ML21055A020

Transmittal of 2020 Owners Activity Report

ML21055A020
Person / Time
Site:	Ginna
Issue date:	02/19/2021
From:	Swift P Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
G1R42
Download: ML21055A020 (28)

Text

Exelon Generation February 19, 2021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 2055-0001

Subject:

RE. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244 Transmittal of 2020 Owner's Activity Report for the R. E. Ginna Nuclear Power Plant Paul M. Swift Site Vice President R.E. Ginna Nuclear Power Plant 1503 Lake Rd.

Ontario. NY 14519 315 7915200 Office www.exeloncorp.com paul.swift@exeloncorp.com Enclosed is a copy of the R. E. Ginna Nuclear Power Plant Owner's Activity Report for the refueling outage conducted in 2020 and a supporting Operability Evaluation. This report is submitted as specified by ASME Code Section XI.

There are no regulatory commitments contained in this letter.

If you have any questions, please contact Kyle Garnish at (585) 469-2837.

Respectfully, Paul M. Swift : FIFTH INSERVICE INSPECTION (ISi) INTERVAL and SECOND CONTAINMENT INSPECTION INTERVAL (GISI), THIRD INSPECTION PERIOD (ISi AND GISI); SIXTH INSERVICE INSPECTION (ISi) INTERVAL and THIRD CONTAINMENT INSPECTION INTERVAL (CISI), FIRST INSPECTION PERIOD (ISi AND GISI) 2020 OWNER'S ACTIVITY REPORT FOR RF0-42 INSERVICE EXAMINATIONS (April 6, 2020 through May 23, 2020) : OPEVAL-20-002, Rev 001 (redacted; non-proprietary) cc:

NRC Regional Administrator, Region 1 NRC Project Manager, Ginna NRC Resident Inspector, Ginna

ATTACHMENT 1 FIFTH INSERVICE INSPECTION (ISi) INTERVAL AND SECOND CONTAINMENT INSPECTION INTERVAL (CISI),

THIRD INSPECTION PERIOD (ISi AND CISI);

SIXTH INSERVICE INSPECTION (ISi) INTERVAL AND THIRD CONTAINMENT INSPECTION INTERVAL (CISI),

FIRST INSPECTION PERIOD (ISi AND CISI) 2020 OWNER'S ACTIVITY REPORT FOR RF0-42 INSERVICE EXAMINATIONS (April 6, 2020 through May 23, 2020)

RE. Ginna Nuclear Power Plant, LLC February 19, 2021

FORM OAR-1 OWNER'S ACTIVITY REPORT Report Number G 1 R42

~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~

Plant R.E. Ginna Nuclear Power Plant, 1503 Lake Road. Ontario New York 14519 UnilNo.

NIA Comtnereial Sef'lice Date G1R42

___ J""'u'--n_e_1..... _19_7_0 _ __ Rerueting Outage Number (if applle:allle)

Fifth Inspection Interval (ISi), Second lnspeclion lnteNal (CISI) [Through December 31, 20191 Current Inspection Interval Sixth Inspection Interval (fSI), Third lnsP!ctiOn Interval (CISI), [Starting January 1, 2020]

{1". 2"". 3". 4'11, other)

Third Inspection Period (ISi and CISI) [Through December 31, 2019):

Current Inspection Period first Inspection period (ISi and CISI) [Starting January 1, 2020J (1", 2"". 3"' I Edition and Addenda of Secliot'I XI applicable to lne Inspection plans ASME Section XI 2004 Edition (Siil intervai)

ASME Section XI 2013 Edition (61.., Interval)

Date ancl RevisiOn of inspeclion plans December 12, 2{) 19, Rev. O

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

Edition and Addenda of Section XI applicabla to tepair/replacemenl activities. Ir different tttan the iospeciion plans Code Ca$es used for inspection and evaluation:

N-532-5, N-706, N-71&-1, N-722*1. N-729-4. N-731. N-770*2 Same as above (11 eippllcallle, lncludlng ceses modi1iad by ca&e N-532 and lilt.er nrvisions)

CERTIFICATE OF CONFORMANCE I certify tnat (a) the $Utl.ements made In lhis report are correct: (b) the examinations and tests meet the Inspection Plan as required by the ASME Code,Section XI; and (c) O'le repair/replacement activilles and evalumns supporting lhe completion of G 1 R42 (refueling outage l"un'lt:ar) conform to the reqvirents of Section XI, s~gned

.,,....- +-~-<-...~ 'j,L 1 Fleet tn-s..ivice Inspection Engineer Date 1112012020 i._~)

ignaa. TiUe}

CERTIFICATE OF lNSERV1CE INSPECTION I, the undersigned, holding a valid commission iSslJed by the National Board or Boiler and Pl'8$$ure Vessel lnspeet.onl and employed by The Hartford Steam Boller Inspection and Insurance Company of HartfOtd, Connecticut have inspected the Items described in lhl$ Owttel's Activity Report, and stale 1tlal, to the best of my knowledge and belief, the Owner nas performed all activities repreaented by this report in accordance with the requirements or Section XI.

By signing this certificate neither Ile Inspector nor his employer makes any warranty. expressed Of implied. concemlng the repalr/replacement activilbe$ and evaluation described In this report Furthermore, neither the Inspector nor his employer shall be liable in any manner any persor181 i ry or property damage or a loss or any kind arising rrom or col"lnected with this inspection.

Dale Page 1of3

Examination Category B-N-3 TABLE 1 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Examination Item Item Description Evaluation Description Number Core Support Structure (270 IR 04335182 813.70 Degree Vessel Side Lower Radial Was evaluated and found acceptable by Clevis Insert) was found disengaged and radially displaced.

OPEVAL-20-002 Page 2 of 3

TABLE 2 ABSTRACT OF REPAIR/REPLACEMENT ACTIVITIES REQUIRED FOR CONTINUED SERVICE Code Item Description Date Repair/Replacement Class Description Of Work Completed Plan Number 2

758A Like for Like replacement of valve 758A 06/10/2020 GORR42-065 to address minor valve leakage WO# C93749626 Page 3 of 3

ATTACHMENT 2 OPEVAL-20-002, REV 001 (REDACTED; NON-PROPRIETARY)

R.E. Ginna Nuclear Power Plant, LLC February 19, 2021 Operability Evaluation Page 1 of 21 1.0 ISSUE IDENTIFICATION 1.1 IR#:

04335182 1.2 OpEval #: OPEVAL-20-002 Revision:....;:0:....::0....:.1 __ _

1.3 EC Number ---'"N""'"/'""'A"------

Revision: N/A General Information:

1.4 Affected Station(s): ___

R'--.E_._G'--i_nn_a~------

1.5 Unit(s)

__:.1 __

1.6 System

....;:0""'2'------'-'R=e=a-=-ct=o-'--r"'""C'"""o"""o"""'la=n=t"'""S"""y-=-s=te"""m'-'-------

1.7 Component(s) Affected: -~L~o....:.w~e~r-'--R""'a~d~ia~l....;:S~u~o~PO~rt~

1.8 Detailed description of what SSC is degraded, nonconforming or unanalyzed condition, by what means and when first discovered, and extent of condition for all similarly affected SSC's:

On April 13, 2020 at 08:55, during the ten-year In-Service Inspection (ISi) of the Ginna reactor vessel Lower Radial Support System (LRSS) relevant conditions were discovered at the 270° position LRSS Clevis Insert. The following conditions were observed:

A. A gap was identified on the back side between the Clevis Insert and the Clevis.

B. There was evidence of wear on the faces between the Clevis Insert and the Clevis.

C. The Clevis Insert had worn into the top face of the Clevis displacing downward.

D. Evidence of wear and contact between the Clevis Insert and the Core Barrel Key.

Follow up interrogation of the Clevis Insert revealed that it had lost some of the originally installed interference fit within the Clevis Block. No relevant conditions were found at the other three LRSS Clevis Inserts or any of the four Core Barrel Keys. No additional Extent of Condition exams were required as all relevant components were examined during the 2020 RFO. The 210* Core Barrel Key showed signs of vibration wear at the top consistent with the wear marks seen on the 210*

Clevis Insert. The Clevis Insert has been repositioned radially towards the Clevis to provide the required radial clearances to allow insertion and installation of the Reactor Vessel Internals. The failed bolting and dowel have been left in place, captured within the Clevis Insert. Approximately 15 Kips of horizontal force was required to reposition the Clevis Insert indicating that in the as-left position there was some nominal interference fit.

The LRSS is designed with four Key and Keyway joints. At equally spaced points around the circumference, an lnconel block (Clevis) is welded to the vessel l.D. Another lnconel block (Clevis Insert) with Keyway geometry is bolted into each Clevis with a 0.002" interference fit and alignment dowel. Opposite each of these is a Key which is attached to the core barrel by cap screws and J-groove welds. As the internals are lowered into the vessel, the Keys engage the Clevis Insert in the axial direction. With this design, the internals are provided with a support at the furthest extremity and may be viewed as a beam fixed at the top and simply supported at the bottom.

Radial and axial (vertical) expansions of the core barrel are accommodated but transverse (tangential) movement of the core barrel is restricted by this design. With this system, normal operation and accident condition stresses remain within design limits.

Operability Evaluation Page 2 of 21 Original design loads included normal operation loads, thermal transient loads, LOCA loads, seismic loads, barrel insertion loads, and associated cyclic load considerations. The Clevis Insert was originally secured in the Clevis Block with an interference fit, eight bolts and one dowel pin.

The 8 bolts failed AND the interference fit was compromised at the 270° Clevis Insert. The bolt heads remain secured in place with anti-rotation tack welded locking-bars. The dowel pin remains secured in place with tack welds. Wear was identified on the tack weld surfaces for the 270° Clevis due to contact with the Core Barrel Key, but no tack welds were found broken and no wear was identified on the lock bars. As the Insert has not been removed and no attempt has been made to remove the bolting, the conclusion that the bolts have broken is based on the observed gap between the Clevis Insert and Clevis in addition to evidence that the bolt heads remain locked in place.

This operability evaluation provides justification for restart from the 2020 RFO and continued operation in Modes 1-6 with the following conditions in the 270° position Clevis Insert:

A. 8 QTY 5/8" cap screws left in as-found condition B. 1 QTY dowel pin left in as-found condition.

C. Loss of original interference fit between the Clevis Insert and Clevis.

2.0 EVALUATION

Operability Evaluation Page 3 of 21 2.1 Describe the safety function(s) or safety support function(s) of the SSC. As a minimum the following should be addressed, as applicable, in describing the SSC safety or safety support function(s):

Does the SSC receive I initiate an RPS or ESF actuation signal?

No.

Is the SSC in the main flow path of an ECCS or support system?

Yes. The LRSS is in the down comer between the vessel wall and core barrel. Under all normal operation conditions flow is downward onto the Clevis Insert. This includes single Reactor Coolant Pump operation which may result in flow at a downward angle (horizontal and downward vertical vector components). During accident conditions involving a Loss of Coolant Accident from the cold leg, flow could be upward past the Clevis Insert.

Is the SSC used to:

Maintain reactor coolant pressure boundary integrity?

No, the Clevis Insert does not directly provide or maintain the RCS pressure boundary. The Clevis Insert transfers loads from the core barrel Key to the LRSS Clevis and the LRSS Clevis is welded to the Reactor Vessel pressure boundary. Loads on the Clevis Insert are designed to be transmitted to the Reactor Vessel pressure boundary.

Shutdown the reactor?

Yes, the LRSS has the potential to impact the ability of the control rods to insert during design basis events due to the allowed motion and distortion of the Core Barrel. In addition, the LRSS provides core support to ensure the fuel assemblies are maintained in a coolable geometry. The LRSS is also relied upon to ensure other reactor internals remain available for their various functions during design basis events.

Maintain the reactor in a safe shutdown condition?

No.

Prevent or mitigate the consequences of an accident that could result in offsite exposures comparable to 10 CFR 50.34(a)(1), 10 CFR 50.67(b)(2), or 10 CFR 100.11 guidelines, as applicable.

Yes, the LRSS has the potential to impact the ability of the control rods to insert due to the allowed motion and distortion of the Core Barrel. In addition, the LRSS provides core support to ensure the fuel assemblies are maintained in a coolable geometry.

Operability Evaluation Page 4 of 21 Does the SSC provide required support (e.g., cooling, lubrication, etc.) to a TS required SSC.

Yes, the LRSS Clevis Insert provides support to the Reactor Vessel internals and associated components by limiting stresses and deflections to ensure the Fuel Assemblies remain intact, Flux Detectors can be inserted into the core, the Control Rods can be freely inserted, and that the core geometry remains coolable.

Is the SSC used to provide isolation between safety trains, or between safety and non-safety ties?

No.

Is the SSC required to be operated manually to mitigate a design basis event?

No.

Have all specified safety functions described in TS been included?

Yes, the Technical Specifications have been reviewed and determined that no described safety functions are attributed to the LRSS. The LRSS is required to resist LOCA loads in Modes 1, 2, 3, and 4, and seismic loads in Modes 1, 2, and 3 operation. The required function of the LRSS is to limit Reactor Vessel Internals motion and stresses as the result of vibrations, seismic and accident conditions (Ref. 2).

TS 3.1.4 - Rod Group Alignment Limits TS 3.1.4 ensures that all control rods remain operable. The LRSS shall be capable of limiting the Reactor Vessel Internals deflections to ensure control rods remain insertable.

TS 3.2.1 - Heat Flux Hot Channel Factor (FQ(Z))

The purpose of the limits on the values of FQ(Z) is to limit the local (i.e., pellet) peak power density. Flux mapping is required to measure (FQ(Z)). The LRSS shall be capable of limiting the Reactor Vessel Internals deflections such that Flux mapping may be performed to support this LCO.

TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor (FN(H))

The purpose of this LCO is to establish limits on the power density at any point in the core so that the fuel design criteria are not exceeded, and the accident analysis assumptions remain valid. Flux mapping is required to measure power density in the core. The LRSS shall be capable of limiting the Reactor Vessel Internals deflections such that Flux mapping may be performed to support this LCO.

TS 3.3.1 - Reactor Trip System (RTS) Instrumentation The RTS initiates a plant shutdown to protect against violating the core fuel design limits and RCS pressure boundary during anticipated operational occurrences and to assist the Engineered Safety Features (ESF) Systems in mitigating accidents. Flux mapping is Operability Evaluation Page 5 of 21 required to compare to NIS Axial Flux Difference. The LRSS shall be capable of limiting the Reactor Vessel Internals deflections such that Flux mapping may be performed to support this LCO.

TS 3.5.2 - ECCS Modes 1.2.3 The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after any of the following accidents:

a. Loss of coolant accident (LOCA) and coolant leakage greater than the capability of the normal charging system;

b. Rod ejection accident;

c. Loss of secondary coolant accident, including uncontrolled steam release or loss of feedwater; and

d. Steam generator tube rupture (SGTR).

The LRSS shall ensure a coolable core geometry is maintained for these events, such that the ECCS system can meet its design function.

TS 3.5.3 - ECCS Mode 4 The function of the ECCS is to provide core cooling and negative reactivity to ensure that the reactor core is protected after a Mode 4 LOCA. The LRSS shall ensure a coolable core geometry is maintained for this event, such that the ECCS system can meet its design function.

TS 3.4.4 - RCS Loops Mode 1 > 8.5% RTP; TS 3.4.5 - RCS Loops Mode 1 < 8.5% RTP,

2. and 3; TS 3.4.6 - RCS Loop Mode 4; TS 3.4.7 - RCS Loops Mode 5. Loops Filled; TS 3.4.8 - RCS Loops Mode 5. Loops Not Filled The function of the RCS Loops in all modes is to provide flow through the core. The LRSS is required to limit motion of the Core Barrel to provide the RCS flow path through the Reactor Vessel.

Have all safety functions of the SSC required during normal operation and potential accident conditions been included?

Yes, the safety functions of the LRSS are evaluated to be satisfied during all modes of operation. The LRSS is required to provide support of the Reactor Vessel Internals and limit their deflections and applied stresses. These functions are evaluated in detail within Ref. 2.

Is the SSC used to assess conditions for Emergency Action Levels (EAL's)?

No.

Operability Evaluation Page 6 of 21 The LRSS performs both safety and non-safety related functions related to the Reactor Vessel and Internals as described in PWROG-15034-P Rev. 0 (Ref. 3). These functions are described below.

Safety Related Functions:

Restrict excessive lateral (tangential) and rotational motion of the Reactor Vessel Internals which can result during normal operation from Flow Induced Vibrations (FIV), and accident conditions from seismic events and Loss of Coolant Accidents (LOCAs).

Ensure displacement and misalignment is limited in order to avoid overstressing the Core Barrel components, interfacing components (including fuel assemblies), and ensure free insertion of the Control Rods.

Provide a structural interface to transmit lateral vibrations and dynamic loads from the Core Barrel structure to the rigid Reactor Vessel.

Non-Safety Related Functions:

Ensures proper alignment and fit-up of the Reactor Vessel Lower Internals during assembly Provides radial and vertical clearances to allow for the relative thermal expansion of the Reactor Vessel Internals Ensures the accurate positioning necessary for engagement of the bottom instrument columns with the lower vessel instrumentation penetrations for installation and mating. Mating of these components is required to ensure a continuous, smooth pathway for introducing the incore flux detectors into the fuel assemblies. Once this mating occurs during assembly, the safety related function to restrict motion ensures the continued use of the flux detectors.

Safety Related Loads (see Figure 1 ):

Cyclic tangential vibration resulting from flow induced vibrations Static tangential loading resulting from flow and pressure distributions across the Reactor Vessel internals Dynamic Faulted Conditions resulting from Operating Basis Earthquakes (OBE)

Faulted Conditions resulting from Safe Shutdown Earthquakes (SSE) and LOCAs Vertical and Radial loads are transferred to the Clevis Insert via friction from the Core Barrel Key during all tangential loads. Seismic loading does not apply in Modes 4, 5, and 6 in accordance with the Safety Evaluation Report for License Amendment No. 139 (Ref. 4) and LOCAs do not apply in Modes 5 and 6 in accordance with Technical Specifications LCO 3.5.2 and 3.5.3.

Safety Related loads occur simultaneously with non-safety related loads discussed below and are considered together to ensure the LRSS remains capable of meeting its design requirements.

Non-Safety Related Loads (see Figure 1 ):

Radial and Vertical frictional loading from the thermal growth of the Reactor Vessel Internals Downward vertical loading from Reactor Vessel Internals insertion Operability Evaluation Page 7 of 21 aevis Oevis Insert Tangenti_

al~---~

Radial Radial Key Oevis Insert Oevis l

Vertical Figure 1: Load Directions Describe the following, as applicable:

(a) the effect of the degraded or nonconforming condition on the SSC safety function(s)

The Clevis Insert has been repositioned radially towards the Clevis to provide the required radial clearances to allow insertion and installation of the Reactor Vessel Internals. The failed bolting and dowel have been left in place, captured within the Clevis Insert. The as-left circumferential clearance of the Clevis Insert is such that an additional 0.001" to 0.002" of transverse motion beyond the nominal design clearances may occur due to wear of the interference fit. The radial inwards motion of the Insert is restricted with the Internals installed and is prevented from unseating due to the engagement with the Core Barrel Key. A gap exists between the top of the Clevis and bottom of the Clevis Insert lip. The effect of these conditions on each of the LRSS functions is described below.

Allowance for Differential Thermal Expansion:

The degraded condition of the Clevis Insert does not impact the ability of the LRSS to allow for differential Radial, Vertical, and tangential thermal expansion of Reactor Vessel and Internals.

Specifically, the as-designed clearances are either unchanged or expanded (tangential direction).

Thermal growth loads are transmitted to the Insert via frictional forces applied between the Stellite wear surfaces. The Clevis Insert was reseated into the Clevis and was left sitting 0.25" proud of the Clevis front surface (See Figure 2). This provides nominally 0.25" radial clearance to the Operability Evaluation Page 8 of 21 Core Barrel Key and provides sufficient clearance for thermal growth without further seating the Clevis Insert.

Figure 2: As-Left Clevis Insert Position The Westinghouse operability evaluation letter (Ref. 2) evaluated thermal loads and found that faulted loads are bounding. Loads are transferred to the Clevis Insert during thermal transients by way of friction between the Radial Key and the Clevis Insert. If interference exists, heatup will produce loading to push the Insert radially out and down. During cooldown the Insert will be pulled in radially and up. The vertical shrinkage of the core barrel during cooldown is insufficient to pull the Insert out of seatin~a of the Clevis because the maximum core barrel shrinkage is less than the seating depth of *. Therefore, faulted conditions (LOCA + SSE) are bounding for this OpEval.

Frictional loading between the Stellite faces will result in the Insert becoming further seated into the Clevis. This does not impact the load carrying capability or the ability of the Insert to accommodate the thermal growth. This does however create a potential for a Loose Parts concern that is addressed in a later section below.

The Clevis Insert is capable of carrying all thermal growth loads due to differential thermal growth by way of bearing on the Clevis which is consistent with the original design.

Wear of the LRSS:

Since the 270° Clevis Insert has less interference than original design in the as-left configuration, movement of the Insert vertically, radially, and tangentially is expected with operational loading.

In addition, the wear from the 270° Insert will cause increased loading and potential subsequent wear on the 90° Insert. This is described in the Westinghouse Operability Letter (Ref. 2).

Operability Evaluation Page 9 of 21 The degraded 270° Clevis Insert was modeled by Westinghouse using a FEA (Finite Element Analysis) RESM (Reactor Equipment System Model) for a similar 2-Loop Westinghouse PWR (Plant A). The model from Plant A was utilized because the Ginna RESM was no longer available (Ref. 2). Relevant plant parameters were compared and determined that Plant A reasonably represents the loading conditions at Ginna. Differences were reconciled in the Westinghouse Letter (Ref. 2) and by Ginna within this OpEval and are not a concern for the purposes of this OpEval. An MPR evaluation (Ref. 6) was also performed for the structural impacts of the vessel, barrel and lower core support plate using bounding loads from Westinghouse. This evaluation demonstrated that displacements and stresses were reasonably expected to be acceptable given the loads provided. The Westinghouse analysis (Ref. 2) performed 4 different analytical cases regarding wear conditions:

Nominal gaps at all Insert locations ("Nominal"), 0.004-inch on either side of the Insert Moderate wear at the 270° Insert (0.039-inch gap on each side of Insert}, remaining Inserts with nominal gaps ("Intermediate")

The 270° Insert removed, remaining Inserts with nominal gaps ("No Clevis")

The 270° Insert removed, remaining Inserts with large (0.070-inch) gaps ("Max Wear")

Loss of Coolant Accident (LOCA) and Safe Shutdown Seismic Event (SSE) loads are the bounding loads on the Clevis Insert (Ref. 2). The analysis performed by Westinghouse evaluated these loading conditions for each Clevis condition cited above. The evaluation and results of the analysis are described in the following sections. The conservatism of the above listed Insert wear cases is discussed below by consideration of the expected 270° Insert wear.

The "Max Wear" case addressed the potential for significant wear of the other three Clevis Inserts (0° I go*

I and 180°). The wear modeled by Westinghouse is significantly greater than what was observed on the existing Clevis Insert, where the existing wear was estimated to be between 0.001" and 0.002". The most likely time for the radial displacement was a thermal transient that resulted in shrinkage of the core barrel (plant cooldown).

For the displacement to occur, the frictional load imparted by the Key on the Clevis Insert is required to be greater than the combined interference fit and friction load between the Insert and the Clevis. The PWROG document (Ref. 3) predicted this condition could occur during plant cooldown where temperature differences could cause a decrease or loss of interference fit.

The heat transfer between the Insert and the Clevis, and between the RCS cold leg water and the former components will be very similar. The main difference is that the Clevis is integrally welded to the Reactor Vessel shell which serves as a heat sink, resulting in the Clevis taking longer to thermally expand and contract. Additionally, the Insert has a greater surface area to volume ratio which provides for greater heat transfer to the RCS cold leg water. During heatup this serves to increase the interference fit and during cooldown decrease the interference fit.

Therefore, it is more likely for this event to have occurred when the core barrel cooled at a more rapid rate than the vessel, i.e. either during a load reject/drop or a plant trip. During a normal plant shutdown, the cold leg temperature changes relatively slowly and is not expected to result in a significant loss of interference. Rapid cooldown transients are expected to cause significant radial and vertical loads that can pull the Insert away from the Clevis. These loads can be large enough to cause a loss of interference fit and retention load. For conservatism it is assumed that the Insert was displaced in June of 2019 during the last plant shutdown prior to the G1R42 refueling outage in April of 2020, but likely occurred during a more rapid plant transient.

With the as-found maximum of 0.002" of tangential wear over 9-months, a linear wear rate results in an additional 0.004" of wear over the next 18-month fuel cycle. This estimate is conservative, because as the 270° Insert gap opens due to wear, the tangential load is transferred to the opposite Insert (90°) as a result of the Lower Core Support Plate stiffness. The frictional radial Operability Evaluation Page 10 of 21 loads proportionally decrease causing less movement and less wear of the Insert. Considering the wear observed vertically (between the top of the Clevis and bottom of the Clevis Insert flange) was much more substantial, it is likely that the frictional wear in the tangential direction is already becoming self-limiting. Therefore, the use of this linear wear rate for another 18-months is conservative.

Further discussion of the operating loads and the expected wear during the operating cycle is provided in the following section.

If the other 3 Clevis Inserts wear at a rate of double that of the 270° Clevis, it would take approximately 157 months, or 13.125-years for the limiting condition analyzed in Ref. 2 to occur.

This provides assurance that wear on the other Inserts will be minimal for one 18-month cycle.

Ref 2 evaluated gap conditions that are considered conservatively large given the prior discussion.

Operating Loads While SSE and LOCA loads are the limiting loads with respect to acceptability of Clevis Insert gaps, operating loads will drive the continued wear and degradation. A discussion of the operating condition is provided to support the wear discussion in the previous section. Particularly, Ref. 2 states that the unrestrained condition of the LRSS was evaluated during dynamic testing in 1972.

This testing determined a -

deflection at the Lower Radial Key for a 4-Loop plant. Since the 2-Loop plants were found to have* th~eflections, the nominal unrestrained Lower Radial Key deflection is approximated as --* Therefore, normal operating conditions should be expected to result in tangential deflection of up to -

to either side of the installed neutral axis.

As the Clevis Insert tangential gap continues to open, the applied tangential load will decrease and the wear rate will slow. Even under conditions where the 90° Insert begins to wear, the tangential wear can reasonably be concluded to slow and arrest as the total gap (as-built+ wear) approaches the unrestrained deflection.

This supports the previous wear discussion that concluded the conditions analyzed for the bounding SSE and LOCA loads are conservative with respect to the expected conditions for the 18-month operating cycle.

The SSE and LOCA cases are evaluated in the following sections. These loads are reduced from those used in the EPU calculations. This reduced loading is acceptable in reference to NSAL 2 which documents a reduction of stiffness in the LRSS. This reduced stiffness has been modeled at Ginna with respect to the Baffle Plates and associated bolting and found to reduce the applied loading. The loads used in Ref. 2 are those determined for Plant A after application of NSAL 2 and are consistent with loads expected at Ginna following a revision of the RESM.

LOCA Loads:

The design basis LOCA loading conditions are a break of the Pressurizer Spray Scoop, PSS, (3" break on the cold leg) and a break of the 4" Upper Plenum Injection, UPI, line (Ref. UFSAR 3.9.2.3.4 Design Basis LOCA based on Leak Before Break). Westinghouse analyzed both UPI and PSS breaks for varying Clevis conditions as described above. The impact loads associated with the PSS break were shown to be bounding in Reference 2 even though the UPI break size is larger than the PSS break. Note that the Ginna UPI break size is larger than that analyzed at Plant A. Since the UPI line is shown in Ref. 2 to be equivalent to the PSS break or limiting for some loads, the potential increased load is assessed in the comparison to previously analyzed loads.

The Westinghouse evaluation shows the loads associated with worn Inserts compared to the nominal (as designed) case.

Reference 2 evaluated reactor internals at 21 area/direction combinations and concluded all locations were acceptable. The largest increase in loading was for the "Max Wear" case with the majority of load increases less than a factor of three, and many Operability Evaluation Page 11 of 21 with almost no increase. Table 1 provides the interaction ratio (Ref. 2 Load/Previous analyzed Load) for Reactor Internals components impacted by the wear. The largest load increases (>2) from both the PSS and UPI break are evaluated to ensure they remain bounded by previous analysis. The UPI break loads are increased by an additional 35% to account differences in UPI break size between Ginna and Plant A based on ratio of flow areas. All interaction ratios consider the previously analyzed and accepted load for EPU except for the RPV Lugs. Additional margin is available for the previously analyzed load and a high interaction ratio is not representative of a limiting condition. The load used for the RPV Lug comparison is the maximum allowed from the EPU calculation.

These increases are for the Clevis Insert removed and the maximum wear (-0.07") applied to the remaining Inserts. As discussed in the previous sections, this extent of wear is not expected to occur. For the more likely and representative intermediate condition, all load increase factors were negligible except for the Lower Radial Key (<2) and the Upper Core Plate Pin (no change).

This demonstrates there is significant margin to the previously analyzed and allowable loads for the expected conditions of the 18-month operating cycle.

a e n erac ion a 10s or mpac e T bl 1 LOCA I t f

Rf f

I tdC t

omponen s Component Interaction Ratio Load Reference RPV Outlet (Horizontal) 0.45 CN-RCDA-05-20 for previously evaluated load RPV Inlet (Horizontal) 0.93 CN-RCDA-05-20 for previously evaluated load RPV Lug (Horizontal) 0.53 CN-RCDA-05-35 for maximum allowable load Lower Radial Key (Circumferential) 0.47 CN-RCDA-04-102 for previously evaluated load Guide Tube Pin 0.13 WCAP-17367-P for previously evaluated load Upper Core Plate Pin (circumferential) 1.8 CN-RCDA-04-102 for previously evaluated load Guide Tube Flanoe N/A N/A Plant analyses (Ref. 13) evaluate loading on the Guide Tubes for the limiting case where deflections interfere with RCCA insertability. Ref. 2 provides an applied moment for the Guide Tube Flange that is not directly reviewable against Ref. 13. However, if the deflection of the Guide Tube is less than the maximum allowed deflection, all Guide Tube loading will remain acceptable.

Ref. 2 provides the Guide Tube deflection for the evaluated Insert wear cases and determine the deflection does not exceed that allowed to ensure RCCA insertability. Therefore, the moment applied to the Guide Tube Flange is acceptable.

Ref. 2 specifies that the increased loading on the Upper Core Plate Alignment Pins results in yielding of the pin. Yielding is demonstrated to be acceptable in Ref. 2 due to core plate rim contact with the core barrel shell which has a much higher allowable load. Review of CN-RCDA-04-102 (Ref. 5) provides a pre-EPU LOCA load in excess of the increased load at "Max_ Wear".

The pre-EPU loading was evaluated as acceptable in WCAP-14460 (Ref. 16). While pin yielding Operability Evaluation Page 12 of 21 may represent a commercial concern, there is no safety impact as the core plate rim will carry the loading from this accident condition. Additionally, pin yielding will not impact RCCA insertability as the gap between the upper core plate rim and core barrel (-0.07") is small compared to the allowable displacement of the Guide Tubes.

Reference 2 also calculated displacements for the Reactor Internals. The displacements for the "Max_Wear" condition remain within acceptable limits as shown in Ref. 2.

Displacements assuming the remaining Inserts have the "nominal" gap condition ("Nominal", "Intermediate", and "No_Clevis") are essentially unchanged. Only when the remaining Inserts are worn to an extreme condition (-0.07") do the displacements change significantly. This implies that no significant change in Internal component displacements will occur until the remaining Inserts begin to experience appreciable wear. The minimal as-found wear of the 270° Insert provides assurance that appreciable wear will not occur on the remaining Inserts over the next operating cycle.

Seismic Westinghouse evaluated the seismic effects for the same varying Clevis Insert conditions.

Detailed results are available using Safe Shutdown Earthquake (SSE)1 acceleration response spectra (ARS) for the same similar plant (Plant A) evaluated above. To validate the reasonability of the response spectra, Westinghouse applied the worst worn condition to the model (270° Insert removed, maximum wear on all other Inserts) using both the Ginna ARS and the Plant "A" ARS.

The load magnitude for both ARS are were then compared.

There are changes in the impact loads in the reactor between the two input spectra, most notably for the reactor vessel supports, fuel bottom nozzle, and the control rod guide tube support pin.

The balance of changes due to the different ARS were quite small. The areas with increased loads were reviewed by Westinghouse and confirmed to remain acceptable. Given the relatively small changes in loading between the two ARSs and the acceptability of differences, use of Plant A is a reasonable method to determine the loads associated with Ginna. The "Max_ Wear" case is the bounding case and is used for comparison to allowable stresses and deflections.

Comparison of Seismic loading for other wear cases is not relevant to the conclusions contained in Ref. 2.

Interaction ratios between previously evaluated loads and the SSE loads assuming "Max-Wear" are shown in Table 2. The Guide Tube Pin was compared to the previously evaluated OBE load.

The Lower Radial Key was evaluated against the OBE load contained in CN-RCDA-05-35 (Ref.

11) for the Reactor Vessel Support Lug. This load is <1/2 the SSE load reported in CN-RCDA-04-101 (Ref. 8) and represents a conservative lower bound OBE. Since Ref. 2 determined SSE loads and does not provide OBE loads, the comparison in Table 2 uses OBE loads to further confirm the Upset condition is bounded for the "Max_ Wear" condition. As applicable, the reported load increase factors for the SSE results were increased by the ratio (Ginna "Max_Wear"/Plant "Max_ Wear"). For example an additional load increase factor of 1.125 is used for the Guide Tube Pin.

T bl 2 SSE I a e nteract1on R.

f I

at1os or mpacte dC omponents Component Interaction Ratio Load Reference Lower Radial Key (Circumferential) 0.75 CN-RCDA-05-35 for previously evaluated load 1 SSE loads are higher than Operating Basis Earthquake (QBE) loads, and allowable load limits are based on the QBE allowable limits; therefore, SSE loads are bounding.

Guide Tube Pin Guide Tube Flange Operability Evaluation Page 13 of 21 0.56 N/A WCAP-17367-P for previously evaluated load N/A The SSE moment applied to the Guide Tube Flange is less than the moment applied under LOCA conditions. Therefore, the Guide Tube Flange loading is bounded by the previous LOCA case and remains acceptable in accordance with Reference 13.

The Guide Tube deflections in Reference 2 for the SSE condition do not exceed the allowed deflections to ensure RCCA insertability and the SSE loading is acceptable.

For the SSE cases, displacement does not change for any case where the remaining Inserts are assumed to have a "nominal" (as-designed) gap. The change in displacement during SSE conditions is likely to progress from no-change to the worst-case condition as the remaining Inserts begin to wear. The minimal as-found wear on the 270° Insert provides assurance that appreciable wear will not occur on the remaining Inserts during the 18-month operating cycle.

LOCA and SSE - Grid Deflections The dynamic analysis is used to provide core plate motions for evaluation of the fuel and determine if grid deformation is expected to occur with the applied loads. Regarding core plate motion and grid deformation, Reference 2 states the Westinghouse dynamic analysis code WE GAP was utilized to determine grid impact forces. The LOCA and SSE loads were combined by Square Root Sum of Squares (SRSS) and compared against allowable loads determined by testing.

The Ref. 2 analysis confirms no grid deformation will occur under the applied loads considering the worst-case wear conditions and therefore the core will be maintained in a coolable geometry for the 18-month operating cycle.

Additionally, Ref. 2 evaluated applied loading to the thimble tubes which provide the structural capacity of the fuel assemblies and passage for the insertion of the RCCAs. The analysis concludes that the acceptable fuel assembly deflections occur under all wear conditions and plastic deformation of the thimble tubes and binding of the RCCAs is not a concern for the 18-month operating cycle.

Loose Parts Loose parts can occur due to wear of the Key directly on the Insert Locking welds, by the Insert being pushed into the bolt remnants such that the locking bars break, or by the Insert falling out of the Clevis. Westinghouse performed an evaluation of the potential for loose parts in the Operability Evaluation Letter (Ref. 2) which is described below.

Cap Screw Head If a cap screw fails, the lock bar can, over time, wear and separate, causing the cap screw head to be loose within the counterbore. The cap screw head would remain captured in the counterbore because the height of the head is larger than the radial gap between the Clevis Insert and the lower radial Key. Over a long period of time, wear of the head may reduce the height of the head to the size of the gap. The cap screw head wear is expected to be small because the cap screw material is much harder than the material it wears against (lnconel X-750 against Type 304 stainless steel).

Cap Screw Remnant Operability Evaluation Page 14 of 21 Industry operating experience (TB-19-5, Ref. 15) demonstrates that cap screw remnants can back-out during normal operating conditions and result in binding of the internals. If sufficient gaps exist, the remnant can also become a loose part. Considering the PWSCC failure location for the cap screws is at the head to shank region, the length of the remaining remnant is -1.6.

This remnant length is greater than the radial clearance (-0.5") between the Insert and the Key and cannot become a loose part. Continued back-out can result in bearing between the cap screw head the remnant during outward radial loading and may result in failure of the lock bar welds. The locking bars are evaluated below for loose part impacts.

Binding of the Lower Internals on the protruding remnant and or cap screw head is not a safety concern. This condition will result in reduced core barrel motion and lower loads across the Reactor Vessel Internals as demonstrated in Ref. 2. The binding condition will likely drive wear on the remnant and/or cap screw head which will eventually open the clearance and restore normal vibratory operation.

Dowel Pin Assuming the dowel pin welds fail, and the interference fit of the dowel pin has degraded, the dowel pin would be free to move along its longitudinal axis. However, the dowel pin is prevented from becoming a loose part due to the length (2") relative to the clearance gap (0.5") to the adjacent component (radial Key).

Locking Bars Break Due to their smaller size, it is determined that lock bars (or fragments thereof) may break down (wear) and with continued operation they may become loose parts. Lock bar wear related degradation has been observed at plants within the industry where cap screw failure has been observed. This wear scenario is unlikely to occur because the Ginna cap screw heads are separated and have insufficient loading to drive the wear condition. However, the locking bars can become loose parts by other means, including radial displacement into the bolt remnants should they be backed out. Based on these observations, additional consideration is conservatively given to the locking bar as a loose part.

Ref. 2 categorizes the effects of loose parts into two categories 1) impact and 2) wedging.

Impact -a moving loose object impacts the reactor vessel or a reactor internals component. This impact may result in unanalyzed stresses in the impacted component.

Impact loads of sufficient magnitude may result in deformation or fracture of the impacted component.

Wedging -a loose object becomes wedged in one of the existing clearances between the reactor vessel and the internals (i.e. the gap between the reactor vessel bottom head and the secondary core support structure). This wedging may result in unanalyzed loads due to closed gaps limiting thermal expansion.

~ew lock bar is an rectangular cuboid with dimensions 1-

---

with a mass of

. While it is more likely the lock bar will be broken off, the case of the worn through locking bar resulting in -50% length becoming a loose part is bounding. In either case, whole or 50% length, the impact load from the lock bar is

~le and bounded by previous analyses, i.e. a loose Baffle-to-Former bolt with a mass of (Ref. 2). Similarly, the wedging concern of the lock bar is evaluated by comparison to previous analyses. The size of the lock bar is insufficient to wedge and restrict movement of critical gaps in the Reactor Vessel Internals. It should also be noted that the lockin bar is of a size where it would likely be caught by the fuel lower debris filter (

and drop onto the core plate under low flow conditions.

Operability Evaluation Page 15 of 21 Based on this discussion, loose parts from the degraded Clevis Insert do not present a safety concern due to being captured or the applied impact and wedging loads on the reactor vessel and internals for normal plant operating conditions being bounded by previous acceptable evaluations.

Conclusions The above discussions provide assurance that none of the Safety Related functions of the LRSS will be adversely impacted for operations in modes 1-6. Specifically, the following functions are maintained:

Restriction of excessive lateral and rotational motion Limitation of displacement and misalignment to avoid component overstress and maintain critical component alignments Provide structural interface to transmit vibrational and dynamic loads to the Reactor Vessel, including o Impact and asymmetrical flow loading resulting from starting and running a single RCP o Seismic (OBE2 and SSE) loading o Faulted LOCA + SSE accident loading (b) any requirements or commitments established for the SSC and any challenges to these There are no commitments established for the LRSS and the associated components. The LRSS is required to perform the safety related functions as discussed in section (a) above and to provide non-safety related functions for installation alignment, allowances for thermal expansion, and alignment for the mating of the bottom mounted instrumentation.

As discussed in section (a) all of these functions are adequately maintained with the current condition of the 270° Clevis Insert. Operations in all modes is restricted to 1 operating cycle as documented in Ref. 2 to preclude the occurrence of unacceptable wear.

(c}

The circumstances of the degraded I nonconforming condition, including the possible failure mechanism(s)

The 270° Clevis Insert was found displaced during the conduct of visual examinations in accordance with ISi program. Further investigation identified that the Clevis Insert was loose and had lost its originally designed interference fit with the Clevis. Due to the displacement of the Clevis Insert, it was determined that all 8 of the installed cap screws had failed and that the alignment dowel had become dislodged. The retention of the Clevis Insert is determined to be the result of its weight bearing on the top of the Clevis and the Core Barrel Key limiting its radial movement during operation.

A Westinghouse report developed for the Pressurized Water Reactor Owner's Group (PWROG}, PWROG-15034-P Rev. 0, details the failure modes and impact of failure for the LRSS and the Clevis Insert. The report details that this as-found condition is not of immediate safety or operational concern. A review of this report provides the following conclusion for the failure mechanisms that led to the as-found condition.

2 OBE loading is bounded by the analyzed SSE loading for all wear cases. The SSE loads are shown to meet the allowable loading for the OBE conditions and explicit OBE loading evaluations are not required.

Operability Evaluation Page 16 of 21 The failure of the bolting is likely the result of Stress Corrosion Cracking (SCC) as a result of the susceptible Alloy X-750 material with a high initial bolt pre-load. sec of Alloy X-750 bolting has been identified in other PWRs LRSS and other components within the Reactor Vessel Internals. After the bolting failed, normal vibratory loads caused the Clevis Insert to wear on both sides of the Clevis and the dowel interface to wear. This wear loosened the fit allowing the Clevis Insert to displace radially inwards. Additional downward vertical loading and vibration resulted in the Clevis Insert wearing down into the top face of the Clevis.

While this as-found condition represents a degraded condition, no loss of Safety Function occurred and therefore the as-found condition is not a non-conforming condition.

(d) whether the potential failure is time dependent and whether the condition will continue to degrade and I or will the potential consequences increase.

The continued degradation and failure of the Clevis Insert is time-dependent and will continue to degrade.

Degradation rates during normal operation are expected to be small and significant degradation is expected to take more than one operating cycle. For one operating cycle, there is ample margin such that the added wear of the Inserts will not have an adverse impact on the LRSS ability to perform its Safety Function.

(e) the aggregate effect of the degraded or nonconforming condition in light of other open OpEvals There are no additional OpEvals related to the LRSS or the Reactor Vessel. Therefore, there is no impact on the conclusions made in this OpEval.

2.3 Is SSC operability supported? Explain basis (e.g., analysis, test, operating experience, [ X] [ ]

engineering judgment, etc.):-----------------------

The LRSS and degraded 210* Clevis Insert has been analyzed in Ref. 2 and this OpEval and shown capable of performing all of its Safety and Non-Safety related functions. All Reactor Internal component loads remain within acceptable limits and bounded by previous analyses and operability of the SSC is supported in all modes. The additional wear that may occur is sufficiently small for the given timeframe and will have acceptable impacts on the load carrying capacity of the LRSS. The maximum deflection of the LRSS in the unrestrained condition is reported in Ref.

2 as -

during normal operating conditions. This unrestrained deflection is similar to the as-designed tangential clearances and does not have an impact on the as-designed Reactor Vessel Internals deflections. However, Ref. 2 determined that should significant wear occur on all four Inserts where the Core Barrel is essentially unrestrained at the LRSS, all deflections will remain within allowable limits during design basis accident conditions. Further, deflections will be limited to ensure insertability of the RCCAs and preclude fuel grid crush.

If Step 2.3 =NO, notify Operations Shift Management immediately.

If Step 2.3 = YES, clearly document the basis for the determination.

Operability Evaluation Page 17 of 21 2.4 Are compensatory measures and I or corrective actions required?

If Step 2.4 =YES, complete section 3.0 (if NO, NIA section 3.0).

2.5 Reference Documents:

[ ]

[ x]

2.5.1 Technical Specifications Section(s): ------------------

TS 3.1.4 - Rod Group Alignment Limits TS 3.2.1 - Heat Flux Hot Channel Factor (FQ(Z))

TS 3.2.2 - Nuclear Enthalpy Rise Hot Channel Factor (FN(H))

TS 3.3.1 - Reactor Trip System (RTS) Instrumentation TS 3.5.2 - ECCS Modes 1,2,3 TS 3.5.3 - ECCS Mode 4 TS 3.4.4 - RCS Loops Mode 1 > 8.5% RTP TS 3.4.5 - RCS Loops Mode 1 < 8.5% RTP. 2. and 3 TS 3.4.6 - RCS Loop Mode 4 TS 3.4. 7 - RCS Loops Mode 5. Loops Filled TS 3.4.8 - RCS Loops Mode 5, Loops Not Filled 2.5.2 UFSAR Section(s):

Section 3.9.5.1.1.1, Section 3.9.5.1.1.5, and Section 4.2.2 provide details related to the Lower Radial Support System Section 3.9.2.3.4 provides discussion of the Leak Before Break LOCA size.

Table 3.9-29 discusses the allowable deflections of the Reactor Vessel Internals components.

Section 6.3.3.11 discusses the Mode 4 LOCA event

2.5.3 Other

1. Westinghouse Dwg 684J823, Lower Radial Support Clevis & Insert Gaging and Assembly.

2. L TR-RIDA-20-87, "R.E. Ginna Reactor Vessel Clevis Insert Degradation Operability Determination"

3. PWROG-15034-P Rev.a - Clevis Bolt Fabrication and Inspection Assessment

4. License Amendment No. 139

5. CN-RCDA-04-102 Rev.2, "Ginna Extended Power Uprate - LOCA Analysis" Operability Evaluation Page 18 of 21

6. 0236-0098-RPT-001 Revision Draft, "Structural Feasibility Assessment of Operation without One Core Barrel Lower Radial Support Clevis Insert"

7. Westinghouse Dwg 675C595, "Lower Radial Support Clevis Insert Hardware"

8. CN-RCDA-04-101, "RGE Extended Power Uprating - Seismic Analysis"

9. NSAL-11-2, "Impact of Change in Lower Radial Key Stiffness Value" 1 O. CN-RCDA-05-20, "R. E. Ginna Extended Power Uprating - Structural Qualification of Vessel Sections - Vessel Support Bracket"

11. CN-RCDA-05-35, "Ginna Station Extended Power Uprate Program Core Support Loads Evaluation"

12. WCAP-17367-P, "R.E. Ginna Replacement CW 316 SS Guide Tube Support Pin", FCMS ECP-11-000173-2003-1-01

13. CN-RCDA-04-175, "RGE Extended Power Uprating - Control Rod lnsertability"

14. CN-LIS-04-100, "RGE Vessel LOCA Hydraulic Forces Analysis for EPU and Fuel Upgrade

15. TB-19-5, "Westinghouse NSSS PWR Thermal Shield Degradation"

16. WCAP-14460, "Reactor Pressure Vessel and Internals System Evaluation for the R.E.

Ginna Plant-T-Average Reduction Program"

3.0 ACTION ITEM LIST:

Operability Evaluation Page 19 of 21 If, through evaluating SSC operability, it is determined that the degraded or nonconforming SSC does not prevent accomplishment of the specified safety function(s) in the TS and the intention is to continue operating the plant in that condition, then record below, as appropriate, any required compensatory measures to support operability and I or corrective actions required to restore full qualification. For corrective actions, document when the actions should be completed (e.g., immediate, within next 13 week period, next outage, etc.) and the basis for timeliness of the action. Corrective action time frames longer than the next refueling outage are to be explicitly justified as part of the OpEval or deficiency tracking documentation being used to perform the corrective action.

Compensatory Measure # 1: NIA Responsible Dept I Supv:

Action Due:

Action Tracking #:

Effects of the Compensatory Measure:

Compensatory Measure # 2: NIA Responsible Dept I Supv:

Action Due:

Action Tracking #:

Effects of the Compensatory Measure:

Compensatory Measure # 3: N/A Responsible Dept I Supv:

Action Due:

Action Tracking #:

Effects of the Compensatory Measure:

Operability Evaluation Page 20 of 21 Corrective Action or ACIT # 1: WEC complete analysis effort to extend the current operability evaluation for 1 O years of operation.

Responsible Dept I Supv: Pascuzzi Action Due: 0612021 Basis for timeliness of action: Timeline for completing a significant amount of design analysis work Action Tracking#: 04335182-25 Corrective Action or ACIT # 2: Develop inspection plan for future outages based on results of the WEC analysis effort and any loose parts evaluations needed by fuels.

Responsible Dept I Supv: Pascuzzi Action Due: 212021 Basis for timeliness of action: Timeline supports planning actions for 2021 Fall RFO Action Tracking#: 04335182-26 Corrective Action or ACIT # 3: NIA Responsible Dept I Supv: NIA Action Due: NIA Basis for timeliness of action: NIA Action Tracking#: NIA Operability Evaluation Page 21of21 4.2 Reviewer -::)._.., JS,,.._~

Date ~/..,.-/;!=*

4.3 Senior Manager Design Engr I Designee Concurrence Date s:/ sf ZIJ'ZZ>

4.4 Operations Shift Management Approval Date 5 /s / 2 tf 4.5 Ensure the completed form is forwarded to the OEPM for processing and Action Tracking entry as appropriate.

4.6 PORC Review Completed as required by Step 4.1.12 of this procedure.

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Digitally signed by Kendrick, Zoe Alexis

'Jl <nr&,.u.;C DN: cn=Kendrick, Zoe Alexis t/'

oa1002021.01.211u1,09-os*oo*

Date -----

(OEPM) 5.0 OPERABILITY EVALUATION CLOSURE:

5.1 Corrective actions are complete, as necessary, and the OpEval is ready for closure Date -----

(OEPM) 5.2 Operations Shift Management Approval ----------

Date -----

5.3 Ensure the completed form is forwarded to the OEPM for processing, Action Tracking entry, and cancellation of any open compensatory measures, as appropriate.

Adobe Acrobat Pro 2017 showing that this document has a valid signature, since redaction changes the document making it not possible to validate signatures on the redacted document. EJ Fischer, Ginna Regulatory Assurance

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4.0 SIGNATURES

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4.1 Preparer(s) M;cb<ee} 6i"i"6).~ Oate $'""/f7'z.o-z:-C>

4.3 Senior Manager Design Engr I Deslgnee ~~nence~

Date f'/~/2#'ZO 4.4 Ope1'8tionsShiftManagementApproval ~

Date.s/s/zf 4.5 Ensure the completed lorm Is forwarded to the OEPM for processing and Action Tracking entry as appropriate.

4.8 PORC Ravlaw C<<npletad as required by Slap 4.1.1.2 of this procedure.

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(OEPM) 5.0 OPERABILITY EVALUATION CLOSURE:

5.1 Corrective actions are complete, as necessary, and the OpEval Is ready for closure Data ___

(OEPM) 5.2 Operations Shift Management Approval--------

Date ___

5.3 Ernsura lhe oomplated fonn is forwarded to the OEPM for processing, Action Tracking entry, and cancellallon of any open compensatory measures, as appropriate.