NL-20-0115, Proposed Inservice Inspection Alternative FNP-ISI-AL T-05-06, Version 1.0

Proposed Inservice Inspection Alternative FNP-ISI-AL T-05-06, Version 1.0

ML20055F672
Person / Time
Site:	Farley
Issue date:	02/24/2020
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-20-0115
Download: ML20055F672 (14)

Text

A Southern Nuclear Ch eryl A. Gayheart Regu latory Affairs Di rector 3535 olonnadc Parkwa Birmingham. AL 35243

_05 992 5.1 16 cagayh~a@ *outhemco.co m FEB 2 4 2020 Docket Nos.: 50-364 NL-20-0115 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant, Unit 2 Proposed lnservice Inspection Alternative FNP-ISI-ALT-05-06, Version 1.0 Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(z}(1 ), Southern Nuclear Operating Company (SNC) hereby requests NRC approval of proposed inservice inspection (lSI) alternative FNP-ISI-AL T-05-06, Version 1.0. This alternative would allow ASME Code Case N-729-4, Inspection Item 84.40 for Reactor Vessel Closure Head (RVCH) nozzle and partial penetration welds of Primary Water Stress Corrosion Cracking (PWSCC)-resistant materials to be re-examined once every 20 years for Farley Nuclear Plant (FNP) Unit 2 in lieu of the current approved interval of 15 years as documented in the associated NRC safety evaluation (ADAMS accession number ML15104A192}. The proposed alternative is provided in the attached enclosure.

SNC requests approval of this alternative prior to October 1, 2020. SNC is aware of NRC draft rulemaking published for public comment (reference 5) , which generically allows a 20-year interval for RVCH penetration nozzles fabricated with resistant material like the FNP Unit 2 component items, but is submitting this proposed alternative to ensure readiness for the October 2020 FNP Unit 2 refueling outage. SNC is prepared to withdraw this proposed alternative if final publication of the rulemaking occurs prior to October 2020.

This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at 205.992 .6611.

Respectfully submitted, CAG/dsp/scm

U. S. Nuclear Regulatory Commission NL-20-0115 Page 2

Enclosure:

Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) cc : Regional Administrator, Region II NRR Project Manager- Farley Nuclear Plant Senior Resident Inspector - Farley Nuclear Plant RTYPE: CFA04.054

Joseph M. Farley Nuclear Plant, Unit 2 Proposed lnservice Inspection Alternative FNP-ISI-ALT-05-06, Version 1.0 Enclosure Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Enclosure to NL-20-0115 Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: Class 1 Description : Reactor Vessel Closure Head (RVCH) Nozzles Examination Category: Code Case N-729-4, Class 1 PWR Reactor Vessel Upper Head Item Numbers: 84.40 - Nozzles and partial-penetration welds of Primary Water Stress Corrosion Cracking (PWSCC)-resistant materials in head Component IDs: APR1-1300, Farley Unit 2 RVCH 2.0 REQUESTED APPROVAL DATE:

Approval is requested by October 1, 2020.

3.0 APPLICABLE CODE EDITION AND ADDENDA:

The Fifth lnservice Inspection (lSI) Interval Code of record for Farley Nuclear Plant (FNP) 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."

4.0 APPLICABLE CODE REQUIREMENT:

The Code of Federal Regulations 10 CFR 50.55a(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.

10 CFR 50.55a(g)(6)(ii)(D)(2) through (4) are NRC conditions to ASME Code Case N-729-4 and are not relevant to this request.

ASME Code Case N-729-4 specifies that the Reactor Pressure Vessel (RPV) upper head components shall be examined on a frequency in accordance with Table 1 of this Code case. The Table 1 inspection requirement for Item 84.40 is a volumetric or surface examination of all nozzles each inspection interval (nominally 10 calendar years) provided that flaws attributed to PWSCC have not previously been identified in the head.

FNP Unit 2 previously received authorization to defer this examination 15 years until the 2020 refueling outage (2R27) as documented in ADAMS Accession No. ML15104A192 (reference 1).

5.0 REASON FOR REQUEST:

ASME Code Case N-729-4 (reference 2) with the conditions of 10 CFR 50.55a(g)(6)(ii)(D) requires volumetric/surface examination of the RPV upper head E-1

Enclosure to NL-20-0115 Proposed Alternative FNP-181 -AL T-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) nozzles and partial-penetration welds once per Inspection Interval (nominal 10 calendar years) after the head was placed into service.

The FNP RVCH penetration nozzles and associated welds are made from Alloys 690/52/152. As discussed in Electric Power Research Institute (EPRI) report: Materials Reliability Program: Recommended Factors of Improvement fo r Evaluating Primary Water Stress Corrosion Cracking (P WSCC) Growth Rates of Thick-Wall A lloy 690 Materials and Alloy 52, 152, and Variants Welds (MRP- 386) EPRI, Palo Alto, CA: 2017 3002010756 (reference 3), that compared to Alloys 600/82/182, these materials have a much greater PWSCC resistance. Excellent operating experience with no observations of PWSCC in almost 30 years of service supports the superiority of Alloy 690 relative to Alloy 600 in PWR primary water environments, as does extensive laboratory testing.

As stated in EPRI MRP report: Materials Reliability Prog ram: Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vesse l Top Head Penetration Nozzles (MRP-375) EPRI, Palo Alto , CA: 201 4 300200244 1 (reference 4):

"This current inspection regime was established in 2004 as a conservative approach and was intended to be subject to reassessment upon the availability of additional laboratory data and plant experience on the performance of Alloy 690 and Alloy 52/152 . Since that time , plant experience and laboratory testing have continued to demonstrate the much greater resistance of these replacement alloys to PWSCC compared to that for Alloys 600/82/182 for the material conditions relevant to partial-penetration welded nozzles. Although laboratory research is ongoing to investigate and understand the times to crack initiation and the crack growth rates for these materials under various conditions, there are now sufficient data available to develop an improved technical basis for inspection of these components."

Research documented in MRP-386 further demonstrates the much greater resistance of these replacement alloys to PWSCC as compared to Alloys 600/82/182 for the conditions relevant to partial-penetration welded nozzles.

The technical bases of MRP-375 and MRP-386 together demonstrate that the reexamination interval can be extended to a 20-year interval while maintaining an acceptable level of quality and safety. The NRC staff acknowledged acceptance of this additional research through the endorsement of ASME Code Case N-729-6 in draft rulemaking published for public comment (reference 5) , which generically allows a 20-year interval for RVCH penetration nozzles fabricated with resistant material like the FNP Unit 2 component items. With final publication of rulemaking not certain to support needed outage scoping and planning, SNC is requesting approval of this alternative to allow the use of a 20-year interval for the affected FNP Unit 2 component items.

E-2

Enclosure to NL-20-0115 Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) 6.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

Proposed Alternative Pursuant to 10 CFR 50.55a(z)(1 ), SNC requests an alternative to performing the required volumetric/surface examinations for the RVCH components identified previously at the frequency prescribed in ASME Code Case N-729-4 as modified by NRC staff authorization (ML15104A192) of FNP alternative FNP-ISI-ALT-17, Version 2.0 (ML15111A387). Specifically, SNC requests to extend the frequency of the volumetric/surface examination of the RVCH penetration nozzles specified in Table 1, Item 84.40 of ASME Code Case N-729-4 for a nominal1 0 year period beyond the one inspection interval (nominally 10 calendar years) from the installation and concurrent baseline volumetric examination of the FNP Unit 2 replacement RVCH . The alternative method proposed provides an acceptable level of quality and safety for the examination frequency of the RVCH for the reasons specified herein.

Application of this alternative is requested for a nominal 20-year period from the time the baseline examination was performed for FNP Unit 2 (2005).

This request to utilize this alternative applies only to the inspection frequencies for volumetric/surface examinations of the RVCH on FNP Unit 2, not the inspection techniques.

Basis for Use Evaluations were performed to demonstrate the resistance of Alloys 690/52/152 to PWSCC under an EPRI MRP initiative provided in MRP-375. This report combines an assessment of the test data and operating experience developed since the technical basis for the 10-year interval of ASME Code Case N-729 (Revisions 1 through 4) was developed in 2004 with deterministic and probabilistic evaluations to assess the improved PWSCC resistance of Alloys 690/52/152 relative to Alloys 600/82/182.

Additional research was recently performed under an EPRI MRP initiative provided in MRP-386. This report compiled over 530 Alloy 690 Crack Growth Rate (CGR) data points and over 130 Alloy 52/152 CGR data points from seven research laboratories further supporting the improved PWSCC resistance of Alloys 690/52/152.

Evaluation of Existing Alloys 690/52/152 Data and Experience by MRP-375 Operating experience to date for replaced and repaired components using Alloys 690/52/152 have shown a proven record of resistance to PWSCC determined through numerous examinations in almost 30 years of application. This includes steam generator tubes, pressurizers, and RVCHs. In particular, the Alloys 690/52/152 operating experience includes inservice volumetric/surface examinations performed in accordance with ASME Code Case N-729 on replacement heads. Some of these examined heads had continuous full power operating temperatures that approached 613°F. However, none of these examinations revealed PWSCC and these examination results further support the low likelihood or potential for the RVCH to experience PWSCC during the extension period.

E-3

Enclosure to NL-20-0115 Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

The evaluation performed in MRP-375 considers a simple Factor of Improvement (FOI) approach applied in a conservative manner to model the increased resistance of Alloys 690/52/152 compared to Alloys 600 and 182 at equivalent temperature and stress conditions. FOis were estimated for the material improvements of Alloys 690/52/152 using an extensive database of material test data. Results for both crack initiation and crack growth conclude that there was a substantially higher resistance to PWSCC than for Alloy 600 base material and Alloy 82/182 weld materials. Figures 3-2, 3-4 , and 3-6 of MRP-375 provide crack growth rate data for Alloys 690/52/152 materials and heat affected zones with curves plotting FOis of 1, 5, 10, and 20 on a statistical basis reflecting the material variability exhibited in Materials Reliability Program (MRP ) Crack Growth Rates fo r Evaluating Primary Water Stress Corrosion Cracking (P WSCC) of Thick-Wall Alloy 600 Materials (MRP-55) Revision 1, EPRI, Palo Alto, CA: 2002. 1006695 (reference 6) for Alloy 600 material and in Materials Reliability Program, Crack Growth Rates fo r Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115), EPRI, Palo Alto, CA: 2004 (reference 7) for Alloys 82/182/132 weld material. An FOI of 20 bounds most of the data plotted and an FOI of 10 bounds essentially all of the crack growth rate data. Table 3-6 of MRP-375 provides a summary of FOis determined on the basis of crack growth rate and crack initiation data. For crack initiation, FOis reported, although significant, are conservatively small because crack initiation of Alloys 690/52/152 was not observed during testing; instead, the initiation time was assumed to be equivalent to the test duration.

Additional Evaluations Performed under MRP-375 MRP-375 applied the FOI results to perform a combination of deterministic and probabilistic evaluations to establish an appropriately conservative inspection interval for Alloy 690 RVCH penetration nozzles. 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 relationsh ip between the improved PWSCC growth resistance of Alloys 690/52/152 and the average time to certain adverse conditions.

These evaluations apply conservative crack growth rate predictions and the assumption of an existing flaw, which is replaced with a PWSCC initiation model for probabilistic evaluation. The evaluations provide a reasonable lower bound on the average time to adverse conditions, from which a conservative inspection interval may be recommended.

This evaluation draws from various EPRI MRP and industry documents that evaluate , for Alloys 600/82/182, the time from a detectable flaw being created to leakage occurring and from a leaking flaw to the time that net section collapse (nozzle ejection) would be predicted to occur. As discussed in MRP-375:

For different analyses and different crack types on an Alloy 690 reactor pressure vessel head (RPVH) , the conservative time between detectable flaw size and leakage varies between 23 and 77 EFPY [effective full power years] at 613°F (or between RIY [Re-lnspection Years]= 31 and 106). This result is supportive of an extension of the ultrasonic (UT) inspection interval to 20 calendar years.

E-4

Enclosure to NL-20-0115 Proposed Alternative FNP-181 -AL T-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

The conservative time between evident leakage and risk of net section collapse varies between 121 and 320 EFPY at 613°F (i. e., between RIY = 167 and 441) for the Alloy 690 RPVH.

These results indicate that more than 20 years is required for leakage to occur and that more than 120 years would be required to reach the critical crack size subsequent to leakage.

The probabilistic model in MRP-375 was developed to predict PWSCC degradation and its associated risks in RVCHs. The model utilized in this probabilistic evaluation is modified from the model presented in Appendix B of Materials Reliability PrograJn: Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Stuface Stress Imp rovement (MRP-335, Revision 1), EPRl, Palo Alto, CA: 2013 3002000073 (reference 8), that evaluated surface stress improvement of RVCHs with Alloy 600 nozzles. The integrated probabilistic model in MRP-375 includes sub-models for simulating component and crack stress conditions, PWSCC initiation, PWSCC growth, and flaw examination. The sub-models for crack initiation and growth prediction for Alloy 600 reactor pressure vessel head penetration nozzles in MRP-335, Revision 1, were adapted for RVCHs with Alloy 690 nozzles by applying FOis to account for the superior PWSCC resistance of Alloys 690/52/152. The average leakage frequency and average ejection frequency were determined using the Monte Carlo simulation model with conservative FOI assumptions.

The results show that, using only modest FOis for Alloys 690/52/152, the potential for developing a safety significant flaw (risk of nozzle ejection) is acceptably small for a volumetric/surface examination period up to 40 years.

The evaluations performed in 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). This number of penetrations bounds the number of penetrations found in the FNP Unit 2 replacement head.

Additional Evaluations Performed under MRP-386 MRP-386 summarizes years of laboratory testing by an international group of experts to quantify the PWSCC growth rates of Alloy 690 and its weld metals, Alloy 52/152, in simulated PWR primary water. Fracture mechanics-based tests were conducted under testing conditions designed to promote PWSCC in several product forms of wrought Alloy 690 and in several alloy variants of weld metal Alloy 52/152. For some Alloy 690 tests, laboratory-added plastic strain (i.e., cold work) of up to 30% reduction in thickness was used to accelerate PWSCC growth rates. Variables known to affect PWSCC were assessed and included in the CGR model and/or disposition equations, including: the mode I stress intensity factor, the test temperature, the yield strength of the material, the electrochemical potential in the test environment, and the orientation of the crack relative to the direction of added cold work. The data were vetted by an international expert panel and were then used to develop predictive models of the PWSCC growth rate in thick walled Alloy 690 (including the heat-affected zone) and its weld metals, Alloys 52 and 152, and variants of these alloys. The lower bound FOI for Alloy 690 compared to Alloy E-5

Enclosure to NL-20-0115 Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) 600 is 25, while the more realistic and recommended FOI is 38. For Alloy 52/152 compared to Alloy 182, the lower bound FOI is 253, while the recommended FOI is 324.

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

PWRs, including the FNP Unit 2 RVCH. The MRP-375 analyses assume a reactor vessel head operating temperature of 613°F to bound the known reactor vessel head temperatures of all U.S. PWRs currently operating. The FNP reactor pressure vessel (RPV) head temperatures are 600 oF and well within the bounds of the assumptions of MRP-375. Based on this, the FNP Unit 2 RVCH average operating temperature, which is the measure of the temperature relevant to potential PWSCC degradation, is bounded by the MRP-375 evaluation , which assumes 613°F to its main deterministic and probabilistic calculations.

As stated in MRP-375 "... to further allow consistent interpretation, all results are adjusted to an operating temperature of 613°F (323°C) using the Arrhenius relationship with an activation energy of 130 kJ/mol. This operating temperature is believed to be an upper bound for operating Alloy 690 top heads in service today." Reduced operating temperature results in a significant improvement in both crack initiation and crack propagation. As discussed in MRP-375 Case M2- Reduced Operating Temperature:

Reducing the head temperature from 613°F to 600°F (323°C to 316°C) reflects that most Alloy 690 hot heads operate below 613°F (323°C} , with a majority operating between 590°F and 600°F (31 ooc to 316°C) . The reduced temperature decreases the thermally activated PWSCC flaw initiation and growth processes (i.e., through the Arrhenius relation in the model).

Reducing the head temperature leads to a more than tenfold reduction in average ejection frequency defined in MRP-375 Appendix A.8 as the average number of predicted ejections per reactor vessel head per year.

Similarly, the average frequency of leakage, defined as the average number of predicted leaks per reactor vessel head per year, is decreased to less than half its base case value.

The FNP Unit 2 RVCH was designed and fabricated using materials and techniques to reduce susceptibility to PWSCC and with enhanced access doors for inspection of RPV head penetrations.

Th e fo II ow1nq ta bl e summanzes t he des1gn

. attn"b utes of th e FNP U n1*t 2 RVCH Feature Description Reactor Vessel (RV) Head SA-508, Grade 3, Class 1, Cladding is 309L Material Coatings The RVCH is not painted to ensure full compliance with the NRC bare metal inspection criteria for RV head inspections.

E-6

Enclosure to NL-20-0115 Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Penetration to Head Weld Alloy 52 ERNiCrFe-7 (UNS N06052) and Alloy 152 Material ERNiCrFe-7 (UNS W86152), Narrow J-Groove Full Length Control Rod 48 CRDM Penetrations Drive Mechanism SB-167 UNS N06690 (Alloy 690)

Penetrations Instrument Port Penetrations 4 core exit thermocouple nozzle assemblies SB-167 UNS N06690 (Alloy 690)

Reactor Vessel Level 2 RVLIS Indication System (RVLIS) SB-167 UNS N06690 (Alloy 690)

Penetrations Reactor Coolant Gas Vent 1 inch Schedule 160 SB-167 UNS N06690 (Alloy 690)

System Penetration Pipe with Alloy 52/152 ERNiCrFe-7 Narrow J-Groove Weld Nozzle Note that the probabilistic analysis in MRP-375 was pertormed assuming a head with 89 partial-penetration welded nozzles. This number bounds the number present in the FNP Unit 2 replacement head (55 nozzles) . The number of penetrations included in the probabilistic model is not a key variable, and the assumed number of penetrations results in a small change in results relative to other sensitivity cases. Thus, the probabilistic calculations of MRP-375 cover all U.S. replacement RVCHs regardless of the precise number of penetrations.

The RVCH was buttered with Alloy 152 in the area of the penetration. A partial penetration J-groove weld using Alloy 52/52M (ERNiCrFe-7 UNS N06052) filler metal was used between the Alloy 690 penetration and the head on the inside of the RVCH. Two modifications were introduced in the weld to reduce residual stress: (1) A narrow gap J-groove weld edge preparation was used to reduce the residual stress by reducing the volume of weld metal deposited, and (2) J-welding with spray cooling inside the CRDM nozzle was utilized to reduce the residual stress by changing the stress distribution through the thickness of the CRDM nozzle. Mock-up testing pertormed to confirm the residual stress reduction showed a maximum of approximately 40 ksi, a 30% reduction over the conventional grove/shielded metal arc welding process. These methods substantially reduce susceptibility beyond that assumed in the generic MRP-375 study.

A bare metal visual (BMV) examination was pertormed of the FNP Unit 2 replacement RVCH in the spring of 2019, in accordance with ASME Code Case N-729-4. The bare metal visual examination was pertormed by VT-2 qualified examiners on the outer surtace of the RVCH including the annulus area of the penetration nozzles. This examination did not reveal any surtace or nozzle penetration boric acid that would be indicative of nozzle leakage. Continued visual inspection in accordance with ASME Code Case N-729-4 will provide prompt detection of any potential leakage. Additionally, the FNP Unit 1 head has been examined as required since the installation in 2004. A volumetric examination was pertormed on the FNP Unit 1 head in the spring of 2015 (1 R26) and the last BMV examination was pertormed in the spring of 2018 (1 R28). None of the required examinations revealed any evidence of defects or leakage from the RVCH.

E-7

Enclosure to NL-20-0115 Proposed Alternative FNP-181-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

As an alternative to the requi rements of A8ME Code Case N-729-4 with conditions of 10 CFR 50 .55a for volumetric/surface examination of the reactor vessel closure heads on a nominal 10-year interval and extended to 15 years by NRC authorization of FNP-181-ALT-17 (reference 1), 8NC requests approval to perform volumetric examination of the RVCH to the requirements of A8ME Code Case N-729-4 with conditions of 10 CFR 50.55a on a nominal 20-year interval.

Minimum FOIImplied by the Requested Inspection Period A8ME Code Case N-729 is based upon conclusions reached in Materials Reliability Program Inspection Plan fo r Reactor Vesse l Closure Head Penetrations in U.S. PWR Plants (MRP-1 17), EPRI, Pal o Alto, CA: 2004. 1007830 (reference 9) that a reexamination interval between volumetric/surface examinations of one 24-month operating cycle is acceptable for a head with Alloy 600 nozzles and operating at a temperature of 605°F. The inspection period for heads with Alloy 690 nozzles in A8ME Code Case N-729-4 is a nominal10 years, which represents a minimum implied FOI of five over Alloy 600.

Per the technical basis documents for A8ME Code Case N-729 for heads with Alloy 600 nozzles (references 9, 10, and 11 ), the effect of differences in operating temperature on the required volumetric/surface reexamination interval for heads with Alloy 600 nozzles can be easily addressed on the basis of the RIY parameter. The RIY parameter adjusts the EFPY of operation between inspections for the effect of head operating temperature using the thermal activation energy appropriate to PW8CC crack growth. For heads with Alloy 600 nozzles, A8ME Code Case N-729-4 as conditioned by 10 CFR 50.55a(g)(6)(ii)(D) limits the interval between subsequent volumetric/surface inspections to RIY = 2.25. The RIY parameter, which is referenced to a head temperature of 600°F, limits the time available for potential crack growth between inspections.

The RIY parameter for heads with Alloy 600 nozzles is adjusted to the reference head temperature using activation energy of 130 kJ/mol (31 kcal/mol) (reference 2). Based on the available laboratory data, the same activation energy is applicable to model the temperature sensitivity of growth of a hypothetical PW8CC flaw in the Alloys 690/52/152 material of the replacement RVCHs. Key laboratory crack growth rate testing data for Alloy 690 wrought material investigating the effect of temperature are as follows:

1 . Results from Argonne National Laboratory indicate that Alloy 690 with 0-26% cold work has an activation energy between 100 and 165 kJ/mol (24-39 kcal/mol).

NUREG/CR-7137 (reference 12) concludes that the activation energy for Alloy 690 at deformation levels up to 20% is comparable to the standard value for Alloy 600 (130 kJ/mol) .

2. Testing at Pacific Northwest National Laboratory, which found activation energy of approximately 120 kJ/mol (28.7 kcal/mol) for Alloy 690 materials with 17-31% cold work, is summarized in Appendix M of Mate rials Reliability Program: Resistance of Alloys 690, 152, and 52 to Primary Water Stress Corrosion Cracking (MRP-237, Rev.2): Summary of Findings Between 2008 and 2012from Completed and Ongoing Test Programs. EPRI, Palo A lto, CA: 2013 30020001 90 (reference 13).

These data show that it is reasonable to assume the same crack growth thermal activation energy as was determined for Alloys 600/82/182 (namely 130 kJ/mol (31 kcal/mol)) for E-8

Enclosure to NL-20-0115 Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) modeling growth of hypothetical PWSCC flaws in Alloys 690/52/152 PWR plant components.

As discussed in Materials Reliability Program Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PWR Plants (MRP -11 7), EPRI, Palo Alto, CA: 2004 1007830 (reference 9) the technical basis document for heads with Alloy 600 nozzles, effective time for crack growth is the principal basis for setting the appropriate reexamination interval to detect any PWSCC in a timely fashion. U.S. PWR inspection experience for heads with Alloy 600 nozzles has confirmed that the RIY = 2.25 interval results in a suitably conservative inspection program .

The FNP Unit 2 RVCH operating temperature is 600°F, which is the reference temperature in Code Case N-729-4. An activation energy of 31 kcal/mol (130 kJ/mol) for crack growth of ASME Code Case N-729-4 is appropriate to apply for modeling crack growth of Alloys 690/52/152 plant components. Conservatively assuming that the EFPYs of operation since the FNP Unit 2 RVCH was replaced is equal to the calendar years, the FOI implied by this RIY value for FNP Unit 2 is:

FOI = (20 RIY69o) I (2.25 RIY6oo ) = 8.9.

Considering the statistical compilation of data provided in Figures 3-2 , 3-4, and 3-6 of MRP-375 , this factor of improvement is less than the FOI of 10 that bounds essentially all of the crack growth rate data presented in MRP-375 and less than half the minimum FOI of 25 presented in Table 5-1 of MRP-386. Furthermore, as discussed in Sections 2 and 3 of MRP-375 , PWR plant experience and laboratory testing have demonstrated a large improvement in resistance to PWSCC initiation of Alloys 690/52/152 in comparison to that for Alloys 600/82/182. Hence, the demonstrated improvements in PWSCC initiation and growth confirm on a conservative basis the acceptability of the requested period of extension.

Conclusions The proposed alternative provides an acceptable level of quality and safety for structural integrity as the Alloy 690 nozzle base and Alloy 52/152 weld materials used in the FNP Unit 2 replacement RVCH provide a 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 minute. The minimum FOI implied by the requested extension period represents a level of reduction in PWSCC crack growth rate versus that for Alloys 600/82/182 that is completely bounded on a statistical basis by the laboratory data compiled in MRP-375. Given the lack of PWSCC detected to date in any PWR plant applications of Alloys 690/52/152, the simple FOI assessment clearly supports the requested period of extension. In addition , as demonstrated in proposed rules the NRC staff has endorsed Code N-729-6 which specifies a 20-year interval for plants with RVCH penetrations fabricated from resistant materials such as FNP Unit 2 (reference 5).

SNC proposes to perform volumetric/surface examinations of the FNP Unit 2 reactor pressure vessel head in accordance with ASME Code Case N-729-4 with conditions of 10 CFR 50.55a on a nominal 20-year interval as an alternative to the nominal 10-year interval required by the ASME Code Case N-729-4.

E-9

Enclosure to NL-20-0115 Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

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

7.0 DURATION OF PROPOSED ALTERNATIVE:

The proposed alternative is requested for the remainder of the 5th lnservice Inspection (lSI) Interval for Farley Unit 2, currently scheduled to end on 11/30/27.

8.0 PRECEDENT

There have been many submittals from multiple plants requesting an alternative from the nominal10-year interval of ASME Code Case N-729-1 (and now ASME Code Case N-729-4) for volumetric/surface examinations of RVCHs with Alloy 690 nozzles. A selection of some of the plants is shown below. Alternative intervals greater than 15 years have previously been granted in order to align with scheduled refueling outages. The alternatives approved for two sites (Items 3 and 4) extended the inspection interval from an initial approved alternative to a total interval of up to 15.5 years. The approved alternative for Calvert Cliff Units 1 and 2 (Item 2) permitted an inspection interval not to exceed 16 years in order to align with scheduled refueling outages. The approved alternative for Prairie Island Generating Stations Units 1 and 2 (Item 1) permitted an inspection interval not to exceed 20 years.

1. Letter from NRC to Northern States Power, "Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Relief Request from the Requirements of the ASME Code (EPID: L-2018-LLR-0023)," dated February 25, 2019 (ADAMS Accession No. ML19046A166)

2. Letter from NRC to Exelon Nuclear, "Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Relief Request for Extension of Volumetric Examination Interval for Reactor Vessel Heads with Alloy 690 Nozzles (CAC Nos. MF5829 and MF5830)," dated December 7, 2015 (ADAMS Accession No. ML15327A367)

3. Letter from NRC to Entergy Operations, Inc., "Arkansas Nuclear One, Unit 1, Request for Alternative AN01-ISI-026 from Volumetric/Surface Examination Frequency Requirements of American Society of Mechanical Engineers Code Case N-729-1 (CAC No. MF8007)," dated February 13, 2017 (ADAMS Accession No. ML17018A283)

4. Letter from NRC to FirstEnergy Nuclear Operating Company, "Beaver Valley Power Station, Unit No. 1 -Issuance of Relief Request 1-TYP-4-RV-05, Revision 0, From Certain ASME Code Reactor Pressure Vessel Penetration Examination Frequency Requirements (CAC No. MF9283)," dated September 8, 2017 (ADAMS Accession No. ML17222A162)

9.

REFERENCES:

1. Joseph M. Farley, Unit 2, Alternative to lnservice Inspection Regarding Reactor Closure Head Nozzle and Partial Penetration Welds, Final Safety Evaluation Report dated May 5, 2015 (ADAMS Accession No. ML15104A192)

E-10

Enclosure to NL-20-0115 Proposed Alternative FNP-ISI-ALT-05-06, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

2. 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

3. Electric Power Research Institute (EPRI) Materials Reliability Program:

"Recommended Factors of Improvement for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) Growth Rates of Thick-Wall Alloy 690 Materials and Alloy 52, 152, and Variants Welds (MRP-386)," EPRI, Palo Alto, CA, Final Report-December 2017 (Report No. 300201 0756)

4. EPRI Materials Reliability Program: "Technical Basis for Reexamination Interval Extension for Alloy 690 PWR Reactor Vessel Top Head Penetration Nozzles (MRP-375) ," EPRI, Palo Alto , CA, Final Report- February 2014 (Report No.

3002002441)

5. Federal Register I Vol. 83, No. 218 I Friday, November 9, 2018 I Proposed Rules

6. EPRI Materials Reliability Program: "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, Final Report- November 2002 (Report No. 1006695)

7. EPRI Materials Reliability Program: "Crack Growth Rates for Evaluating Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (M RP-115)," EPRI , Palo Alto, CA, Final Report- November 2004 (Report No. 1006696)

8. EPRI Materials Reliability Program : "Topical Report for Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement (MRP-335, Revision 1)," EPRI , Palo Alto, CA, Final Report- January 2013 (Report No. 3002000073)

9. EPRI Materials Reliability Program : "Inspection Plan for Reactor Vessel Closure Head Penetrations in U.S. PWR Plants (MRP-117) ," EPRI , Palo Alto , CA, Final Report - December 2004 (Report No. 1007830) (ADAMS Accession No. ML043570129)

10. EPRI Materials Reliability Program: "Reactor Vessel Closure Head Penetration Safety Assessment for U.S. PWR Plants (MRP-11 ONP) ," EPRI, Palo Alto , CA, Final Report- April 2004 (Report No. 1009807-NP) (ADAMS Accession No. ML041680506)

11. EPRI Materials Reliability Program: "Probabilistic Fracture Mechanics Analysis of PWR Reactor Pressure Vessel Top Head Nozzle Cracking (MRP-1 05 NP)," EPRI ,

Palo Alto , CA, Final Report- April2004 (Report No. 1007834) (ADAMS Accession No. ML041680489)

12. U.S. NRC, "Stress Corrosion Cracking in Nickel-Base Alloys 690 and 152 Weld in Simulated PWR Environment-2009, NUREGICR-7137, ANL-1 0136," published June 2012 (ADAMS Accession No. ML12199A415)

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

EPRI , Palo Alto, CA, Technical Update- April 2013 (Report No. 3002000190)

E-11