and 3-5 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 of613°F, the re-inspection year (RIY) parameter for a requested examination interval of 20 years is compared with the N-729-1 interval for Alloy 600 nozzles ofRIY = 2.25. The representative head operating temperatures of613°F corresponds to an RIY temperature adjustment factor of 1.38 (versus the reference temperature of 600°F) using the activation energy of 31 kcal/mo! (130 kJ/mol) for crack growth of ASME Code Case N-729-1. Conservatively assuming that the effective full power years (EFPY) ofoperation accumulated since RV top head replacement is equal to 98% of the calendar years since replacement, the RIY for a requested extended period of20 years would be (1.38)(19.6) =

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

Dominion fn~ineerin~, Inc.

TN-5696-00-02, Rev. 0 laboratories. No screening process was applied to the data on the basis of test characteristics such as minimum required crack extension or minimum required extent of transition along the crack front to intergranular cracking. Instead, an inclusive process was applied to conservatively as~ess the factors of improvement apparent in the data for spjcimens with less than 10 percent adlied 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 600/82/182 treats the full variability in both original and replacement alloys, rather than comparing the variability of the replacement alloy against a conservative mean (75th percentile) growth rate for the original alloys. By considering the cumulative distributions, a fuller perspective of the improved resistance of Alloys 690/521152 emerges where over 70% of the data in each of Figures 3-2, 3-4, and 3-6 of MRP-3 75 indicate a factor of improvement beyond 20 and all of the data3 correspond to a factor of improvement of 12 or greater.

It is emphasized that the deterministic MRP-55 and MRP-115 crack growth rate equations were developed not to describe bounding crack growth rate behavior but rather reflect 75th percentile values of the variability in crack growth rate due to material variability. Twenty-five percent of the material heats (MRP-55) and test welds (MRP-115) assessed in these reports on average showed crack growth rates exceeding the deterministic equation values. Thus, the most appropriate FOI comparisons are made on a statistical basis (e.g., Figures 3-2, 3-4, and 3-6 of MRP-3 75). 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 Figures 3-2, 3-4, and 3-6 ofMRP-375 show cumulative distribution functions of the variability in crack growth rate normalized for temperature and crack loading (i.e., stress intensity factor). Each ordinate value in the plots shows the fraction of data falling below the corresponding normalized crack growth rate. Thus, the cumulative distribution function has the benefit of illustrating the variability in crack growth rate data for a standard set of conditions.

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

2

Dominion En~ineerin~. Inc.

TN-5696-00-02, Rev. 0 growth rates being many times higher than the deterministic 75th percentile values per MRP-55 and MRP-115. Nevertheless, as described below, the large majority of the data points for the conditions directly relevant to plant conditions (e.g., constant load conditions) are located more than a factof of 12.0 below the deterministic (75th percentile) MRP-551and 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 75th percentile curve, out of a total of 75 points shown in Figure 3-1 of MRP-375.

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

Two of the points tested by CIEMAT are from specimen 9ARB1, comprised of Alloy 690 plate material, loaded to 37 MPa(m)0*5, and tested at 340°C and 15 cc H2/kg H20

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

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

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

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

U.S. PWRs operate with a dissolved hydrogen concentration per EPRI guidelines in the range of 25-50 cc/kg for Mode 1 operation. Testing at 15 cc/kg results in accelerated crack growth rates versus that for normal primary water due to the proximity of the Ni-NiO equilibrium line [2].

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

Dominion [n~ineerin~, Inc TN-5696-00-02, Rev. 0 material. CRDM and CEDM nozzles in U.S. PWRs are fabricated from extruded pipe or bar stock material. Note that term CRDM nozzle is used henceforth to refer to both CRDM and CEDM nozzles (CEDM is the terminology used by plants designed by Combustion Engineering).

The wide variability in crack growth rate within even the samtj testing segment indi ates that significant experimental variability exists. ThusJ 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)05-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)0*5), and the remaining six data are from specimens 9T5 or 9T6 (loaded to roughly 27 MPa(m)0*5). The results for 9Tl and 9T2 are contained in Reference [9]; the final data for 9T5 and 9T6 are contained in EPRI MRP-340, but have not been openly published. As discussed later in Section 2.4, the addition of cold work may result in a material that is substantially more susceptible than the as-received material. The extent of transition along the crack front to intergranular cracking for these data was extremely low (::;

10%) for the ten points from specimens tested at constant temperature. A plot of crack growth rate versus K for the Alloy 690 data from MRP-375 for heat WP787 is provided here as Figure 2. As in Figure l, there is significant growth rate variability within the data for the same heat of material. The median for the CIEMA T specimens is more than a factor of 12 below the MRP-55 curve. Additionally, the Pacific Northwest National Laboratory (PNNL) data indicate that the specific laboratory that produces the data can significantly influence the reported growth rate, such that there is a substantial possibility that a small number of reported data points with relatively high crack growth rates from a single laboratory are not characteristic of the true susceptibility of a specific heat of Alloy 690 material.

The three ANL data points are for CT specimens C690-CR-1 and C690-LR-2, comprised of Valinox heat number WP 142 CRDM nozzle material that were not cold worked and were tested at 21 to 24 MPa(m)0*5, 320°C, and 23 cc H2/kg H20 [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-3 7 5. Figure 3-3 shows the complete set of data points compiled for Alloy 690 heat affected zone (HAZ) specimens at the time MRP-375 was completed by the PWSCC Expert Panel that was organized by EPRI. The following points are within a factor of 12.0 below the MRP-55 deterministic crack growth rate for Alloy 600:

4

Dominion fn~ineerin~, Inc.

TN-5696-00-02, Rev. 0 There are eight points within a factor of 12.0 below the MRP-55 75th percentile curve, out of a total of 34 points shown in Figure 3-3 of MRP-3 75. 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 pf Valinox CROM nozzl~ 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)05) [1 ll, and the other point is from specimen CF690-CR-4 (loaded to roughly 22 MPa(m)0* ) [12]. A plot of crack growth rate versus K for all the Alloy 690 HAZ data from MRP-3 75 for heat WP142 is provided here as Figure 4. As discussed below, PPU conditions-under which five of these six points were obtained-may result in accelerated crack growth relative to plant conditions.

The remaining two points are from CIEMAT testing of specimens 19ARH1 and 19ARH2, comprised of welded Alloy 690 plate material, tested at 340°C and 15 cc H2/kg H20, and loaded to roughly 37 MPa(m)0*5 [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 75th percentile curve, out of a total of212 points shown in Figure 3-5 ofMRP-375. Five of these points are not relevant to PWR conditions and should not be considered further, as discussed in the following bullets.

One of these points is from PNNL testing of the dilution zone of a dissimilar metal weld between 152M (Special Metals heat WC83F8) and carbon steel, tested at 360°C and 25 cc H2/kg H20 [13]. This material condition is not applicable to the wetted surfaces of CRDM nozzle J-groove welds because the dilution zone where Alloy 52/152 contacts the low-alloy steel RV head is below the stainless steel cladding. A plot of crack growth rate versus K for the Alloy 152 data from MRP-375 for heat WC83F8 is provided here as Figure 6.

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

The remaining 14 data points represent four specimens from Alloy 152 weld material (Special Metals heat WC04F6) that were tested by ANL at 320°C and 23 cc H2/kg H20 ([15] and [10]). Ten of these* points are for specimen A152-TS-5 at loads of about 28, 32, and 48 MPa(m)°-5 [ 14]. The other four points were obtained at loads of 5

Dominion fn~ineerin~. Inc.

TN-5696-00-02, Rev. 0 27 MPa(m)0*5 for specimen N152-TS-1and30 MPa(m)0*5 for specimens A152-TS-2 and A152-TS-4. The Alloy 152 specimens all came from welded plate material. A plot of crack growth rate versus K for the Alloy 15 2 data from MRP-3 7 5 for heat WC04F6 is provided here as Figure 7. All but three of these points were for PPU conditions, whf ch may result in accelerated crack growth rates that are n9t directly representative ~f plant conditions. Figure 7 shows a very large variabilitt in the crack growth rate reported by different laboratories for this heat of Alloy 152 weld material.

Roughly one third the ANL data (specimen N152-TS-l), all of the General Electric Global Research Center (GE-GRC) data, and all the PNNL data for this heat are for specimens from a single weld made by ANL [16], illustrating the role of experimental variability. A small number of elevated data points for a weld produced by a single laboratory may not be representative of the true material susceptibility.

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

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

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

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

Figure 10 for Alloy 690 HAZ.

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

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

Figure 12 for Alloys 52/152.

Only three of 83 points are within a factor of 12.0 below the MRP-115 deterministic crack growth rate for Alloy 182.

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

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

Dominion fn~ineerin~. Inc.

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

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

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

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

The Alloy 52/1 52 weld metal data shown in Figure 3-5 and Figure 3-6 of MRP-375 include data reflecting a range of weld dilution levels. The data presented in Figure 12 and Figure 13 exclude the weld dilution data points because of the limited number of data points available, the variability in results, and the limited area of continuous weld dilution for potential flaws to grow through. The weld dilution data are not reflective of the full chromium content of Alloy 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 of MRP-375.

The limited number of remaining points in Figure 8 and Figure 12 that lie within a factor of 12.0 below the deterministic MRP-55 and MRP-115 lines represent the upper end of material and/or experimental variability. Figure 9, Figure 11, and Figure 13 consider the variability in crack growth rate among different heats/welds of Alloys 600/82/182 and compare this against the full variability of the Alloy 690/52/152 data most applicable to plant conditions. The lack of any 7

Dominion fn~ineerin~. Inc.

TN-5696-00-02, Rev. 0 points within a factor of 12 when accounting for variability in Alloy 600/82/ 182 crack growth rates supports a reexamination interval longer than the requested interval corresponding to an FOi of 12.0. The volumetric or surface inspection interval for heads with Alloy 600 nozzles reflects consideration of crack growth rates 9n a statistical basis, with crack growth rates often higher than that given by the deterministic equations of MRP-55 and MRP-115.

2.3 Data Specific to Argonne National Laboratory (ANL) and Pacific Northwest National Laboratory (PNNL)

The U.S. NRC is most familiar with the crack growth data for Alloys 690/521152 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 52/152 material for the purpose of including the limited number (i.e.,

five) of weld metal data points generated for added cold work conditions are shown in Figure 23 4 The data in Reference [16] are augmented by the crack growth rate data for Alloys 52/152 produced by PNNL and previously published in an NRC NUREG contractor report [17]. While these PNNL data are shown graphically in of Reference [16], the enclosures of tabular data in this NRC document omitted all of the PNNL data for Alloys 52/152. It is also noted that contrary to the enclosure titles of Reference [16], Enclosure 2 contains the PNNL tabular data, and Enclosure 4 contains the ANL tabular data.

8

Dominion fn~ineerin~. Inc.

TN-5696-00-02, Rev. 0 and Figure 24. Added cold work for weld metals is not directly relevant to plant material conditions.

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

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

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

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

9

Dominion [n~ineerin~, Inc.

TN-5696-00-02, Rev. 0 Figure 3-1 of MRP-375 [2]. The only Alloy 690 CRDM material for which crack growth rate data were available at added cold work of less than 10% (the threshold for inclusion in Figure 3-1 ofMRP-375) was supplied by Valinox Nucleaire. The few data using CRDM material from other suppliers were obtained at cold works of 20% or higher and were not included in the assessment. The data do not indicate any porrelation between material I supplier and susceptibility to crack growth rate. FourteenJ 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 of PPU for eight of these 14 data), this similarity in no way indicates any specific concern for elevated PWSCC susceptibility of the head nozzle material provided by any one supplier.

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

Figure 3-5 of MRP-375 [2]. There are no relevant similarities between (a) the Alloy 52 and 152 data points above a crack growth rate 12.0 times lower than the MRP-115 [2]

Alloy 182 deterministic crack growth rate in Figure 3-5 of MRP-3 7 5 and (b) the Alloy 521152 weld material used in any particular categories of replacement heads. The variability among test welds with respect to PWSCC crack growth susceptibility reflects a combination of how the weld was made (welding procedure, weld design, degree of constraint, etc.) and perhaps the material variability in the weld consumable (e.g.,

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

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

10

Dominion [n~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 l 0% added cold work are supplied by Valinox. Consequently, there is no basis to suggest material from any one supplier is more susceptible than that from another based on the presence or absence of data points within a given factor of the determi11istic crack growth rate curve from MRP-55.

The datal points showing the highest crack growth rates for the testld 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 690/52/152 than for Alloys 600/82/182/132 considering the greater testing challenges associated with the more resistant replacement alloys.

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

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

Laboratory and plant experience show that Alloys 690/52/152 are substantially more resistant to PWSCC initiation than Alloys 600/82/182 [2]. PWSCC has not been shown to be an active degradation mode for Alloys 690/521152 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 I A WS ERNiCrFe-7), Alloy 52M (UNS N06054 I AWS ERNiCrFe-7A), Alloy 52MSS (UNS N06055 I AWS ERNiCrFe-13), Alloy 52i (AWS ERNiCrFe-15), Alloy 152 (UNS W86152 I AWS ENiCrFe-7), and Alloy 152M (UNS W86 l 52 I A WS ENiCrFe-7). Considering the overall set of available crack growth rate data for the various variants of Alloy 52 and 152, there is no basis for concluding at this time any significant difference in the average behavior between the Alloy 52 and Alloy 152 variants in use at U.S. PWR RV heads with Alloy 690 nozzles.

In addition, it should be recognized that PWSCC of Alloy 690 RV head penetration nozzles or their Alloy 52/152 attachment welds is not an active degradation mode. Thus, it is premature to single out individual materials or fabrication categories of heads with Alloy 690 nozzles for additional scrutiny on the basis of subsets of laboratory crack growth rate data. In the case of 11

Dominion En~ineerin~, Inc.

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

Based on the additional! information and discussion provided above, it is conclud~ 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.epri.com]

2.

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]

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/or Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375), EPRI, Palo Alto, CA: 2014. 3002002441. [freely available at www.epri.com]

5.

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

6.

Materials Reliability Program: Reactor Vessel Closure Head Penetration Safety Assessment for U.S. PWR Plants (MRP-1 JONP), EPRI, Palo Alto, CA: 2004. 1009807-NP.

[ML041680506]

7.

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

2004. 1007834. [ML041680489]

8.

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

9.

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

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

12

Dominion fn~ineerin~, Inc.

TN-5696-00-02, Rev. 0 Presented at: Alloy 6901152152 Research Collaboration Meeting, Tampa, FL, November 29-December 3, 2011.

10. Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld in Simulated PWR Environment - 2009, NUREG/CR-7137, June 2012.

11.

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

12.

B. Alexandreanu, Y. Chen, K. Natesan and B. Shack, "Update on SCC CGR Tests on Alloys 690/52/1 52 at ANL - June 2011," Presented at: US NRCIEPRI Meeting, June 6-7, 2011. [MLI 11661946]

13.

M. Toloczko, M. Olszta, N. Overman, and S. Bruemmer, "Stress Corrosion Crack Growth Response For Alloy 152/52 Dissimilar Metal Welds In PWR Primary Water," 16th International Conference on Environmental Degradation 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-7103, Vol. 2, April 2012.

19.

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

13

Dominion En~ineerin~, Inc.

TN-5696-00-02, Rev. 0 Data from Individual Heats 1.E-09....-=====-------------------------..

I

..t. CIEMATI MRP-55 i.-----1 1 __ =-==---=---=----==-~==-~~~~;;:::::::::::::~c~u~~~e/~

1 L__~I 1.E-10 ~-

~

~

11.E-11

(!)

.:.: e u

1.E-12 6 --

t,. -- --~

PPU data are represented with open s rnbols Data are adjusted for temperature (325°C).

Q = 130 kJ/mol

1. E-13

+--'-~-'-+....L.....L....L.....1.-+-'-'-'--L..-t-'-'-'--+-L......J.....L......J.....+-'--.._._-'-+-.............L....L.-+-'-~-'-+....L.....L....L.....l.-+-'-'--'--L..-t 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa'1m)

Figure 1.

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

• PNNL MRP-55 Cu~e/1

~

1.E-10 6f:P.

.s co t,.aA tf>t,.

-~

c:::

.r:;

A----

3:

0

~ 6

(!)

PPU data are u

/

11! ~

represented with co u

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

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

Figure 2.

Plot of da/dt versus K1 for Alloy 690 Data from Heat WP787 14

Dominion [n~ineerin~, Inc.

TN-5696-00-02, Rev. 0 1.E-09 1--=== :::;-----=-- = -=-=-=-==----::::- =-=- ::-::-==-""""-=-::--===-="'-:--:-:""""-=--=-=-==-=--=-=--=-=-=-1 OANL

• PNNL

--~

= ---~--

___---------~ !

0 q,,._--

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PPU data are

/

0 represented with

_ L open s mbols I

Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 1.E-13 _.,_............. """""""'_._..................._.~__.__.,....._............. '-0-._._._.__.._._~..._._..._._+-'-..._...-'-+..................... -+-'"~__.__.

10 15 20 25 30 35 40 45 Stress Intensity Factor (MPa-'m)

Figure 3.

Plot of da/dt versus K1 for Alloy 690 Data from Heat WP142 OANL

~

1.E-10 50 55 60 MRP-55 i-----t Curve/1

~

-- ~

ex oO o

~

~ 1.E-11 +---_,_----------------=--n---f':i!i'-'=..,....-- -

(!)

.:it!

!,,)

E u

1.E-12

/

/

I oO PPU data are represented with o ens mbols Data are adjusted for temperature (325°C).

Q = 130 kJ/mol

1. E-13,.._...__._._--'-+...............__..4-'-_._.__.__............ _..._.'-+-._._._._+-'-.._._.........,............._._....................... -'-+..................... -t-1-........... -'-t 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-'m)

Figure 4.

Plot of da/dt versus K1 for Alloy 690 HAZ Data from Heat WP142 15

Dominion [n~ineerin~, Inc.

TN-5696-00-02, Rev. 0 1.E-09 ~~~----------------------~

I CIEMATI MRP-55~--i 1.E-10

~

__ =-.:.=---~

- ~:::=-===-~ -=-=~-=:-1 Curve/1

~

~

~

~

~

--~

i.... o 1.E-11..,-------==---==--=:-;;

~-~,.._----====-----1

<.?

PPU data are ti

~ _,,.

A represented with Ue

~

o ens mbols 1.E-12 -F-----------------;::==:!::::::~~====~1 I

Data are adjusted for temperature (325°C).

Q = 130 kJ/mol

1. E-13 +-"-............. _.__......................... -+-'-............... """""----+...................... -+-'.....&......J........._t-'-.._._....._..._._..._._........................ _.__......................... -+-'-............... ~

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

Figure 5.

Plot of da/dt versus K1 for Alloy 690 HAZ Data from Plate Material Tested by CIEMAT

~

1.E-10 MRP-115 Curve/1

~ 1.E-11

<.?

.:a:

~

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

Q = 130 kJ/mol 0

1. E-13 +-"-............. _.__.............. _._._-+-'-............... """""----+....................... -+-'.....&......J.._._t-'-......_....._..._._..._._........................ _.__................ _._._-+-'-...._..~

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

Figure 6.

Plot of da/dt versus K1 for Alloy 152 Data from Heat WC83FB 16

Dominion En~ineerin~. Inc.

TN-5696-00-02, Rev. 0 1.E-09 ~-ro~A~N~L=1~==-:

- :=-~-:-;- :::-:- :::;~~-~~-~ _;_~-~;;;;;;;;;;-....-1 o GE-GRC -

MRP-115 -

Curve/1 1.E +PNNL

~

<r ~ ~

~

....,.;_--~ ----

.§.

~

0

'V 6 o_

~

~e 1.E-11 ---=---*--=:....:===,.__-=----~~=--------___,....,.,.------..,.--1 X)9 PPU data are 0

Black-filled ANL data present ti represented with growth rates during the

~

0 en s mbols

~

environmental pre-crack period u

1.E-12 -H========:!..===.-~=-=-=-=--~-=::----l and should not be included.

Data are adjusted for

-* JI temperature (325°C).

Q = 130 kJ/mol B *

1. E-13

+-'-....._._-'--+...._._...._._-+-'-__.__.-'---t__.__.__.__.-+-'._._._,_1--"-,_._z...+-,_,__,_,__+-'-....._._-'--+...._._...._._-+-"__._.~

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

Figure 7.

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

Dominion fn~ineerin~, Inc.

TN-5696-00-02, Rev. 0 Data Most Applicable to Plant Conditions

~

s

.s 1.E-09 1.~-10 1.E-11

~

(.!)

~

(,)

E u

1.E-12 OANL

• Bettis MRP-55 CIEMAT Curve/1 PNNL

!iii

-~

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.,__..=----==----------

-=-~ =---

Q..- ~-

-: ~ --

';;t" 0

6

/

0

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Ii ~

Data are adjusted for temperature (325°C).

Q = 130 kJ/mol

1. E-13 +-'-4......_....._.......,_....._.___.._._........_...................... ~..-._._t-'-.........._"'-T-..._._..._._+-'-..................i...+..................... +-'-............... -'-I Figure 8.

Figure 9.

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

Plot of da/dt versus Kr for Alloy 690 Data from All Laboratories, :s 10% Cold Work, Constant Load or Kr o ANL

• Bettis 60

~CIEMAT *---~-----.,..------1*----------1 PNNL 15.~----1------1--------- 1

~

1.E-12 1.E-11 1.E-1 O 1.E-09 1.E-08 Crack Growth Rate (mis)

Cumulative Distribution Function of Adjusted da/dt for Alloy 690 Data from All Laboratories, :s 10% Cold Work, Constant Load or Kr 18

Dominion fo~ineerin~, Inc.

TN-5696-00-02, Rev. 0 1.E-09 o ANL CIEMAT MRP-55 DGE-GRC Curve/1 Vi' 1.E-10

+ P NL g

.s

--~

IQ c::

~

i 1.E-11 0....

(.!)

~

u

/

~

0 A

(.)

1.E-12

/

0 0

I 0

A ¢ Data are adjusted for El temperature (325°C).

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

Figure 10.

Plot of da/dt versus K1 for Alloy 690 HAZ Data from All Laboratories, S 10% Cold Work, Constant Load or K1 1.0 0.9 0.8

§ 0.7 0

s

@ 0.6 VI

~ 0.5

~ 0.4 E

a o.3 0.2 0.1 0.0 1.E-13 0

0 OANL 0 -----------,,_---~!

0 0

I I

L IF~ =~21 7

1.E-12

/

7 1

1.E-11 1.E-10 Crack Growth Rate (mis)

~CIEMAT DGE-GRC -

PNNL The data points at 1 E-13 were reported as "no growth."

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

Q = 130 kJ/mol K = 30 MPavm 1.E-09 1.E-08 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 1.E-09 lfco~A~N~L=r~-~=-=-=-~--=-=-=-=-:-==

- =-===-::~~;;;;.:::::=---1

~CIEMAT

- DGE-GRC MRP-115 Curve/1 1.E-10 PNNL

~

0 0

0

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0 0

~

~

°i 1.E-11.. --~-~=~---_,.~--------------

• e

~

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u

~

u 1.E-12 1.E-13 10 Data are adjusted for temperature (325°C).

Q = 130 kJ/mol 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa-..Jm)

Figure 12.

Plot of da/dt versus K1 for Alloy 52/152 Data from All Laboratories, s 10% Cold Work, Constant Load or K1 1.0 eP 0 OANL

/

0.9 0

~CIEMAT 0.8 DGE-GRC

~ ~

1 PNNL

§ 0.7 I

• .=:

I I

@ 0.6 Cll I-

~ 0.5 The data points at 1 E-13 I-were reported as "no

~ 0.4 growth."

E a o.3 Data are adjusted for FOi = 12 temperature (325°C) and 0.2 stress intensity factor.

0.1 I

Q = 130 kJ/mol K = 30 MPa-..Jm 0.0 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (mis)

Figure 13.

Cumulative Distribution Function of Adjusted da/dt for Alloy 52/152 Data from All Laboratories, S 10% Cold Work, Constant Load or K1 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 r--=----:=-------=------.,,..----;::======= =:;--;:::====;i 1 _ -- -

10 100 1000 10000 Hold Time (Hours)

Figure 14.

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

Dominion fn~ineerin~, Inc.

TN-5696-00-02, Rev. 0 Compilation of ANL and PNNL Data 1.E-09,-;:::=======::::;-- -----::--::-:':":'"-=--- --=-----==-----,

Box and arrow show the

- ratio between the MRP-55 curve and the data point MRP-55 1---_..,

r~~~~~~~~~---=l_..,.......~F~~---1 Curve/1 1.E-10 ~-

1 1.E-11 11Mll~oQ~11+/-~ - ~ - - -~

C!>

~

.,,.-Q..-

oANLCL ti 0

o OANL PPU b

/ /

PNNL 1.E-12 0--

--::<>-------;::::========.1 1---

Data are adjusted for temperature (325°C).

Q = 130 kJ/mol

1. E-13 +-'-_........-'-+-~........... -+-'-'--'-L-i--'---'-<><X:ll)--L--.L...()-ol><>-..._..""-<>...._._...._._+-'-_,_,_-'-+....._......L......L-+-'~--'-l 10 15 20 25 30 35 40 45 50 55 60 Stress Intensity Factor (MPa~m)

Figure 15.

Plot of da/dt versus K1 for Alloy 690 Data Produced by ANL and PNNL and Available in Reference [17]; S 22% Cold Work 1.0 -r-------------a---,,~__..~---'.=:1-------,

o.9 +---------~e>J---.L----~

/

0.8 +---------

-=--- -------1-----------*

.g o.7 +--------,:c--u---1 OANLCL

s E 0.6 1------1--------1

• PNNL en

~ 0.5

/

The data points at 1 E-13 are

/-

treated as "no growth,"

~ 0.4 consistent with MRP-375.

E 8 0.3 Data are adjusted for 0.2 R-----=IF=O=I==1=2!...__I ----/-------I temperature (325°C) and

/

stress intensity factor.

0.1 - ---7 Q = 130 kJ/mol K = 30 MPavm 0.0 1,1---=....................,. __

~:..........~--'--......... _._............ l--__.__::i::==q=====i:l..I 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 Crack Growth Rate (mis)

Figure 16.

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

Dominion fn~ineerin~. Inc.

TN-5696-00-02, Rev. 0

~

s

~

=

~

t!)

~

u E 1.E-09 -============.,...----------------,

o ANL CL Box and arrow show the 0 ANL PPU ratio between the MRP-55 PNNL curve and the data point MRP-55 ~---i l~~~~=:::::~~~~:::;:~~~~::::::::::::::=~c~u~rve~/~

1 t..=_-=-=-_j 1.E-10

~

,._(

--~

,~~,-..._,...,._, __ - - -

00

/

u 1.E-12 1-~:------::---=------:::-------------;:========;I Data are adjusted for temperature (325°C).

a = 130 kJ/mol 1.E-13 +-'".................................... ~-t-"~""""'-t

.......... ~-o--~'--'-'C>----"-+-....................... _........................ ~-.~

.......... -'-i Figure 17.

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

Plot of da/dt versus K1 for Alloy 690 HAZ Data Produced by ANL and PNNL and Available in Reference (17); S 22% Cold Work 60 1.0.,.------------- --,,,--=---.------:=------i 0.9 +----------.,,..__

.,__ _____,,'----------I

~

0.8 +-----'---------

~

.g 0.7 --~-.-------

'------l--------r--o-A_N_L_CL~

I

~

~ 0.6 +----------1------*f--------1 + PNNL Vi

~

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

/-

treated as "no growth,"

~ 0.4 - ------- -----*

consistent with MRP-375.

E 8 0.3 Data are adjusted for IFOI = 121 temperature (325°C) and 0.2 +------=;==----.r-------I stress intensity factor.

0.1 ----~~-------------*

Q=130kJ/mol K = 30 MPa.Ym 0.0 --=....._._.,... __

i=:::i::._,__'-'-"+ _

_..___.__..........._............ +--___.__.::.:==::q:::::====:::i::il 1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 Crack Growth Rate (mis)

Figure 18.

Cumulative Distribution Function of Adjusted da/dt Alloy 690 HAZ Data Produced by ANL and PNNL (17); s 22% Cold Work and Constant Load/Ki 23

Dominion fn~ineerin~, Inc.

TN-5696-00-02, Rev. 0 1.E-09 Box and arrow show the ratio between the MRP-115 MRP-115 Curve/1 CEJO-~

curve and the data point 1.E-1r Ci) g 0

Cl.I 0

0

-- ao co 0

ei::

~

.c 1.E-11

~

0

(.!)

o ANL CL u

OANL PPU co

+ PNNL CK NUREG u

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

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

Figure 19.

Plot of da/dt versus K1 for Alloy 52/152 Data Produced by ANL and PNNL and Available in References [17] and [18]; s 22% Cold Work 1.0 0.9 0.8 5 0.7

',;l

l

@ 0.6

'lii

~ 0.5

~ 0.4 E 8 0.3 0.2 0.1 0.0

~

4 4

4 OANL CL

• PNNL CK (NUREG) 0 --

/

0 7 0-

/

/

0 1 v

I MRP-1151 t

<FOi = 1)

I I

I I

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

L Data are adjusted for I

temperature (325°C) and IFOI = 121 I

stress intensity factor.

Q = 130 kJ/mol

/

K= 30 MPa..Jm

~

1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 Crack Growth Rate (mis)

Figure 20.

Cumulative Distribution Function of Adjusted da/dt Alloy 52/152 Data Produced by ANL and PNNL ([17] and [18]); S 22% Cold Work and Constant Load/ K1 24

Dominion fn~ineerin~, Inc.

TN-5696-00-02, Rev. 0 Data for Less than 20% Cold Work from All Laboratories 1.E-09,--;===::::::;---==---=----::------=-----~---------,

1.E-10

~

.s vAMEC

~

11:1 e:::

~ 1.E-11

~ A =- :._---~ ~

==

-.-- -"no o

o ~

DO A A

(.!)

u E u

1.E-12

/

OD 0

A

-L------~~----,~----;::::============;t I

Data are adjusted for temperature (325°C).

Q = 130 kJ/mol

1. E-13........................ -'-+-......._._...................................... -'-+-........................ +-'-............ "'--+-........................ +-'-............ "'--+-........................ +-'-............ '-+-........................._.

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

Figure 21.

Plot of da/dt versus K1 for Alloy 690 Data from All Laboratories, > 10 & s 20% Cold Work, CROM and Bar Material, Constant Load or K1 Testing 1.0,-;;;;===-- ------,.----:-

__,,.__.------=:om------,

v AMEC

~

0.9

~CIEMAT 0.8 oGE-GRC 1------<>-.------1----------1

§ 0.7

• .=:

s PNNL

@ 0.6 -t---------7'<'-- ______

Vi

~ 0.5 1-----)(--,----;:::::::::::~::::::::;---1

~ 0.4 +----~

E 8 0.3 +---------

0.2 +----

0.1 ---'----*~-----.1--------1 The data points at 1 E-13 were reported as "no growth."

Data are adjusted for temperature (325°C} and stress intensity factor.

Q = 130 kJ/mol K = 30 MPavm 0.0 ~-:ic.-=............... "+-_~~.L..U-JL..f----L.--'-....L.......L...L..U-J+---'-~==a:i:====~

1.E-13 1.E-12 1.E-11 1.E-10 1.E-09 1.E-08 Crack Growth Rate (m/s)

Figure 22.

Cumulative Distribution Function of Adjusted da/dt Alloy 690 Data from All Labs, S 20% Cold Work, CROM and Bar Material, Constant Load or K1 25

Dominion En~ineerin~, Inc.

TN-5696-00-02, Rev. 0 1.E-09 rc::;:r~===----======-===::~::;:;:::;;;;;;;;;;;;;;;;-1 II DGE-GRC 1.E-10 PNNL Iii'

].

---0

~

C'CI 0::

~

.&; 1.E-11

~

~

.:it:.

D u

C'CI D

D

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

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

Figure 23.

Plot of da/dt versus K1 for Alloy 52/152 Data from All Laboratories,> 10 & S 20% Cold Work, Constant Load or Ki 1.0 oANL 0.9

~CIEMAT

/

0.8 DGE-GRC

• PNNL

§ 0.7

• g :s o.6 Cll

~ 0.5

~ 0.4 The data points at 1 E-13 I

were reported as "no growth."

E a o.3 0.2 0.1 FOi = 12 Data are adjusted for temperature (325°C) and I

stress intensity factor.

Q = 130 kJ/mol K = 30 MPa" m 0.0 a..._.__....._................. __....::::::....,..........,.__ ___

~U::::.::L+===:.:::::;:::;:::;::::;~

1.E-13 1.E-12 1.E-11 1.E-10 Crack Growth Rate (mis)

Figure 24.

Cumulative Distribution Function of Adjusted da/dt Alloy 521152 Data from All Laboratories, S 20% Cold Work, Constant Load or K1 1.E-09 26