View source for NL-16-1329, Proposed Inservice Inspection Alternative VEGP-ISI-ALT-11, Version 2.0 and Proposed Inservice Inspection Alternative FNP-ISI-ALT-19, Version 2.0

{{Adams
| number = ML16221A066
| issue date = 08/04/2016
| title = Proposed Inservice Inspection Alternative VEGP-ISI-ALT-11, Version 2.0 and Proposed Inservice Inspection Alternative FNP-ISI-ALT-19, Version 2.0
| author name = Wheat J
| author affiliation = Southern Co, Southern Nuclear Operating Co, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000348, 05000424, 05000425
| license number = 
| contact person = 
| case reference number = NL-16-1329
| document type = Inservice/Preservice Inspection and Test Report, Letter type:NL
| page count = 25
}}

=Text=
{{#Wiki_filter:Justin T. Wheat               Southern Nuclear Nuclear Licensing Manager     Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 Tel 205.992.5998               SOUTHERN Fax 205.992.7601 NUCLEAR A SOUTHERN COMPANY AUG O 4 2016 Docket Nos.: 50-424         50-348                                     NL-16-1329 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Vogtle Electric Generating Plant, Units 1 and 2 Joseph M. Farley Nuclear Plant, Unit 1 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11, Version 2.0 and Proposed lnservice Inspection Alternative FNP-ISl-ALT-19, Version 2.0 Ladies and Gentlemen:
By letters dated June 28, 2016, and June 30, 2016 (Agencywide Documents Access and Management System Accession Nos. ML16180A046 and ML16182A475, respectively), Southern Nuclear Operating Company (SNC) submitted a request for relief for the Vogtle Electric Generating Plant (VEGP),
Units 1 and 2 and Joseph M. Farley Nuclear Plant (FNP), Unit 1. SNC requests relief to eliminate the reactor pressure vessel threads in flange examination requirement for the remainder of the inservice inspection intervals. Please note that FNP Unit 2 already performed this examination for their current inservice inspection interval in 2010.
Based on the Nuclear Regulatory Commission (NRC) staff's acceptance review, SNC is resubmitting these two Alternatives. Enclosure 1 resubmits the VEGP
* Alternative, and Enclosure 2 resubmits the FNP Alternative. These Enclosures supersede Version 1.0 of the respective Alternative in their entirety. Enelosure 3 contains the Electric Power Research Institute (EPRI) report that forms the basis of the two Alternatives. Please note that Enclosure 3 is being submitted to the NRC staff for information only and is publicly available for download on EPRl's website.
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.
Respectfully submitted,
 ~~!(~
Nuclear Licensing Manager JTW/RMJ

U.S. Nuclear Regulatory Commission NL-1S-1329 Page 2

==Enclosures:==
: 1. Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0
: 2. Proposed lnservice Inspection Alternative FNP-ISl-ALT-19, Version 2.0
: 3. EPRI Nondestructive Evaluation Report- Reactor Pressure Vessel Threads in Flange Examination Requirements.
3002007626; Dated: March 2016 cc:  Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. 0. G. Bost, Executive Vice President & Chief Nuclear Officer Ms. C. A. Gayheart, Vice President - Farley Mr. 0. R. Madison, Vice President - Fleet Operations Mr. B. K. Taber, Vice President - Vogtle 1 & 2 Mr. M. 0. Meier, Vice President - Regulatory Affairs Mr. B. J. Adams, Vice President- Engineering Mr. C. R. Pierce, Regulatory Affairs Director Ms. B. L. Taylor, Regulatory Affairs Manager - Farley Mr. G. W. Gunn, Regulatory Affairs Manager- Vogtle 1 & 2 RType: Farley=CFA04.054; Hatch=CHA02.004; Vogtle=CVC7000 U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. S. A. Williams, NRA Project Manager - Farley Ms. N. R. Childs, Senior Resident Inspector- Vogtle 1 & 2 Mr. P. K. Niebaum, Senior Resident Inspector - Farley Mr. R. E. Martin, NRA Project Manager - Vogtle 1 & 2

Vogtle Electric Generating Plant, Units 1 and 2 Joseph M. Farley Nuclear Plant, Unit 1 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 and Proposed lnservice Inspection Alternative FNP-ISl-ALT-19, Version 2.0 Enclosure 1 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 to NL-16-1329 Proposed lnservice Inspection Alternative VEGP-ISl:ALT-11 Version 2.0 Plant Site-Unit:
Vogtle Electric Generating Plant (VEGP) - Units 1 and 2 Interval-Interval Dates:
3rd lnservice Inspection (ISi) Interval, May 31, 2007 through May 30, 2017 Requested Date for Approval and Basis:
Approval is requested by February 28, 2017 to allow implementation of the proposed alternative during the 20th Refueling Outage of VEGP Unit 1. This proposed alternative would also apply to all the remaining refueling outages for both units during the third ISi Interval.
ASME Code Components Affected:
ASME Category B-G-1, Item 86.40, Pressure retaining bolting greater than 2-inches, Reactor Vessel - threads in flange require a volumetric examination.

===Applicable Code Edition and Addenda===
The applicable Code edition and addenda is ASME Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2001 Edition with Addenda through 2003.
Applicable Code Requirements:
The Reactor Pressure Vessel (RPV) threads in flange are examined using a volumetric examination technique with 100% of the flange ligament areas examined every ISi interval_. The examination area is the one-inch area around each RPV stud hole, as shown on Figure IWB-2500-12.
Background and Reason for Request:
Southern Nuclear Operating Company (SNC) has worked with the industry to evaluate eliminating the RPV threads in flange examination requirement. Licensees in th.e US and internationally have worked with the Electric Power Research Institute (EPRI) to produce a technical report (Reference 1) which provides the basis for elimination of the requirement. Reference 1 evaluates potential degradation mechanisms and includes a stress analysis I flaw tolerance evaluation. Also included in Reference 1 is a review of operating experience based on a survey of inspection results from over 168 units related to RPV flange/bolting and related RPV assessments. The conclusion from Reference 1 is that the current requirements. are not commensurate with the associated burden (worker exposure, personnel safety, radwaste, increased time at reduced RCS inventory) of the examination. The technical basis for this alternative is discussed in more detail below.
E1-1 to NL-16-1329 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 Potential Degradation Mechanisms An evaluation of potential degradation mechanisms that could impact flange/threads reliability was performed as part of Reference 1. Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue, stress corrosion cracking, crevice corrosion, velocity phenome, dealloying corrosion and general corrosion, stress relaxation, creep, mechanical wear and mechanical/thermal fatigue. Other than the potential for mechanical/thermal fatigue, there are no active degradation mechanisms identified for the threads in flange component.
The EPRI report notes a general conclusion from Reference 2, (which includes work supported by the United States Nuclear Regulatory Commission (USN RC))
that when a component item has no active degradation mechanism present, and a preservice inspection has confirmed that the inspection volume is in good condition (i.e. no flaws I indications), then subsequent inservice inspections do not provide additional value going forward. As discussed in the OE review summary below, the RPV flange ligaments have received the required pre-service examinations, and over 10,000 inservice inspections, with no relevant findings.
To address the potential for mechanical/thermal fatigue, Reference 1 documents a stress analysis and flaw tolerance evaluation of the flange thread area to assess mechanical/thermal fatigue potential. The evaluation consists of two parts. In the first part, stress analysis is performed considering all applicable loads on the threads in flange component. In the second part, the stresses at the critical locations of the component are used in a fracture mechanics evaluation to determine the allowable flaw size for the component as well as how much time it will take for a postulated initial flaw to grow to the allowable flaw size using guidelines in the ASME Code, Section XI IWB-3500.
Stress Analysis A stress analysis was performed in Reference 1 to determine the stresses at critical regions of the threads in flange component as input to a flaw tolerance evaluation.
Sixteen nuclear plant units (ten PWRs and six Boiling Water Reactors (BWRs))
were considered in the analysis. The evaluation was performed using a geometrical configuration that bounds the sixteen units considered in this effort. The details of the RPV parameters for VEGP Units 1 and 2 as compared to the bounding values used in the evaluation are shown in Table 1. As can be seen from this table, the diameter of the stud used in the analysis is smaller than that at VEGP Units 1 and 2; all other parameters are the same. The smaller stud diameter results in higher preload per bolt. Hence, the stresses from the analyzed configuration would be conservative in application to VEGP Units 1 and 2. Dimensions of the analyzed geometry are shown in Figure 1.
E1-2 to NL-16-1329 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 Table 1: Comparison of VEGP Units 1 and 2 Parameters to Bounding Values Used in Analysis Flange Stud         RPV Inside Thickness      Design No. of        Nominal       Diameter at Plant                                                           at Stud     Pressure Studs        Diameter         Stud Hole Hole        (psi a)
(inches)         (inches)
(inches)
VEGP-1, 2 54             7.0              173            16          2500 Range for 16 54- 60        6.5 - 7.0       157 - 173        15 -16        2500 Units Considered Bounding Values Used in           54             6.0              173            16          2500 Analysis The analytical model is shown in Figures 2 and 3. The loads considered in the analysis consisted of:
* A design pressure of 2500 psia at an operating temperature of 600&deg;F was applied to all internal surfaces exposed to internal pressure.
* Bolt/stud preload -The preload was calculated as detailed in VEGP Units 1 and 2 RPV manual. The preload on the bounding geometry is calculated as:
C*P*ID 2      1.1*2500*1732 Ppreload =             =                  =42,338 psi S *D 2           54*6 2 where:
Ppreload     =       Preload pressure to be applied on modeled bolt (psi) p            =       Internal pressure (psi)
ID           =       Largest inside diameter of RPV (in.)
c            =       Bolt-up contingencies (+ 10%)
s            =       Least number of studs D            =       Smallest stud diameter (in.)
* Thermal stresses - The only significant transient affecting the bolting flange is heat-up/cooldown. This transient typically consists of a steady 100&deg;F/hour ramp up to the operating temperature, with a corresponding pressure ramp up to the operating pressure.
The ANSYS finite element analysis program was used to determine the stresses in the threads in flange component for the three loads described above.
E1-3 to NL-16-1329 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 Flaw Tolerance Evaluation A flaw tolerance evaluation was performed using the results of the stress analysis to determine how long it would take an initial postulated flaw to reach the ASME Code, Section XI allowable flaw size. A linear elastic fracture mechanics evaluation consistent with ASME Code, Section XI, IWB 3600 was performed.
Stress intensity factors (Ks) at four flaw depths of a 360&deg; inside-surface-connected, partial-through-wall circumferential flaw are calculated using finite element analysis techniques with the model described above. The maximum stress intensity factor (K) values around the bolt hole circumference for each flaw depth (a) are extracted and used as to perform the crack growth calculations. The circumferential flaw is modeled to start between the 10th and 11th flange threads from the top end of the flange because that is where the largest tensile axial stress occurs. The modeled flaw depth-to-wall thickness ratios (alt) are 0.02, 0.29, 0.55, and 0.77, as measured in any direction from the stud hole. This creates an ellipsoidal flaw shape around the circumference of the flange, as shown in Figure 4 for the flaw model with alt= 0.77 crack model. The crack tip mesh for the other flaw depths follows the same pattern.
When preload is not being applied, the stud, stud threads, and flange threads are not modeled. The model is otherwise unchanged between load cases.
The maximum K results are summarized in Table 2 for the four crack depths.
Because the crack tip varies in depth around the circumference, the maximum K from all locations at each crack size is conservatively used for the K vs. a profile.
Table 2: Maximum K vs. alt K at Crack Depth (ksivin)
Load 0.02 alt 0.29 alt 0.55 alt 0.77 alt Pre load                   11.2       17.4     15.5       13.9 Preload + Heatup + Pressure             13.0       19.8     16.1       16.3 The allowable stress intensity factor was determined based on the acceptance criteria in ASME Section XI, IWB-3610/Appendix A which states that:
K, < K1J-V1 o = 69.6 ksivin Where, K1 =Allowable stress intensity factor (ksivin)
Kie = Lower bound fracture toughness at operating temperature (220 ksivin)
As can be seen from Table 2, the allowable K is not exceeded for all crack depths up to the deepest analyzed flaw of alt =0.77. Hence the allowable flaw depth of the 360&deg; circumferential flaw is at least 77% of the thickness of the flange. The allowable flaw depth is assumed to be equal to the deepest modeled crack for the purposes of this analysis.
For the crack growth evaluation, an initial postulated flaw size of 0.2 in. (5.08 mm) is chosen consistent with the ASME Code, Section XI IWB-3500 flaw acceptance E1-4 to NL-16-1329 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 standards. The deepest flaw analyzed is alt= 0.77 because of the inherent limits of the model. Two load cases are considered for fatigue crack growth: heat-up/cooldown and bolt preload. The heat-up/cooldown load case includes the stresses due to thermal and internal pressure loads and is conservatively assumed to occur 50 times per year. The bolt preload is assumed to be present and constant during the load cycling of the heat-up/cooldown load case. The bolt preload load case is conservatively assumed to occur five times per year, and these cycles do not include thermal or internal pressure. The resulting crack growth was determined to be negligible due to the small delta K and the relatively low number of cycles associated with the transients evaluated. Because the crack growth is insignificant, the allowable flaw size will not be reached and the integrity of the component is not challenged for at least 80 years (original 40-year design life plus additional 40 years of plant life extension).
The stress analysis I flaw tolerance evaluation presented above show that the threads in flange component at VEGP Units 1 and 2 is very flaw tolerant and can operate for 80 years without violating ASME Code, Section XI safety margins. This clearly demonstrates that the threads in flange examinations can be eliminated without affecting the safety of the RPV.
Operating Experience Review Summary As discussed above, the results of the survey, which includes results from VEGP, confirmed that the RPV threads in flange examination are adversely impacting outage activities (dose, safety, increased time at reduced RCS inventory) while not identifying any service induced degradations. Specifically, for the US fleet, a total of 94 units have responded to date and none of these units have identified any type of degradation. As can be seen in Table 3 below, the data is encompassing. The 94 units represent data from 33 BWRs and 61 PWRs. For the BWR units, a total 3,793 examinations were conducted and for the PWR units a total of 6,869 examinations were conducted, with no service induced degradation identified. The response data includes information from all of the plant designs in operation in the US and includes BWR-2, -3, -4, -5 and -6 designs. The PWR plants include the 2- loop, 3-loop and 4-loop designs and each of the PWR NSSS designs (i.e., Babcock & Wilcox, Combustion Engineering and Westinghouse).
Table 3: Summary of Survey Results - US Fleet Plant Type         Number of        Number of          Number of Units        Examinations         Reportable Indications BWR               33               3,793                0 PWR               61               6,869                0 Total             94              10,662                0 Related RPV Assessments In addition to the examination history and flaw tolerance discussed above, Reference 1 discusses studies conducted in response to the issuance of the E1-5 to NL-16-1329 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 Anticipated Transient Without Scram (ATWS) Rule by the USN RC. This rule was issued to require design changes to reduce expected ATWS frequency and consequences. Many studies have been conducted to understand the A TWS phenomena and key contributors to successful response to ATWS event. In particular, the reactor coolant system and its individual components were reviewed to determine weak links. As an example, even though significant structural margin was identified in USNRC SECY-83-293 for PWRs, the ASME Service Level C pressure of 3200 psig was assumed be to an unacceptable plant condition. While a higher ASME service level might be defensible for major Reactor Coolant System (RCS) components, other portions of the RCS could deform to the point of inoperability. Additionally, there was the concern that steam generator tubes might fail before other RCS components, with a resultant bypass of containment. The key take away for these studies is that the RPV flange ligament was not identified as a _
weak link and other RCS components were significantly more limiting. Thus, there is substantial structural margin associated with the RPV flange.
In summary, Reference 1 identifies that the RPV threads in flange are performing with very high reliability based on operating and examination experience. This is due to the robust design and a relatively benign operating environment (e.g. the number and magnitude of transients is small, generally not in contact with primary water at plant operating temperatures/pressures, etc.). The robust design is manifested in that plant operation has been allowed at several plants even with a bolt/stud assumed to be out of service. As such, significant degradation of multiple bolts/threads would be needed prior to any RCS leakage.
Proposed Alternative and Basis for Use:
In lieu of the requirements for a volumetric ultrasonic examination, SNC proposes eliminating the requirement for the RPV threads in flange examination for VEGP.
SNC has confirmed that VEGP plant specific parameters (e.g. vessel diameter, number of studs, inservice inspection findings) are consistent with or bounded by Reference 1.
Since there is reasonable assurance that the proposed alternative is an acceptable alternate approach to the performance of the ultrasonic examinations, SNC requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(z)(1) on the basis that the use of the alternative provides an acceptable level of quality and safety.
To protect against non-service related degradation, each outage SNC uses a detailed procedure for the removal, care and visual inspection of the RPV studs and the threads in flange. Care is taken to not only remove the studs, but once the studs are removed to inspect the RPV threads for damage and to install RPV stud plugs to protect threads from damage. Prior to reinstallation, the studs and stud holes are cleaned and lubricated. The studs are then replaced and tensioned into the Reactor Vessel. This activity is performed each refueling outage and the procedure documents each step.
These controlled maintenance activities provide further assurance that degradation is detected and mitigated prior to returning the reactor to service.
In addition, it is important to note all other inspection activities, including the system leakage test (examination category 8-P), which is conducted each refueling outage, will continue to be performed.
E1-6 to NL-16-1329 Proposed lnservice Inspection Alternative VEGP-ISI-ALT-11 Version 2.0 Duration of Proposed Alternative:
The alternative is requested for the current Third lnservice Inspection Interval, which began May 31, 2007 and is scheduled to end on May 30, 2017 for both Unit 1 and 2.

==References:==
: 1. EPRI Nondestructive Evaluation Report - Reactor Pressure Vessel Threads in Flange Examination Requirements. 3002007626; Dated: March 2016 Electric Power Research Institute (EPRI).
: 2. American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR)
Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.
Status:
Pending NRG approval.
E1-7 to NL-16-1329 Proposed lnservice inspection Alternative VEGP-ISl-ALT-11 Version 2.0 J
R86.5" I
J 8.5" 17.0"                            7.0" 16.0" R83.75" R4.5" 10.75" R85.69" Figure 1 Modeled Dimensions E1-8 to NL-16-1329 Proposed lriservice Inspection Alternative VEGP-ISl-ALT-11 Versio.n 2.0 l ELEMNTS REAL NU1 ANO_Vessel_Flange Figure 2 Finite Element Model Showing Bolt and Flange Connection E1-9

Enclosure
   .       1 to NL-16-1329                               .
Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 Figure 3 Finite Element Model Mesh with Detail at Thread Location E1-10

Vogtle Electric Generating Plant, Units 1 and 2 Joseph M. Farley Nuclear Plant, Unit 1 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 and Proposed lnservice Inspection Alternative FNP-181-ALT-19, Version 2.0 Enclosure 2 Proposed lnservice Inspection Alternative FNP-ISl-ALT-19 Version 2.0 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-ISl-ALT-19 Version 2.0 Plant Site-Unit:
Joseph M. Farley Nuclear Plant (FNP) - Unit 1 Interval-Interval Dates:
4th lnservice Inspection {ISi) Interval, December 1, 2007 through November 30, 2017 Requested Date for Approval and Basis:
Approval is requested by July 1, 2017 to allow the proposed alternative during the 28th Refueling Outage of FNP Unit 1. This proposed alternative would apply to Unit 1 for the fourth ISi Interval.
ASME Code Components Affected:
ASME Category B-G-1, Item 86.40, Pressure retaining bolting greater than 2-inches, Reactor Vessel - threads in flange require a volumetric examination.

===Applicable Code Edition and Addenda===
The applicable Code edition and addenda is ASME Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2001 Edition with Addenda through 2003.
Applicable Code Requirements:
The Reactor Pressure Vessel (RPV) threads in flange are examined using a volumetric examination technique with 100% of the flange ligament areas examined every ISi interval. The examination area is the one-inch area around each RPV stud hole, as shown on Figure IWB-2500-12.
Background and Reason for Request:
Southern Nuclear Operating Company (SNC) has worked with the industry to evaluate eliminating the RPV threads in flange examination requirement. Licensees in the US and internationally have worked with the Electric Power Research Institute (EPRI) to produce a technical report (Reference 1) which provides the basis for elimination of the requirement. Reference 1 evaluates potential degradation mechanisms and includes a stress analysis I flaw tolerance evaluation. Also included in Reference 1 is a review of operating experience (OE) based on a survey of.inspection results from over 168 units related to RPV flange/bolting and related RPV assessments.
The conclusion from Reference 1 is that the current requirements are not commensurate with the associated burden (worker exposure, personnel safety, radwaste, increased time at reduced RCS inventory) of the examination. The technical basis for this alternative is discussed in more detail below.
E2-1 to NL-16-1329 Proposed lnservice Inspection Ahernative FNP-ISl-ALT-19 Version 2.0 Potential Degradation Mechanisms An evaluation of potential degradation mechanisms that could impact flange/threads reliability was performed as part of Reference 1. Potential types of degradation evaluated included pitting, intergranular attack, corrosion fatigue, stress corrosion cracking, crevice corrosion, velocity phenome, dealloying corrosion and general corrosion, stress relaxation, creep, mechanical wear, and mechanical/thermal fatigue. Other than the potential for mechanical/thermal fatigue, there are no active degradation mechanisms identified for the threads in flange component.
The EPRI report notes a general conclusion from Reference 2, (which includes work supported by the United States Nuclear Regulatory Commission (USNRC))
that when a component item has no active degradation mechanism present, and a preservice inspection has confirmed that the inspection volume is in good condition (i.e. no flaws I indications), then subsequent_inservice inspections do not provide additional value going forward. As discussed in the OE review summary below, the RPV flange ligaments have received the required pre-service examinations, and over 10,000 inservice inspections, with no relevant findings.
To address the potential for mechanical/thermal fatigue, Reference 1 documents a stress analysis and flaw tolerance evaluation of the flange thread area to assess mechanical/thermal fatigue potential. The evaluation consists of two parts. In the first part, stress analysis is performed considering all applicable loads on the threads in flange component. In the second part, the stresses at the critical locations of the component are used in a fracture mechanics evaluation to determine the allowable flaw size for the component as well as how much time it will take for a postulated initial flaw to grow to the allowable flaw size using guidelines in the ASME Code, Section XI IWB-3500.
Stress Analysis A stress analysis was performed in Reference 1 to determine the stresses at critical regions of the threads in flange component as input to a flaw tolerance evaluation.
Sixteen nuclear plant units (ten PWRs and six Boiling Water Reactors (BWRs))
were considered in the analysis. The evaluation was performed using a geometrical configuration that bounds the sixteen units considered in this effort. The details-of the RPV parameters for FNP as compared to the bounding values used in the evaluation are shown in Table 1. (Please note that Unit 1 and Unit 2.values are identical for these Table 1 parameters.) As can be seen from this table, the configuration used in the analysis has a larger RPV diameter, the same bolt diameter, and fewer studs compared to those at FNP. The larger RPV diameter results in higher pressure and thermal stresses and the fewer number of studs results in higher preload per stud. Hence, the stresses from the analyzed configuration would be conservative in application to FNP. Dimensions of the analyzed geometry are shown in Figure 1.
E2-2 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-181-ALT-19 Version 2.0 Table 1: Comparison of FNP Parameters to Bounding Values Used in Analysis Stud        RPV Inside        Flange Design No. of         Nominal        Diameter at     Thickness Plant                                                                           Pressure Studs          Diameter        Stud Hole     at Stud Hole (psia)
(inches)         (inches)       (inches)
FNP 58              6.0             167             15          2500 Range for 16 54- 60         6.0 - 7.0       157 - 173         15 -16        2500 Units Considered Bounding Values 54              6.0             173             16         2500 Used in Analysis The analytical model is shown in Figures 2 and 3. The loads considered in the analysis consisted of:
* A design pressure of 2500 psia at an operating temperature of 600&deg;F was applied to all internal surfaces exposed to internal pressure.
* Bolt/stud preload - The preload was calculated as detailed in FNP Unit 1 RPV manual. The preload on the bounding geometry is calculated as:
C
* P* ID 2    1.1* 2500* 1732 Ppreload= - - - -   2 =                     =42,338 psi S *D              54*62 where:
Ppreload      =        Preload pressure to be applied on modeled bolt (psi) p             =        Internal pressure (psi)
ID            =        Largest inside diameter of RPV (in.)
c             =        Bolt-up contingencies (+ 10%)
s             =        Least number of studs D             =        Smallest stud diameter (in.)
* Thermal stresses - The only significant transient affecting the bolting flange is heat-up/cooldown. This transient typically consists of a steady 100&deg;F/hour ramp up to the operating temperature, with a corresponding pressure ramp up to the operating pressure.
The ANSYS finite element analysis program was used to determine the stresses in the threads in flange-component for the three loads described above.
E2-3 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-ISl-ALT-19 Version 2.0 Flaw Tolerance Evaluation A flaw tolerance evaluation was performed using the results of the stress analysis to determine how long it would take an initial postulated flaw to reach the ASME Code, Section XI allowable flaw size. A linear elastic fracture mechanics evaluation consistent with ASME Code, Section XI, IWB 3600 was performed.
Stress intensity factors (Ks) at four flaw depths of a 360&deg; inside-surface-connected, partial-through-wall circumferential flaw are calculated using finite element analysis techniques with the model described above. The maximum stress intensity factor (K) values around the bolt hole circumference for each flaw depth (a) are extracted and used as to perform the crack growth calculations. The circumferential flaw is modeled to start between the 10th and 11th flange threads from the top end of the flange because that is where the largest tensile axial stress occurs. The modeled flaw depth-to-wall thickness ratios (alt) are 0.02, 0.29, 0.55, and 0.77, as measured in any direction from the stud hole. This creates an ellipsoidal flaw shape around the circumference of the flange, as shown in Figure 4 for the flaw model with alt= 0.77 crack model. The crack tip mesh for the other flaw depths follows the same pattern.
When preload is not being applied, the stud, stud threads, and flange threads are not modeled. The model is otherwise unchanged between load cases.
The maximum K results are summarized in Table 2 for the four crack depths.
Because the crack tip varies in depth around the circumference, the maximum K from all locations at each crack size is conservatively used for the K vs. a profile.
Table 2: Maximum K vs. alt Load                          K at Crack Depth (ksi"in) 0.02 alt 0.29 alt 0.55 alt 0.77 alt Preload                    11.2        17.4    15.5       13.9 Preload + Heatup + Pressure             13.0        19.8    16.1       16.3 The allowable stress intensity factor was determined based on the acceptance criteria in ASME Section XI, IWB-3610/Appendix A which states that:
K1 < K1cf"1 O = 69.6 ksi"in Where, K1 = Allowable stress intensity factor (ksi"in)
Kie= Lower bound fracture toughness at operating temperature (220 ksi"in)
As can be seen from Table 2, the allowable K is not exceeded for all crack depths up to the deepest analyzed flaw of alt =0.77. Hence the allowable flaw depth of the 360&deg; circumferential flaw is at least 77% of the thickness of the flange. The allowable flaw depth is assumed to be equal to the deepest modeled crack for the purposes of this analysis.
For the crack growth evaluation, an initial postulated flaw size of 0.2 in. (5.08 mm) is chosen consistent with the ASME Code, Section XI, IWB-3500 flaw acceptance E2-4 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-181-ALT-19 Version 2.0 standards. The deepest flaw analyzed is alt= 0.77 because of the inherent limits of the model. Two load cases are considered for fatigue crack growth: heat-up/cooldown and bolt preload. The heat-up/cooldown load case includes the stresses due to thermal and internal pressure loads and is conservatively assumed to occur 50 times per year. The bolt preload is assumed to be present and constant during the load cycling of the heat-up/cooldown load case. The bolt preload load case is conservatively assumed to occur five times per year, and these cycles do not include thermal or internal pressure. The resulting crack growth was determined to be negligible due to the small delta K and the relatively low number of cycles associated with the transients evaluated. Because the crack growth is insignificant, the allowable flaw size will not be reached and the integrity of the component is not challenged for at least 80 years (original 40-year design life plus additional 40 years of plant life extension).
The stress analysis I flaw tolerance evaluation presented above show that the threads in flange component at FNP is very flaw tolerant and can operate for 80 years without violating ASME Code, Section XI safety margins. This clearly demonstrates that the threads in flange examinations can be eliminated without affecting the safety of the RPV.
Operating Experience Review Summary As discussed above, the results of the survey, which includes results from FNP, confirmed that the RPV threads in flange examination are adversely impacting outage activities (dose, safety, increased time at reduced RCS inventory) while not identifying any service induced degradations. Specifically, for the US fleet, a total of 94 units have responded to date and none of these units have identified any type of degradation. As can be seen in Table 3 below, the data is encompassing. The 94 units represent data from 33 BWRs and 61 PWRs. For the BWR units, a total 3,793 examinations were conducted and for the PWR units a total of 6,869 examinations were conducted, with no service induced degradation identified. The response data includes information from all of the plant designs in operation in the US and includes BWR-2, -3, -4, -5 and -6 designs. The PWR plants include the 2-loop, 3-loop and 4-loop desi.gns and each of the PWR NSSS designs (i.e., Babcock
       & Wilcox, Combustion Engineering and Westinghouse).
Table 3: Summary of Survey Results - US Fleet Plant Type         Number of        Number of          Number of Units        Examinations         Reportable Indications BWR               33               3,793                0 PWR               61               6,869                0 Total             94              10,662                0 Related RPV Assessments In addition to the examination history and flaw tolerance discussed above, Reference 1 discusses studies conducted in response to the issuance of the E2-5 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-ISl-ALT-19 Version 2.0 Anticipated Transient Without Scram (ATWS) Rule by the USN RC. This rule was issued to require design changes to reduce expected ATWS frequency and consequences. Many studies have been conducted to understand the ATWS phenomena and key contributors to successful response to ATWS event. In particular, the reactor coolant system and its individual components were reviewed to determine weak links. As an example, even though significant structural margin was identified in USNRC SECY-83-293 for PWRs, the ASME Service Level C pressure of 3200 p'sig was assumed to be to an unacceptable plant condition.
While a higher ASME service level might be defensible for major Reactor Coolant System (RCS) components, other portions of the RCS could deform to the point of inoperability. Additionally, there was the concern that steam generator tubes might fail before other RCS components, with a resultant bypass of containment. The key take-away for these studies is that the RPV flange ligament was not identified as a weak link and other RCS components were significantly more limiting. Thus, there is substantial structural margin associated with the RPV flange.
In summary, Reference 1 identifies that the RPV threads in flange are performing with very high reliability based on operating and examination experience. This is due to the robust design and a relatively benign operating environment (e.g. the number and magnitude of transients is small, generally not in contact with primary water at plant operating temperatures/pressures, etc.). The robust design is manifested in that plant operation has been allowed at several plants even with a bolt/stud assumed to be out of service. As such, significant degradation of multiple bolts/threads would be needed prior to any RCS leakage.
Proposed Alternative and Basis for Use:
In lieu of the requirements for a volumetric ultrasonic examination, SNC proposes eliminating the requirement for the RPV threads in flange examination for FNP Unit 1.
SNC has confirmed that FNP plant specific parameters (e.g. vessel diameter, number of studs, inservice inspection findings) are consistent with or bounded by Reference 1.
Since there is reasonable assurance that the proposed alternative is an acceptable alternate approach to the performance of the ultrasonic examinations, SNC requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a(z)(1) on the basis that the use of the alternative provides an acceptable level of quality and safety.
To protect against non-service related degradation, each refueling outage SNC uses a detailed procedure for the removal, care and visual inspection of the RPV studs and the threads in flange. Care is taken to not only remove the studs, but once the studs are removed, to inspect the RPV threads for damage and to install RPV stud plugs to protect threads from damage. Prior to reinstallation, the studs and stud holes are cleaned and lubricated. The studs are then replaced and tensioned into the Reactor Vessel. This activity is performed each refueling outage and the procedure documents each step. These controlled maintenance activities provide further assurance that degradation is detected and mitigated prior to returning the reactor to service.
In addition, it is important to note all other inspection activities, including the system leakage test (examination category 8-P), which is conducted each refueling outage, will continue to be performed.
E2-6 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-ISl-ALT-19 Version 2.0 Duration of Proposed Alternative:
The alternative is requested for the current Fourth lnservice Inspection Interval, which began December 1, 2007 and is scheduled to end on November 30, 2017 for Unit 1.

==References:==
: 1. EPRI Nondestructive Evaluation Report - Reactor Pressure Vessel Threads in Flange Examination Requirements. 3002007626; Dated: March 2016 Electric Power Research Institute (EPRI).
: 2. American Society of Mechanical Engineers, Risk-Based Inspection: Development of Guidelines, Volume 2-Part 1 and Volume 2-Part 2, Light Water Reactor (LWR)
Nuclear Power Plant Components. CRTD-Vols. 20-2 and 20-4, ASME Research Task Force on Risk-Based Inspection Guidelines, Washington, D.C., 1992 and 1998.
Status:
Pending NRC approval.
E2-7 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-ISl-ALT-19 Version 2.0
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8.5" 17.0"                               7.0" 16.0" R83.75" R4.5"
: 10. 75" R85.69" Figure 1 Modeled Dimensions E2-8 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-ISl-ALT-19 Version 2.0
 '- El.EMNI'S REAL NUM ANO_Vessel_Flange Figure 2 Finite Element Model Showing Bolt and Flange Connection E2-9 to NL-16-1329 Proposed lnservice Inspection Alternative FNP-ISl-ALT-19 Version 2.0 Figure 3 Finite Element Model Mesh with Detail at Thread Location E2-10

Vogtle Electric Generating Plant, Units 1 and 2 Joseph M. Farley Nuclear Plant, Unit 1 Proposed lnservice Inspection Alternative VEGP-ISl-ALT-11 Version 2.0 and Proposed lnservice Inspection Alternative FNP-ISl-ALT-19, Version 2.0 Enclosure 3 EPRI Nondestructive Evaluation Report- Reactor Pressure Vessel Threads in Flange Examination Requirements. 3002007626; Dated: March 2016}}