ML21132A079

Additional Information Concerning the Proposed Alternative to Use the Successive Inspection Requirements of IWB-2420(c) of the 2017 Edition of ASME Section XI

ML21132A079
Person / Time
Site:	Ginna
Issue date:	05/12/2021
From:	David Gudger Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML21132A079 (6)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.55a May 12, 2021 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 R.E. Ginna Nuclear Power Plant Renewed Facility Operating License No. DPR-18 NRC Docket No. 50-244

Subject:

Additional Information Concerning the Proposed Alternative to Use the Successive Inspection Requirements of IWB-2420(c) of the 2017 Edition of ASME Section XI

Reference:

1) Letter from D. Gudger (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Proposed Alternative to Use the Successive Inspection Requirements of IWB-2420(c) of the 2017 Edition of ASME Section XI," dated November 11, 2020 In the Reference 1 letter, Exelon Generation Company, LLC (Exelon) requested a proposed alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (B&PV), Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," on the basis that the proposed alternative would provide an acceptable level of quality and safety. Specifically, Exelon requested the use of the successive inspection requirements of paragraph IWB-2420(c) of the ASME B&PV Code Section XI, 2017 Edition for the 270-degree lower radial support clevis insert in lieu of paragraph IWB-2420(b) of the 2004 Edition and 2013 Edition. This relief request applies to the fifth and sixth ten-year Inservice Inspection (ISI) intervals for the Ginna Nuclear Power Plant.

Attached is a summary of the analytical evaluation discussed in the relief request.

There are no regulatory commitments in this letter.

If you have any questions concerning this letter, please contact Tom Loomis at (610) 765-5510.

Respectfully, David T. Gudger Senior Manager - Licensing Exelon Generation Company, LLC

Additional Information Concerning the Proposed Alternative to Use the Successive Inspection Requirements of IWB-2420(c) of the 2017 Edition of ASME Section XI May 12, 2021 Page 2

Attachment:

Summary of Analytical Evaluation cc: Regional Administrator - NRC Region I NRC Senior Resident Inspector - R.E. Ginna Nuclear Power Plant NRC Project Manager - R.E. Ginna Nuclear Power Plant A. L. Peterson, NYSERDA

Attachment Summary of Analytical Evaluation

Summary of Analytical Evaluation Page 1 An analytical evaluation is being performed to justify continued operation with the observed and continuing Clevis degradation for the next 10 years when the Reactor Vessel 10-Year Inservice Inspection (ISI) is required. In order to qualify the Clevis Insert and the Lower Radial Support System (LRSS) for the next 10 years, the following items are being performed. The revised equipment loading will be compared to the existing design basis to determine if revision is required. A new design analysis will not be required if the new loads are bounded by the original analysis:

1. Wear Assessment (discussed below)

This analysis will determine the maximum wear that will occur over the next 10 years of operation.

2. Reactor Equipment System Model (RESM) (discussed below)

The current Ginna RESM exists in an obsolete program (WECAN) and is being migrated to ANSYS. The RESM, once migrated, will determine the loading outputs to utilize in the downstream analyses of the Reactor Vessel, Internals, and interconnected equipment.

3. Load Comparisons The revised loading for Flow Induced Vibration, Operating Basis Earthquake, Safe Shutdown Earthquake, and Loss of Coolant Accidents as generated by the RESM is compared to the original design basis analyses for the below components. Revisions to the design bases will be issued as required for the following components:

i. Reactor Internals ii. Nuclear Fuel iii. Reactor Pressure Vessel iv. Reactor Coolant Loop Piping

v. Reactor Pressure Vessel Supports vi. Reactor Pressure Vessel Closure Head Equipment The above evaluations are in-progress and will be used to update the existing Ginna design basis analyses to reflect the potential Clevis Insert wear conditions. Once the design change package is completed, the Ginna Reactor Vessel equipment will be qualified for 10 years (next required Reactor Vessel 10-year ISI examination) assuming the developed wear conditions.

Wear Analysis A wear analysis was developed for the Reactor Vessel LRSS Clevis Inserts. This wear analysis evaluated the LRSS Clevis Insert from two approaches: 1) best estimate based on the relative motion and contact conditions, and; 2) limiting wear based on the limiting assembly gaps and assuming this to be wear volume. For the purposes of the analysis, the wear condition developed using the best estimate approach was utilized. It should be noted that the self-limiting wear case was consistent with that utilized for the one-cycle justification.

The best estimate wear case utilized the Archard equation and vibrational motion of the Core Barrel to determine expected wear over the period of ten years. The applied loading between the radial key and the Clevis Insert decays with operating time. Additionally, due to the loss of material from the wear condition, additional load decay will occur. To maximize the amount of wear, the following assumptions were made:

Summary of Analytical Evaluation Page 2

1. Worst case initial interference was assumed between the Clevis Insert and the Radial Key.

2. The Clevis Insert is assumed to follow with the Core Barrel motion, no slip between the Clevis Insert and Radial Key.

3. One removal and reinstallation of the Core Barrel during the 10 Year cycle is assumed.

The core barrel movements utilized for the wear determination occur due to:

1. Core Barrel Vibrations

2. Flow Induced Vibrations of the Insert

3. Thermal Heat-up/Cool-down cycles The vertical wear (loss of Clevis Insert flange thickness) was assessed assuming sliding between the Clevis Insert flange and the support block. Flange material loss is compared to the flange thickness to determine if the flange will be worn through and lose vertical retention.

Results show that the material loss is less than the flange thickness and is acceptable for a 10-year operating period.

The wear developed by this analysis was utilized in the RESM analysis to modify the loading inputs to the Ginna design basis analyses which will be revised to demonstrate qualification with the degraded Clevis Insert.

Reactor Equipment System Model (RESM)

Modeling improvements to the RESM were made to address the Clevis Insert project:

1. The wear inputs from the wear analysis were utilized to vary the interface at the Lower Radial Support System

2. Five wear cases were utilized as follows:

a. Nominal (as-designed) gaps at all inserts

b. Maximum gap at 270 insert, all others at nominal gap

c. Maximum gap at 270 and 90 inserts, all others at nominal gap

d. Maximum gap at all inserts

e. 270 insert wedged due to loose parts, all others at nominal gap

3. Detailed modeling of the Simplified Head Assembly The maximum wear used in the RESM analysis is the wear determined by the best estimate approach, versus the more limiting, overly conservative self-limiting wear.

This revised RESM is equivalent to the original and other industry RESMs except that provisions have been added to vary the gaps at the LRSS Clevis Inserts. Loading is developed for Ginna and applied to the RESM as follows:

1. Vertical Steady State Loads

2. Seismic Accelerations (SSE and OBE) from the Ginna Containment Building at the Reactor Vessel Nozzle Supports

3. LOCA input loading

4. FIV forcing functions

Summary of Analytical Evaluation Page 3 A standard set of outputs (loads, accelerations, and displacements) are generated at various points for use in downstream analyses.

Remaining Work in Progress The following products are currently in progress. These products update the existing Ginna design basis and provide qualification of the Clevis Insert degradation for the next 10 years of operation. Evaluations are performed to show the existing design basis remains bounding.

Items 7, 8, and 9 are new or revised documents which will be used to update the design basis:

1. Reactor Vessel Internals Structural Evaluation

2. Nuclear Fuel Evaluation

3. RCS Piping Structural Evaluation

4. Reactor Vessel Structural Evaluation

5. RCS Piping Aging Management Evaluation

6. Reactor Vessel Support Evaluations

7. Reactor Vessel Simplified Head Assembly Analysis (Revised Calculation)

8. CRDM Analysis (New Calculation)

9. CETNA Design Specification/Report Update The results of the analyses and evaluations will be provided to Ginna in a final summary letter.

Conclusions This qualification will meet the requirements of ASME Section XI IWB-3142.4 to demonstrate the observed degradation and expected ongoing degradation does not impact the qualification of the Reactor Vessel equipment for the next 10-year ISI interval.