WO 20-0029, License Amendment Request to Revise Specification 5.5.16 for Permanent Extension of Type a and Type C Leak Rate Test Frequencies

License Amendment Request to Revise Specification 5.5.16 for Permanent Extension of Type a and Type C Leak Rate Test Frequencies

ML20111A327
Person / Time
Site:	Wolf Creek
Issue date:	04/20/2020
From:	Mccoy J Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WO 20-0029
Download: ML20111A327 (175)

Text

Wolf Creek~,~

Nuclear Operating Corporation Jaime H. McCoy Site Vice President April 20, 2020 WO 20-0029 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: License Amendment Request to Revise Technical Specification 5.5.16 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies To Whom It May Concern:

Pursuant to 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests an amendment to Renewed Facility Operating License Number NPF-42 for the Wolf Creek Generating Station (WCGS). The license amendment request (LAR) proposes to revise Technical Specification (TS) 5.5.16, "Containment Leakage Rate Testing Program," to reflect the following:

• Increase the Type A integrated leakage rate test program interval from 10 years to 15 years in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A and the conditions and limitations specified in NEI 94-01, Revision 2-A.

• Adopt an extension of the containment isolation valve leakage rate testing (Type C) frequency from 60 months to 75 months for selected components in accordance with NEI 94-01, Revision 3-A.

• Adopt the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2002, Containment System Leakage Testing Requirements.

• Adopt a more conservative allowable test interval extension of nine months for Type A, Type B, and Type C leakage rate tests in accordance with NEI 94-01, Revision 3-A.

It has been determined that this application does not involve a significant hazard consideration as determined per 10 CFR 50.92, "Issuance of amendment." Pursuant to 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," Section (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment. The amendment application was reviewed by the WCNOC Plant Safety Review Committee.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation," a copy of this amendment application is being provided to the designated Kansas State Official.

P.O. Box 411 I Burlington, KS 66839 I 620-364-8831

WO 20-0029 Page 2 of 3 Attachment I provides the Evaluation of Proposed Changes. Attachment II provides the Proposed TS Changes (Mark-up) Pages. Attachment Ill provides the Retyped TS Pages. The Enclosure provides the risk assessment supporting the proposed amendment.

As the containment leakage rate testing program exists currently, an integrated leak rate test would be required during Refueling Outage 24 (RF24) which is scheduled to begin March 25, 2021. WCNOC requests approval of this proposed license amendment by January 28, 2021, to allow proper planning for deferral of this test in accordance with this request. WCNOC also requests that the NRG provide notification by November 12, 2020, as to whether the requested approval date can be met. It is anticipated that the license amendment, as approved, will be effective upon issuance, to be implemented within 90 days from the date of issuance.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4156, or Ron Benham at (620) 364-4204.

Sincerely, Jaime H. McCoy JHM/rlt Attachments: I - Evaluation of Proposed Changes II - Proposed Technical Specification Changes (Mark-Up) Pages Ill - Retyped Technical Specification Pages

Enclosure:

Evaluation of Risk Significance of Permanent ILRT Extension cc: S. A. Morris (NRG), w/a, w/e N. O'Keefe (NRG), w/a, w/e K. S. Steves (KDHE), w/a, w/e B. K. Singal (NRG), w/a, w/e Senior Resident Inspector (NRG), w/a, w/e

WO 20-0029 Page 3 of 3 STATE OF KANSAS )

) ss COUNTY OF COFFEY )

Jaime H. McCoy, of lawful age, being first duly sworn upon oath says that he is Site Vice President of Wolf Creek Nuclear Operating Corporation; that he has read the foregoing document and knows the contents thereof; that he has executed the same for and on behalf of said Corporation with full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and belief.

By_ q~ j-/ o/f/XI_

Jaime H. McCoy Site Vice President SUBSCRIBED and sworn to before me this d r')D+~day of ApP -1 I , 2020.

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Notary Public cf ;km~

Expiration Date uVrvnlla/Ll( *//, diJJ;;_

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Attachment I to WO 20-0029 Page 1 of 122 EVALUATION OF THE PROPOSED CHANGE

SUBJECT:

License Amendment Request to Revise Wolf Creek Generating Station, Unit 1 Technical Specification 5.5.16, "Containment Leakage Rate Testing Program," for Permanent Extension of Type A and Type C Leak Rate Test Frequencies 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusion

6.0 REFERENCES

Attachment I to WO 20-0029 Page 2 of 122 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for Amendment of License, Construction Permit, or Early Site Permit," Wolf Creek Nuclear Operating Corporation (WCNOC) requests an amendment to Renewed Facility Operating License No. NPF-42, for Wolf Creek Generating Station (WCGS), Unit 1.

The proposed change revises Technical Specification (TS) 5.5.16, "Containment Leakage Rate Testing Program," to reflect the following:

• Increases the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with Nuclear Energy Institute (NEI) Topical Report (TR)

NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A (Reference 2), and the conditions and limitations specified in NEI 94-01, Revision 2-A (Reference 8).

• Adopts an extension of the containment isolation valve (CIV) leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.

• Adopts the use of American National Standards Institute/American Nuclear Society (ANSI/ANS) 56.8-2002, Containment System Leakage Testing Requirements (Reference 29).

• Adopts a more conservative allowable test interval extension of nine months, for Type A, Type B and Type C leakage rate tests in accordance with NEI 94-01, Revision 3-A.

Specifically, the proposed change contained herein, would revise WCGS TS 5.5.16, paragraph a.,

by replacing the references to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," (Reference 1) and NEI 94-01, Revision 0, (Reference 5) with a reference to NEI 94-01, Revision 3-A, and the limitation and conditions specified in NEI 94-01, Revision 2-A, dated October 2008, as the documents used by WCGS to implement the performance-based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J.

Attachment I to WO 20-0029 Page 3 of 122 2.0 DETAILED DESCRIPTION WCGS TS 5.5.16, "Containment Leakage Rate Testing Program," paragraph a., currently states, in part:

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Testing Program,"

dated September 1995, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

The proposed changes to WCGS TS 5.5.16 will replace the reference to RG 1.163 with reference to TR NEI 94-01, Revisions 2-A and 3-A. There are no changes being made to the exceptions identified in paragraph a., items 1 and 2, as a result of this proposed change.

The proposed change revises the WCGS TS 5.5.16, paragraph a., to read as follows (with recommended changes using strike-out and bold-type for clarification purposes):

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program, dated September 1995, Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

Attachment I to WO 20-0029 Page 4 of 122

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

Therefore, the retyped ("clean") version of TS 5.5.16, paragraph a., would appear as follows:

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

The marked-up TS page for WCGS TS 5.5.16 are provided in Attachment II.

The typed ("clean") TS page for WCGS TS 5.5.16 are provided in Attachment III.

The Enclosure to this submittal contains the plant specific risk assessment conducted to support this proposed change. This risk assessment follows the guidelines of NRC RG 1.174, Revision 3 (Reference 3) and RG 1.200, Revision 2 (Reference 4). The risk assessment concludes that increasing the ILRT test frequency on a permanent basis to a one-in-fifteen-year frequency is not considered to be significant because it is considered to be a small change to the WCGS Unit 1 risk profile.

3.0 TECHNICAL EVALUATION

3.1 Description of Primary Containment System The containment structure for WCGS is a prestressed, post-tensioned concrete structure with a cylindrical wall, a hemispherical dome, and a flat foundation slab. The wall and dome form a prestressed, post-tensioned system consisting of horizontal tendons in the wall and inverted U-

Attachment I to WO 20-0029 Page 5 of 122 shaped vertical tendons in the wall and dome. The foundation slab is reinforced with carbon steel.

The inside surface of the structure is lined with a carbon steel liner to ensure a high degree of leak tightness. The containment structure completely encloses the reactor and reactor coolant system, i.e., the reactor pressure vessel, the steam generators, the reactor coolant loops and portions of the associated auxiliary systems, the pressurizer, accumulator tanks, and associated piping. The design ensures that the containment structure is protected against postulated missiles from both equipment failures and external sources. The containment design provides means for the integrated leak rate testing of the containment structure and for local leak rate testing of individual piping, electrical, and access penetrations of the containment.

The containment structure design is the same standard state-of-the-art design that has been applied to several other Bechtel-designed plants. The basic structure is similar to the containment structures at Farley, Palo Verde, and Turkey Point and identical to the structure at Callaway.

Table 3.1-1 Principal Nominal Dimensions of the WCGS Reactor Building Interior diameter 140 ft Interior height 205 ft Height to spring line 135 ft Base slab thickness 10 ft Cylinder wall thickness 4 ft Dome thickness 3 ft Liner plate thickness 0.25 in 3.1.1 Containment Foundation The reactor building foundation is a 10-foot-thick reinforced concrete mat, 154 feet in diameter, founded 11 feet below plant grade. The central reactor cavity and instrumentation tunnel extend below the reactor building foundation, with the bottom of the 5.5-foot-thick foundation slab located 36 feet below grade. The 8-foot-wide tendon access gallery, located beneath the perimeter of the

Attachment I to WO 20-0029 Page 6 of 122 reactor building mat, has a 4.25-foot-thick foundation slab, the bottom of which is 25.25 feet below grade.

Located below the base slab is a continuous peripheral tendon gallery that provides access for the installation and inspection of the vertical posttensioning system.

3.1.2 Liner Plate A carbon steel liner plate covers the entire inside surface of the reactor building (excluding penetrations). The liner is 1/4-inch thick but is thickened locally around the penetrations, large brackets, and major attachments. The liner plate, including the thickened plate, is anchored to the concrete structure. The vertical and dome liner plates are also used as forms for concrete placement.

The 1/4-inch-thick liner plate material conforms to the requirements of the Specification for Low and Intermediate Tensile Strength Carbon Steel Plates for Pressure Vessels (ASME SA 285),

Grade A. Thickened liner plates, ranging from 1/2 inch to 2 inches in thickness, were used at penetrations, brackets, and embedded assemblies and conform to the requirements of the Specification for Carbon Steel Plates for Pressure Vessels for Moderate and Lower Temperature Service (ASME SA516), Grade 70. In the event that significant loads are to be transmitted through the thickness dimension of the liner, nondestructive tests were performed to determine the capability of the liner materials used in these locations. Leak chase channels and angles are also attached at seam welds where the welds are inaccessible to nondestructive examination after construction.

3.1.3 Containment Shell (Cylinder and Dome) Post-Tensioning System A tendon system is employed to post-tension the cylindrical shell and dome of the reactor building.

The system uses unbonded tendons, each consisting of approximately 170 one-quarter-inch-diameter high strength steel wires and anchorage components consisting of stressing washers.

The prestressing load is transferred by cold-formed button heads on the ends of the individual wires, through stressing washers, to the steel bearing plates embedded in the structure. The ultimate strength of each tendon is approximately 1,000 tons.

The vertical tendons consist of 86 inverted U-shaped tendons, which extend through the full height of the cylindrical wall over the dome and are anchored at the bottom of the base slab. The cylinder circumferential (hoop) tendons consist of 135 tendons anchored at three buttresses equally spaced around the outside of the reactor building. Each tendon is anchored at buttresses located 240 degrees apart. Three adjacent tendons, anchored at alternate buttresses, result in two complete hoop tendons

Attachment I to WO 20-0029 Page 7 of 122 Prestressing of the hemispherical dome is achieved by a two-way pattern of the inverted U-shaped tendons and 30 hoop tendons, which start at the springline (i.e. transition point from cylinder to dome) and continue up to an approximate 45-degree vertical angle from the springline.

The unbonded tendons are installed in tendon ducts (sheathing) and tensioned in a predetermined sequence. The ducts consist of galvanized, spiral-wrapped, semirigid corrugated steel tubing.

After stressing, a petroleum-based corrosion inhibitor is pumped into the duct.

3.1.4 Post Tensioning System Long Term Surveillance The long-term surveillance program consists of evaluating the general conditions of the post-tensioning system. Data on wire corrosion levels and tendon lift-off forces are obtained and analyzed. The surveillance tendons are designated as part of the inservice inspection program, which conforms with the American Society of Mechanical Engineers (ASME) ASME Boiler and Pressure Vessel Code (BPVC),Section XI, Division 1, Subsections IWE, Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants," and IWL, Requirements for Class CC Concrete Components of Light-Water Cooled Plants, and applicable addenda as required by 10 CFR 50.55a, Codes and standards, except where an exemption, relief, or an alternative has been authorized by the NRC.

This surveillance program provides assurances of the continuing ability of the structure to meet the design functions.

The ASME Section XI, Subsection IWE containment inservice inspection program, provides aging management inspection of the steel liner of the concrete containment building, including the containment liner plate, piping and electrical penetrations, access hatches, and the fuel transfer tube. The ASME Section XI, Subsection IWL containment inservice inspection program, manages aging inspection of the concrete containment structure (including the tendon gallery ceiling), the concrete dome, and the post-tensioning system. Inspections in these Programs are credited as license renewal aging management activities.

In conformance with 10 CFR 50.55a(g)(4)(ii), both of these WCGS Programs are updated during each successive 120-month inspection interval to comply with the requirements of the latest edition and addenda of the Code specified twelve months before the start of the inspection interval.

3.1.5 Containment System Steel Penetration Items The major containment penetrations and portions of penetrations intended to resist pressure which by design, are not backed by structural concrete, and are part of the containment pressure boundary, include access openings, such as the equipment hatch and personnel hatches, piping penetration sleeves, fuel transfer tube penetration sleeves, electrical penetration sleeves, and the purge line penetration sleeves.

Attachment I to WO 20-0029 Page 8 of 122 3.1.5.1 Equipment and Personnel Access Hatches and Penetration Sleeves The equipment hatch is a welded steel assembly with a double-gasketed, flanged, and bolted cover. Provision is made for leak testing of the flange-gasket combination by pressurizing the space between the gaskets. One personnel hatch and one auxiliary hatch, both of which are welded steel assemblies. Each hatch has two doors with double gaskets in series.

In order to assure leak tightness, the space between the gaskets are normally pressurized. The doors are mechanically interlocked to ensure that one door cannot be opened unless the second door is sealed. The interlock can be deliberately overridden by the use of special tools and procedures. Each door is equipped with quick-acting valves for equalizing the pressure across the doors. The doors will not operate unless the pressure is equalized. Pressure equalization is possible from every point at which the associated door can be operated. The valves for the two doors are properly interlocked so that only one valve can be opened at a time and only when the opposite door is closed and sealed.

Each door is designed so that, with the other door open, it will withstand and seal against design and testing pressure of the containment vessel. There is visual indication outside each door showing whether the opposite door is open or closed. Provision is made outside each door for remotely closing and latching the opposite door so that in the event that one door is accidentally left open it can be closed remotely.

The access hatch barrels have nozzles which permit pressure testing of the hatch at any time. The hatches are protected from tornado missiles by enclosure structures or shields.

The personnel hatch is enclosed within the auxiliary building. The auxiliary hatch is enclosed within an exterior tornado - resistant concrete structure. The personnel and auxiliary access hatch barrels are designated as ASME Section III, Class MC components.

The hatch penetration sleeves project into the reactor building and are used to support the hatches.

These items are made from carbon steels and conform to the requirements of ASME Section III, Subsection NE.

3.1.5.2 Pipe Penetration Sleeves Piping penetrations are divided into three general groups: Type 1, Type 2, and Type 3. They are described, as follows:

Type 1 - Flued head penetrations used for most high energy piping. Examples of Type 1 penetrations are the main steam and main feedwater lines. Type 1 piping penetrations consist of the following major steel items:

a. Process Pipe - made of welded or seamless carbon or stainless steel and is welded to the flued head and conforms to the requirements of ASME Section III, Subsection NC.

Attachment I to WO 20-0029 Page 9 of 122

b. Flued Head - made from forged carbon or stainless steel and conforms to the requirements of ASME Section III, Subsection NC. It is designed to contain the full pressure of the process fluid and full reactor building pressure in parts adjoining the pipe sleeve. The connecting process pipes and the flued heads are designed and analyzed to be capable of carrying loads resulting from the failure of the process pipe.

c. Pipe Sleeve - consists of the portion which projects into the reactor building and supports the flued head. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

Type 2 - Closure plate penetrations used for some high energy, all moderate energy, and all low energy general piping. The use of this type of penetration for high energy piping is limited to only those cases where an analysis based on combination of pressure, temperature, and line size has demonstrated the adequacy of the design. Type 2 piping penetrations consist of the following major steel items:

a. Process Pipe - made of welded or seamless carbon or stainless steel and is welded to the closure plate, conforms to the applicable requirements of ASME Section III, Subsection NC.

b. Closure Plate - made from carbon or stainless steel plate and conforms to the requirements of ASME Section III, Subsection NC.

c. Pipe Sleeve - consists of the portion which projects into the reactor building and supports the closure plate. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

Type 3 - Spare penetrations reserved for future use or small access penetrations. Type 3 spare penetrations consist of the following major items:

a. Solid Closure Plate (mechanical or welded) or Pipe Cap - made from carbon steel and conforms to the requirements of ASME Section III, Subsection NE or NC.

b. Pipe Sleeve - consists of the portion which projects into the reactor building. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

3.1.5.3 Fuel Transfer Tube Penetration Sleeve The fuel transfer tube penetration is provided to transfer fuel between the refueling canal and the fuel storage pool during refueling operations of the reactor. The penetration consists of a 20-inch-diameter stainless steel pipe installed inside a 26-inch steel sleeve. The steel sleeve which projects into the reactor building conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed. The inner pipe acts as the fuel transfer tube. The

Attachment I to WO 20-0029 Page 10 of 122 steel sleeve is designed to provide integrity of the reactor building, allow for differential movement between structures, and prevent leakage through the fuel transfer tube in the event of an accident.

3.1.5.4 Electrical Penetration Sleeves Steel sleeves, which form a portion of the containment pressure boundary, are provided for electrical/fiber optic penetrations. The sleeve consists of the portion which projects out of the reactor building and supports the electrical/fiber optic assembly. It conforms to ASME Section III, Subsection NE, except that authorized inspection and stamping are not performed.

3.1.5.5 Purge Line Penetration Sleeves The steel sleeves, which are embedded in the reactor building wall concrete, are welded to the purge line piping and form a part of the ASME Section III, Class 2 purge line piping system. The sleeves conform to ASME Section III, Subsection NC.

3.2 Emergency Core Cooling System (ECCSA Net Positive Suction Head (NPSH)

Analysis The injection mode of ECCS operation consists of the ECCS pumps (centrifugal charging pumps, safety injection (SI) pumps, and residual heat removal (RHR) pumps) and the containment spray (CS) pumps take suction from the refueling water storage tank (RWST) delivering water to the reactor coolant system (RCS) and containment spray system, respectively. Upon depletion of the RWST water volume, manual valve repositioning is performed to align the suction source of key ECCS pumps to the reactor building sump where cooling flow via recirculation can be maintained.

Based on the containment pressure-temperature analyses which assume runout flows of all pumps, including the containment spray pumps, which draw from the RWST, switchover of the RHR pumps occurs approximately 13.7 minutes after the accident.

A minimum water volume of 394,000 gallons is maintained in the RWST to ensure that, after a LOCA, sufficient water is injected for emergency core cooling and for rapidly reducing the containment pressure and temperature. In addition, this volume ensures that; sufficient water is available in the containment sump to permit recirculation flow to the core and the containment; ensures that sufficient water is available to meet the NPSH requirements of the RHR and containment spray pumps; and, assures that a sufficient water volume is available in the RWST to allow for manual switchover of the containment spray pumps.

The safety intent of RG 1.1, Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps (Safety Guide 1), is met by the design of the ECCS so that adequate NPSH is provided to system pumps. In addition to considering the static head and suction line pressure drop, the calculation of available NPSH in the recirculation mode assumes that the vapor pressure of the liquid in the sump is equal to the containment ambient pressure. This ensures that the actual available NPSH is always greater than the calculated NPSH.

Attachment I to WO 20-0029 Page 11 of 122 Additionally, the containment recirculation sump design meets the intent of RG 1.82, "Sump for Emergency Core Cooling and Containment Spray Systems."

3.2.1 Residual Heat Removal System (RHRS)

In its capacity as the low head portion of the ECCS, the RHRS provides long-term recirculation capability for core cooling following the injection phase of a LOCA. This function is accomplished by aligning the RHRS to take fluid from the containment sump, cool it by circulation through the residual heat exchangers, and supply it to the core directly as well as via the centrifugal charging pumps and SI pumps.

The design parameters for the NPSH of the RHR pumps as shown below in Table 3.2.1-1, are the specific operational parameters for the RHR pumps.

Table 3.2.1-1 Input and Results of NPSH Analysis for Residual Heat Removal Pumps NPSH Reference Elevation (2) 1972.07 ft Static head available (LOCA) (1) 30.015 ft Suction line losses @ 4,760 gpm 3.945 ft Available NPSH @ 4,760 gpm (3) 23.79 ft Required NPSH @ 4,760 gpm 21.01 ft Notes:

(1) Large LOCA conditions are provided for the RHR pumps since the flow rates, line losses, and NPSH required are greater than those associated with an MSLB wherein the RCS pressure remains above the RHR shutoff head at switchover to recirculation.

(2) NPSH reference elevation is 3-3/8 inches above the discharge centerline.

(3) Includes 1.724 ft. total head loss across the sump strainer with both the Spray Pump and RHR Pump running in Recirculation, and a 0.56 ft. allowance for emergency diesel generator (EDG) frequency uncertainties.

3.2.2 Safety Injection (SI) and Centrifugal Charging In the event of an accident, the SI pumps are started automatically upon receipt of a Safety Injection Signal (SIS), and take suction from the RWST via normally open, motor-operated valves, and

Attachment I to WO 20-0029 Page 12 of 122 deliver water to the RCS during the injection phase. Additionally, the SIS automatically aligns the centrifugal charging pumps to take suction from the RWST and deliver flow through the Boron Injection Tank (BIT) to the RCS at the prevailing RCS pressure.

Available and required NPSH analyses for SI pumps and centrifugal charging pumps are provided.

This analysis indicates that sufficient NPSH margin exists for each pump. Specifically, for the SI pumps the required NPSH is 25 ft with the available NPSH at 44 ft and for the centrifugal charging pumps the required NPSH is 28 ft and the available NPSH is 44 ft.

When a predetermined low RWST level is reached, the SI and centrifugal charging pumps are manually re-aligned to the recirculation phase of ECCS and receive flow from the containment sump via the RHR pumps. Suction flow and sufficient NPSH to both these pumps are satisfied by the discharge of the RHR pumps. Additionally, to ensure that the required NPSH is available during recirculation, restriction orifices are provided in the four discharge lines into the RCS cold legs and in the two discharge lines into the RCS hot legs.

3.2.3 Containment Spray Although not an ECCS component, the CS piping size and layout provide adequate NPSH to the CS pump during all anticipated operating conditions, in accordance with Regulatory Guide 1.1. In calculating the available NPSH, the conservative assumption has been made that the water in the containment sump after a design basis LOCA is a saturated liquid, and no credit has been taken for anticipated subcooling. Additionally, in calculating the water level within the reactor building which contributes to the NPSH available to the CS pumps at the beginning of its recirculation phase, consideration has been given to the potential mechanisms of water loss within the reactor building.

These water loss mechanisms include water present in the vapor phase, water loss to compartments below elevation (EL) 2,000, water loss above EL 2,000, and water loss due to wetted surfaces. The reduction in water level due to potential water loss mechanisms is considered in the calculated NPSH available.

The design parameters for the NPSH of the CS pumps as shown below in Table 3.2.3-1, are the specific operational parameters for the CS pumps.

Attachment I to WO 20-0029 Page 13 of 122 Table 3.2.3-1 Input and Results of NPSH Analysis for Containment Spray Pumps Pump elevation (discharge centerline) 1971 ft 3/4 in Static head available (MSLB) 31 ft 3/16 in Suction line losses @ 3,950 gpm 9.56 ft Available NPSH @ 3,950 gpm (1) 20.1 ft Required NPSH @ 3,950 gpm 16.5 ft (from USAR Figure 6.2.2-5)

Note:

(1) Includes 1.724 ft. total head loss across the sump strainer with both the Spray Pump and RHR Pump running in Recirculation, and a 0.56 ft. allowance for EDG frequency uncertainties.

Calculated NPSH exceeds required NPSH by at least 10 percent. The recirculation piping penetrating the containment sumps is nearly horizontal to minimize vortexing. In addition, a vortex breaker is provided in the inlet of the piping from the sump.

3.2.4 Conclusion NPSH margin is adequately provided for proper RHR, Safety Injection, Centrifugal Charging and Containment Spray pump operation. In addition to considering the static head and suction line pressure drop, calculation of available NPSH for the RHR and CS Systems in the recirculation mode assumes that the vapor pressure of the liquid in the sump is equal to the containment ambient pressure. This ensures that the actual available NPSH is always greater than the calculated NPSH.

WCGS does not rely upon containment overpressure for ECCS performance.

3.3 Justification for the TS Change 3.3.1 Chronology of Testing Requirements of 10 CFR 50, Appendix J The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the TS. 10 CFR 50, Appendix J also ensures that periodic surveillances of reactor containment penetrations and isolation valves are performed so that proper maintenance and repairs are made during the service life of the containment and of the systems and components penetrating primary containment. The limitation on containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant design basis accident (DBA). Appendix J identifies three types of required tests:

Attachment I to WO 20-0029 Page 14 of 122

1) Type A tests, intended to measure the primary containment overall integrated leakage rate;

2) Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage-limiting boundaries (other than valves) for primary containment penetrations, and;

3) Type C tests, intended to measure containment isolation valve (CIV) leakage rates.

Types B and C tests identify the vast majority of potential containment leakage paths. Type A tests identify the overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Types B and C testing.

In 1995, 10 CFR 50, Appendix J, was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50, Appendix J refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.

Also, in 1995, RG 1.163 (Reference 1) was issued. The RG endorsed NEI 94-01, Revision 0, (Reference 5) with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) test from three tests in 10 years to one test in 10 years. This relaxation was based on an NRC risk assessment contained in NUREG-1493, (Reference 6) and Electric Power Research Institute (EPRI) TR-104285 (Reference 7), both of which showed that the risk increase associated with extending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval, provisions for extending the test interval an additional 15 months were considered in the establishment of the intervals allowed by RG 1.163 and NEI 94-01, but that this extension of interval "should be used only in cases where refueling schedules have been changed to accommodate other factors."

In 2008, NEI 94-01, Revision 2-A (Reference 8), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC safety evaluation (SE) report (SER) on NEI 94-01. NEI 94-01, Revision 2-A, includes provisions for extending Type A ILRT intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163. The document also delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies.

Justification for extending test intervals is based on the performance history and risk insights.

Attachment I to WO 20-0029 Page 15 of 122 In 2012, NEI 94-01, Revision 3-A (Reference 2), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J and includes provisions for extending Type A ILRT intervals to up to 15 years. NEI 94-01 has been endorsed by RG 1.163 and NRC SERs of June 25, 2008 (Reference 9), and June 8, 2012 (Reference 10), as an acceptable methodology for complying with the provisions of Option B in 10 CFR 50, Appendix J. The regulatory positions stated in RG 1.163, as modified by references 9 and 10, are incorporated in NEI 94-01, Revisions 3-A. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights.

Extensions of Type B and Type C test intervals are allowed based upon completion of two consecutive periodic As-Found tests where the results of each test are within a licensees allowable administrative limits. Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment airlocks) and up to a maximum of 75 months for Type C tests. If a licensee considers extended test intervals of greater than 60 months for Type B or Type C tested components, the review should include the additional considerations of As-Found tests, schedule and review as described in NEI 94-01, Revision 3-A, Section 11.3.2.

The NRC has provided the following guidance concerning the use of test interval extensions in the deferral of ILRTs beyond the 15-year interval in NEI 94-01, Revision 2-A, NRC SER Section 3.1.1.2:

As noted above, Section 9.2.3, NEI TR 94-01, Revision 2, states, "Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 15 years based on acceptable performance history." However, Section 9.1 states that the "required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes." The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

NEI 94-01, Revision 3-A, Section 10.1, Introduction, concerning the use of test interval extensions in the deferral of Type B and Type C LLRTs, based on performance, states, in part:

"Consistent with standard scheduling practices for Technical Specifications Required Surveillances, intervals of up to 120 months for the recommended surveillance frequency for Type B testing and up to 75 months for Type C testing given in this section may be extended by up to 25 percent of the test interval, not to exceed nine months.

Attachment I to WO 20-0029 Page 16 of 122 Notes: For routine scheduling of tests at intervals over 60 months, refer to the additional requirements of Section 11.3.2.

Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. This provision (nine-month extension) does not apply to valves that are restricted and/or limited to 30-month intervals in Section 10.2 (such as BWR MSIVs) or to valves held to the base interval (30 months) due to unsatisfactory LLRT performance."

The NRC also provided the following concerning the extension of ILRT intervals to 15 years in NEI 94-01, Revision 3-A, NRC SER Section 4.0, Item 2, which states, in part:

"The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time" 3.3.2 Current WCGS Primary Containment Leakage Rate Testing Program Requirements 10 CFR Part 50, Appendix J was revised, effective October 26, 1995, to allow licensees to choose containment leakage testing under either Option A, Prescriptive Requirements, or Option B, Performance-Based Requirements.

Currently, WCGS TS 5.5.16, "Containment Leakage Rate Testing Program," requires that a program be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The program is required to be in accordance with the guidelines contained in RG 1.163. RG 1.163 endorses, with certain exceptions, NEI 94-01, Revision 0, as an acceptable method for complying with the provisions of Appendix J, Option B.

RG 1.163, Section C.1 states that licensees intending to comply with 10 CFR 50, Appendix J, Option B, should establish test intervals based upon the criteria in Section 11.0 of NEI 94-0 rather than using test intervals specified in ANSI/ANS 56.8-1994. NEI 94-01, Section 11.0 refers to Section 9, which states that Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once-per-ten years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage was less than 1.0La (where La is the maximum allowable leakage rate at design pressure). Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine performance shall be at least 24 months.

Adoption of the Option B performance-based containment leakage rate testing program altered the frequency of measuring primary containment leakage in Types A, B, and C tests but did not alter

Attachment I to WO 20-0029 Page 17 of 122 the basic method by which Appendix J leakage testing is performed. The test frequency is based on an evaluation of the "as found" leakage history to determine a frequency for leakage testing which provides assurance that leakage limits will not be exceeded. The allowed frequency for Type A testing as documented in NEI 94-01 is based, in part, upon a generic evaluation documented in NUREG-1493 (Reference 6). The evaluation documented in NUREG-1493 included a study of the dependence or reactor accident risks on containment leak tightness for differing containment types.

NUREG-1493 concluded in Section 10.1.2 that reducing the frequency of Type A tests (ILRT) from the original three (3) tests per 10 years to one (1) test per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Types B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements. Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, NUREG-1493 concluded that increasing the interval between ILRTs is possible with minimal impact on public risk.

3.3.3 WCGS 10 CFR 50, Appendix J, Licensing History November 10, 1993 - License Amendment No. 69 - Power Level Uprate License Amendment No. 69 increases the maximum allowable core power from 3411 megawatts thermal (MWt) to 3565 MWt (Reference 12). Peak Containment pressure Pa was unchanged due to the power uprate and remained at 48.0 psig. The Safety Analysis for this change included the statement that the staff reviewed the licensees evaluations and determined that the licensee adequately demonstrated that the containment will satisfy its design functions under the uprated conditions.

August 10, 1994 - License Amendment No. 76 - ILRT Test Interval Extension License Amendment No. 76 and an exemption from the requirements of 10 CFR 50, Appendix J, Section III.D.1(a), were granted for WCGS on August 12, 1994, and August 10, 1994, respectively (Reference 13). License Amendment No. 76 and the exemption from Appendix J provided one-time relief from the requirement to perform the overall integrated containment leakage rate test at intervals of 40 months plus or minus 10 months. The approval of the one-time exemption would extend the interval between the second and third ILRTs and would result in the third test being performed approximately six months after the end of the first 10-year service period. The third test would then coincide with the 10-year plant inservice inspections scheduled for the eighth refueling outage. (March of 1996).

March 1, 1996 - License Amendment No. 97 - Option B Testing License Amendment No. 97 approved changes to the TS (Appendix A to Facility Operating License) for WCGS (Reference 14). The changes allowed revision of the TS to reflect their approval to use 10 CFR Part 50, Appendix J, Option B for the WCGS containment leakage rate test program.

Attachment I to WO 20-0029 Page 18 of 122 Option B to Appendix J, requires a Containment Leakage Rates Testing Program to govern Type A, Type B and Type C, containment leakage rate surveillance testing. The WCGS Containment Leakage Rate Testing Program, was written to govern the Program. This procedure followed the recommendation of Nuclear Energy Institute 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 0, with exceptions.

March 17, 2004 - License Amendment No. 152 - Tendon and IWE/IWL Program License Amendment No. 152 (Reference 15) received approval from the NRC to revise TS Section 5.5.6, "Containment Tendon Surveillance Program," for consistency with the requirements of 10 CFR 50.55a(g)(4) of Title 10 of the Code of Federal Regulations (10 CFR) for components classified as Code Class CC. The approval revised TS 5.5.6 is to indicate that the Containment Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with the ASME BPVC Section XI, Division Subsection IWL and the applicable addenda as required by 10 CFR 50.55a, except where an exemption or relief has been authorized by the NRC. Approval was also received to delete the provisions of Surveillance Requirement (SR) 3.0.2 from this TS.

In addition, this license amendment approval revised TS 5.5.16, "Containment Leakage Rate Testing Program," to add exceptions to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Testing Program." Specifically, the amendment changed TS 5.5.16 at the end of the paragraph where the following verbiage was added to modify compliance with RG 1.163:

"as modified by the following exceptions to modify compliance with RG 1.163: "as modified by the following exceptions to: (1) the visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC. (2) The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC." Lastly, the Amendment approved the corresponding changes to the TS Bases for SR 3.6.1.1.

November 20, 2008 - Renewed Facility Operating License The NRC approved and issued the renewed facility operating license No. NPF-42 for the Wolf Creek Generating Station (WCGS) (Reference 16). NRC issued the renewed facility operating license based on the staff's review of the WCGS application dated September 27, 2006, as supplemented by letters submitted to the NRC through August 1, 2008. Their review did not result in an amendment of the TS for WCGS. The renewed facility operating license No. NPF-42 expires at midnight March 11, 2045.

The technical basis for issuing the renewed license was set forth in NUREG-1915, Safety Evaluation Report Related to the License Renewal of the Wolf Creek Generating Station,

Attachment I to WO 20-0029 Page 19 of 122 published in October 2008. The results of the environmental reviews related to the issuance of the renewed license are found in NUREG-1437, Supplement 32, Generic Environmental Impact Statement for License Renewal of Nuclear Plants Regarding Wolf Creek Generating Station, published on May 8, 2008.

(Section 3.6, License Renewal Aging Management, of this LAR provides a discussion of the approval and a description of the License Renewal activities, commitments, and tracking for containment items at WCGS.)

January 6, 2020 - License Amendment No. 223 The NRC approved and issued Amendment No. 223 (Reference 39) which revised TS 3.6.3, "Containment Isolation Valves," to remove the use of a blind flange to meet Limiting Condition for Operation (LCO) 3.6.3, Condition D, Required Action D.1. Also, it revises Surveillance Requirement 3.6.3.1 to remove the use of blind flange while performing the surveillance consistent with the requirements of the LCO.

3.3.4 Continued Acceptability of TS Amendment 152 for WCGS By application dated October 17, 2003, WCNOC requested changes to the WCGS TS. The amendment approved by the NRC on March 17, 2004, revised TS Section 5.5.6, "Containment Tendon Surveillance Program," for consistency with the requirements of 10 CFR 50.55a(g)(4) for components classified as Code Class CC. The amendment to TS 5.5.6 indicated that the Containment Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with ASME BPVC Section XI, Subsection IWL and the applicable addenda as required by 10 CFR 50.55a, except where an exemption or relief has been authorized by the NRC.

The amendment also deleted the provisions of Surveillance Requirement (SR) 3.0.2 from the TS.

In addition, this amendment revised TS 5.5.16 to add the following exceptions to Regulatory Guide (RG) 1.163.

Technical Specification Amendment 152 revised TS 5.5.16 by adding the following exceptions to RG 1.163, "Performance-Based Containment Leak-Test Program":

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

Attachment I to WO 20-0029 Page 20 of 122 NRC SE Section 3.2, Evaluation (Reference 15), stated the following as the basis for acceptability of the requested TS Amendment:

The requirements of 10 CFR 50.55a were amended (61 FR 41303) to incorporate, by reference, Subsections IWE and IWL of Section XI of the ASME Code, for the inspection of containments of light water-cooled reactors. Subsection IWE provides the requirements for inservice inspection, repair, and replacement of Class MC pressure retaining components, and metallic shell and penetration liners of Class CC pressure retaining components, and their integral attachments. Subsection IWL provides the requirements for preservice examination, inservice inspection and repair of the reinforced containments.

For WCGS TS 5.5.16, the licensee stated in its application that:

Currently, WCGS TS 5.5.16 contains requirements for the containment leakage rate testing program, and it specifies that the program shall be in accordance with the guidelines contained in RG 1.163. Regulatory Position C.3 of RG 1.163 states that "Section 9.2.1, "Pretest Inspection and Test Methodology," of NEI 94-01 provides guidance for the visual examination of accessible interior and exterior surfaces for the containment system for structural problems. These examinations should be conducted prior to initiating a Type A test, and during two other refueling outages before the next Type A test, if the interval for the Type A test has been extended to 10 years, in order to allow for early uncovering of evidence of structural deterioration."

There are no specific requirements in NEI 94-01 for the visual examination, except that it is to be a general visual examination of accessible interior and exterior surfaces of the primary containment components.

The licensee proposes to modify TS 5.5.16 to specify that, in addition to the requirements of RG 1.163 and NEI 94-01, the concrete surfaces of the containment must be visually examined in accordance with the ASME Section XI Code, Subsection IWL, and the liner plate inside containment must be visually examined in accordance with Subsection IWE. The frequency of visual examination of the concrete surfaces per Subsection IWL is once every five years, and the frequency of visual examination of the liner plate per Subsection IWE is, in general, three visual examinations over a 10-year period. The visual examinations performed pursuant to Subsection IWL may be performed at any time during power operation or during shutdown, and the visual examinations performed pursuant to Subsection IWE are performed during refueling outages, since this is the only time that the liner plate is fully accessible.

As a result of this modification to TS 5.5.16, one less visual examination will be conducted during the 10-year interval. The licensee indicated, however, that the requirements of Subsection IWE and IWL are more rigorous than those performed pursuant to RG 1.163 and NEI 94-01. Further, the licensee stated that with respect to

Attachment I to WO 20-0029 Page 21 of 122 examinations performed pursuant to both Subsections IWL and IWE, visual examinations of both the concrete surfaces and the liner plate must be reviewed by an inspector regularly employed by an insurance company authorized to write boiler and pressure vessel insurance, in accordance with IWA-2110 and IWA-2120 of the ASME Code. The staff agrees that the combination of the Code requirements for the visual examinations plus the third-party review will offset the fact that there will be one less visual examination during the 10-year interval.

The staff agreed with WCNOC in that the visual examinations of the containment concrete surfaces and the liner plate performed pursuant to Subsections IWL and IWE, respectively, are more rigorous than those performed pursuant to RG 1.163 and NEI 94-01. The requirements in Subsections IWL and IWE of the ASME Code constitute acceptable requirements for the inspection of the concrete surfaces and the liner plate in the WCGS containment, in that the requirements in Subsections IWL and IWE meet 10 CFR 50.55a(b)(2)(vi) and 50.55a(g)(4). Therefore, the staff finds that the changes proposed with respect to TS 5.5.16 are acceptable.

Acceptability of TS Amendment 152 Section 9.2.3.2, Supplemental Inspection Requirements, of NEI 94-01, Revision 3-A, dated July 2012, and Revision 2-A, dated October 2008, both incorporated the inspection requirements of RG 1.163, Regulatory Position C.3 and expanded upon this requirement for ILRT intervals of up to fifteen years in Section 9.2.3.2, Supplemental Inspection Requirements, as follows:

To provide continuing supplemental means of identifying potential containment degradation, a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. It is recommended that these inspections be performed in conjunction or coordinated with the ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE/IWL required examinations.

ASME Section XI, Division 1, Subsection IWE ASME Section XI, Subsection IWE requires the performance of a complete visual examination of the containment liner once each inspection period. If the interval for the Type A test has been extended to 15 years (a minimum of 4 CISI Inspection Periods) a minimum of four complete inspections of the containment liner will be performed during the interval between ILRT performances. This frequency exceeds both the initial requirement of RG 1.163 and NEI 94-01 Section 9.2.3.2 as proposed by TS Amendment 152.

ASME Section XI, Division 1, Subsection IWL ASME Section XI, Subsection IWL requires the performance of a complete visual examination of the containment exterior concrete once every five years. If the interval for the Type A test has been

Attachment I to WO 20-0029 Page 22 of 122 extended to 15 years a minimum of two complete inspections of the containment exterior concrete will be performed during the interval between ILRT performances.

NRC SE Section 3.2 states the following regarding the acceptability of the use of Subsection IWL:

The staff agrees with the licensee in that the visual examinations of the containment concrete surfaces performed pursuant to Subsection IWL, are more rigorous than those performed pursuant to RG 1.163 and NEI 94-01. The staff agrees that the combination of the Code requirements for the visual examinations plus the third-party review will offset the fact that there will be one less visual examination during the 10-year interval.

This satisfies the original intent of the SE in that the combination of the Code requirements for the visual examinations plus the third party review will offset the fact that there will be one less visual examination during the 15-year interval instead of the original one less visual examination during the 10-year interval.

Conclusion It is the position of WCNOC that it will continue to be acceptable to perform the visual inspections of the containment interior and exterior in accordance with the requirements and frequency specified by ASME BPVC,Section XI, Subsections IWE and IWL following the approval of this proposed TS amendment.

No further evaluation of TS Amendment 152 in regard to the proposed activity is required.

3.3.5 Integrated Leakage Rate Testing (ILRT) History As noted previously, WCGS TS 5.5.16 currently requires Types A, B, and C testing in accordance with RG 1.163, which endorses the methodology for complying with 10 CFR 50, Appendix J, Option B. Since the adoption of Option B, the performance leakage rates are calculated in accordance with NEI 94-01, Section 9.1.1 for Type A testing.

Table 3.3.5-1 below, lists the test results of past periodic Type A ILRT tests at WCGS.

Table 3.3.5-1, WCGS ILRT Test History 95% Upper Confidence Limit Test Date (% Containment Air Mass/Day)1 December 1984 - Preoperational AL= 0.058 October 1988 AF= 0.1122, AL = 0.1121 September 1991 AF = 0.21692, AL = 0.0704 October 2000 AF = 0.1400, AL = 0.1328 May 2011 AF = 0.0824, AL = 0.0823

Attachment I to WO 20-0029 Page 23 of 122 Table 3.3.5-1, WCGS ILRT Test History 95% Upper Confidence Limit Test Date (% Containment Air Mass/Day)1 Notes:

(1) As specified in WCGS Technical Specification 5.5.16, the maximum allowable containment leakage rate La, at Pa of 48.0 psig, is 0.2% of primary containment air weight per day.

Additionally, TS 5.5.16 specifies that during the first startup following Type A testing, acceptable leakage test criteria shall be less than 0.75 La or 0.15 wt%/day.

(2) As Found ILRT leakage computation was penalized by Type C LLRT on Penetrations P-28 (ESW) and P-190 (Purge Exhaust). Specifically, the As-Found Type C LLRT on penetration P-28 was 310,043 sccm and was corrected to an As-Left of 2732 sccm. The As-Found for P-160 was 1150 sccm and was corrected to an aAs-Left of 750 sccm. These two As-Found LLRT tests when combined, penalized (leakage savings addition) the ILRT As-Found test results by contributing to a value of 0.1465 wt%/day of the computed 0.2169 wt%/day leakage value. WCGS program procedure addressed the impact (penalty) of individual LLRT excessive leakage on the ILRT overall summation and the deferral of the penalty to the Type B and C testing program (instead of Type A) along with appropriate corrective actions.

Table 3.3.5-2 below, provides the two most recent ILRT containment tests that encompass the 10-year test interval along with the specific test data. The current ILRT test interval for WCGS is ten years. This interval is verified, and the test data is presented in Table 3.3.5-2.

The acceptance criteria for this verification is contained in NEI 94-01, Revisions 2-A and 3-A, Section 5.0, Definitions, and is as follows:

"The performance leakage rate is calculated as the sum of the Type A upper confidence limit (UCL) and as-left minimum pathway leakage rate (MNPLR) leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position (i.e., drained and vented to containment atmosphere) prior to performing the Type A test. In addition, leakage pathways that were isolated during performance of the test because of excessive leakage must be factored into the performance determination. The performance criterion for Type A tests is a performance leak rate of less than 1.0La."

Attachment I to WO 20-0029 Page 24 of 122 Table 3.3.5-2, WCGS ILRT Test Results Verification of Current Extended ILRT 10-year Interval Test Upper Volume As Left Adjusted ILRT ILRT Test 95% Level Min As Left Method/Da Date Confiden Correctio Pathway Leak Total Acceptan ta ce Level ns Penalty Rate ce Analysis As Left Criteria for (wt.%/da Leakag Technique (wt.%/day (wt.%/day Isolated y) s

) ) e Pathway

[Test s (wt.%/

Pressure] day)

(wt.%/da y) 0.75 La Absolute Octob 0.1175 Method er -0.0001 0.0000 0.0154 0.1328 (0.150 /Mass

[48.46 2000 weight Point psig]

%/day) Analysis 0.75 La Absolute 0.0754 Method May (0.150

[48.84 0.0000 0.0000 0.0068 0.0822 /Mass 2011 weight psig] Point

%/day) Analysis 3.4 Plant Specific Confirmatory Analysis 3.4.1 Methodology A plant specific confirmatory analysis was performed to provide a risk assessment of permanently extending the currently allowed containment Type A ILRT from ten years to fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for WCGS. The risk assessment follows the guidelines from the following:

• NEI 94-01, Revision 3-A (Reference 2)

• NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals from November 2001 (Reference 20)

Attachment I to WO 20-0029 Page 25 of 122

• NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 Revision 2 as applied to ILRT interval extensions (Reference 4

• Risk insights in support of a request for a plants licensing basis as outlined in Regulatory Guide (RG) 1.174 (Reference 3)

• The methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval (Reference 32)

• The methodology used in EPRI 1018243, Revision 2-A of EPRI 1009325 (Reference 34).

Revisions to 10 CFR 50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing frequency requirement from three in ten years to at least once in ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage rate was less than the limiting containment leakage rate of 1.0 La.

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of the performance-based Option B to Appendix J.

Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, September 1995 (Reference 6), provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals (Reference 7).

The NRC report on performance-based leak testing, NUREG-1493, analyzed the effects of containment leakage on the health and safety of the public and the benefits realized from the containment leak rate testing. In that analysis, it was determined that for a representative PWR plant (i.e., Surry), containment isolation failures contribute less than 0.1% to the latent risks from reactor accidents. Consequently, it is desirable to show that extending the ILRT interval will not lead to a substantial increase in risk from containment isolation failures for WCGS.

NEI 94-01 Revision 3-A supports using EPRI Report No. 1009325 Revision 2-A (EPRI 1018243),

Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, for performing risk impact assessments in support of ILRT extensions (Reference 34). The Guidance provided in Appendix H of EPRI Report No. 1009325 Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the ASME BPVC Section XI requirements. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class

Attachment I to WO 20-0029 Page 26 of 122 MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment.

The associated change to NEI 94-01 requires that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements are not changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.

In the NRC SE dated June 25, 2008 (Reference 9), the NRC concluded that the methodology in EPRI TR-1009325, Revision 2, was acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.0 of the SE. Table 3.4.1-1 below addresses each of the four (4) limitations and conditions from Section 4.2 of the SE for the use of EPRI 1009325, Revision 2.

Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition WCGS Response (From Section 4.2 of SE)

1. The licensee submits documentation WCGS PRA technical adequacy is addressed indicating that the technical adequacy of in Section 3.4.2 of this LAR and Enclosure, "

their PRA is consistent with the Evaluation of Risk Significance of Permanent requirements of RG 1.200 relevant to the ILRT Extension, Appendix A, "PRA ILRT extension application. Acceptability."

Attachment I to WO 20-0029 Page 27 of 122 Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition WCGS Response (From Section 4.2 of SE) 2.a The licensee submits documentation Since the ILRT does not impact core indicating that the estimated risk increase damage frequency (CDF), the relevant associated with permanently extending criterion is large early release frequency the ILRT surveillance interval to 15 years (LERF). The increase in LERF resulting is small, and consistent with the from a change in the Type A ILRT test clarification provided in Section 3.2.4.5 of interval from 3 in 10 years to 1 in 15 years is this SE. estimated as 1.48E-7/year using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Total Internal Events LERF (baseline and change in LERF due to the ILRT extension) is 2.58E-7. Therefore, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174.

When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 2.81E-7/year using the EPRI guidance, and total LERF is 4.89E-7/year. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174. (See Enclosure, Section 7.0 of this submittal)

Attachment I to WO 20-0029 Page 28 of 122 Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition WCGS Response (From Section 4.2 of SE) 2.b Specifically, a small increase in The effect resulting from changing the Type population dose should be defined as an A test frequency to 1-per-15 years, increase in population dose of less than measured as an increase to the total or equal to either 1.0 person-rem per year integrated plant risk for those accident or 1% of the total population dose, sequences influenced by Type A testing is whichever is less restrictive. 0.645 person-rem/year. NEI 94-01 (Reference 8) states that a small population dose is defined as an increase of 1.0 person-rem per year, or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria.

Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible. (See Enclosure, Section 7.0 of this submittal) 2.c In addition, a small increase in CCFP The increase in the conditional containment should be defined as a value marginally failure probability (CCFP) from the 3 in 10 greater than that accepted in a previous year interval to 1 in 15 year interval is one-time 15-year ILRT extension 0.911%. NEI 94-01 (Reference 8) states requests. This would require that the that increases in CCFP of 1.5% is small.

increase in CCFP be less than or equal to Therefore, this increase is judged to be 1.5 percentage point. small. (See Enclosure, Section 7.0 of this submittal)

3. The methodology in EPRI Report No. The representative containment leakage for 1009325, Revision 2, is acceptable except Class 3b sequences used by WCGS is 100 for the calculation of the increase in La, based on the guidance provided in EPRI expected population dose (per year of Report No. 1009325, Revision 2-A. (See reactor operation). In order to make the Enclosure, Section 4.0 of this submittal) methodology acceptable, the average leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

Attachment I to WO 20-0029 Page 29 of 122 Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition WCGS Response (From Section 4.2 of SE)

4. A licensee amendment request (LAR) is Containment overpressure is not required for required in instances where containment ECCS performance and is discussed in over-pressure is relied upon for ECCS Section 3.2 of this submittal. Therefore, no performance. additional request is needed.

3.4.2 PRA Acceptability 3.4.2.1 PRA Quality Statement for Permanent 15-Year ILRT Extension Revision 9 of the WCGS PRA model is the most recent evaluation of internal event risk. The WCGS PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the WCGS PRA is based on the single top fault tree methodology, which is a well-known PRA methodology in the industry.

The internal events PRA model (Reference 24) has been assessed against RG 1.200 (Reference 4). The internal events PRA model was subject to a full-scope peer review conducted in accordance with RG 1.200 in June 2019 (Reference 25). The independent assessment of the findings and observations (F&O) closures was performed between November 2019 and March 2020, including a two-day meeting at the Wolf Creek site in Burlington, Kansas, on December 10 and 11, 2019 (Reference 26).

The F&O closure independent assessment was performed following the guidance of Appendix X of the Nuclear Energy Institute (NEI) peer review guidance document NEI 05-04 (Reference 27) as well as the process clarifications provided by the United States Nuclear Regulatory Commission (USNRC), which are documented in Reference 28. This independent assessment also considered the lessons learned from the USNRC staff observations of three independent assessment team pilot reviews (Reference 33). After this F&O closure, three of the original peer review F&Os remained open. During the F&O closure review, one F&O was judged to be closed with a PRA upgrade, which required a focused scope peer review. This triggered a focused scope peer review of the Supporting Requirements associated with the upgrade. Following the focused scope peer review, all the involved SRs have been judged to be met at Capability Category II or higher, but a new F&O (AS-B3-01) has been assigned and remains open. The four open F&Os are detailed in the Enclosure to this submittal in Section A.2. F&Os 3-8 and 6-8 remain open following the closure assessment. Since F&Os 3-8 and 6-8 relate to documentation, they do not affect the PRA acceptability for use in the ILRT extension analysis. F&O 4-10 was not reviewed during this F&O closure. One part of F&O 4-10 pertains to the ISLOCA modeling only meeting Capability Category

Attachment I to WO 20-0029 Page 30 of 122 I, but this does not negatively affect the acceptability of the PRA for the ILRT extension analysis.

The other part of F&O 4-10 relates to documentation, so it does not affect the PRA acceptability for use in the ILRT extension analysis. New F&O AS-B3-01 pertains to the disposition of the potential for sump strainer blockage during feed and bleed events with open PORVs. A bounding sensitivity study, detailed in the Enclosure, Section A.2, demonstrates no impact on the conclusions of this analysis. Therefore, the PRA is of sufficient quality and level of detail to support the ILRT extension analysis.

WCNOC employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models. This approach includes both a proceduralized PRA maintenance and update process and the use of self-assessments and independent peer reviews.

3.4.2.2 PRA Maintenance and Update The WCNOC risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant. The process delineates the responsibilities and guidelines for updating the PRA models, and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, and industry operational experience), for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files. The process will assess the impact of these changes on the plant PRA model in a timely manner.

3.4.3 Summary of Plant-Specific Risk Assessment Results Based on the results from Enclosure, Section 5.2 and the sensitivity calculations presented in Enclosure, Section 5.3, the following conclusions regarding the assessment of the plant risk are associated with extending the Type A ILRT test frequency to 15 years:

• RG 1.174 (Reference 3) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines small changes in risk as resulting in increases of CDF greater than 1.0E-6/year and less than 1.0E-5/year and increases in LERF greater than 1.0E-7/year and less than 1.0E-6/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 1.48E-7/year using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Total Internal Events LERF (baseline and change in LERF due to the ILRT extension) is 2.58E-7. Therefore, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174. The risk change resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change. Considering the increase in LERF resulting from a

Attachment I to WO 20-0029 Page 31 of 122 change in the Type A ILRT test interval from 1 in 10 years to 1 in 15 years is estimated as 6.18E-8/year, the risk increase is very small using the acceptance guidelines of RG 1.174.

• When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 2.81E-7/year using the EPRI guidance, and total LERF is 4.89E-7/year. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of RG 1.174. The risk change resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change. When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 10 years to 1 in 15 years is estimated as 1.17E-7/year, and the total LERF is 3.25E-7/year. Therefore, the risk increase is small using the acceptance guidelines of RG 1.174.

• The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing is 0.645 person-rem/year. NEI 94-01 (Reference 2) states that a small population dose is defined as an increase of 1.0 person-rem per year, or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

• The increase in the conditional containment failure probability (CCFP) from the 3 in 10-year interval to 1 in 15-year interval is 0.911% for WCGS. NEI 94-01 states that increases in CCFP of 1.5% is small. Therefore, this increase is judged to be small.

Therefore, increasing the ILRT interval to 15 years is considered to be small since it represents a small change to the WCGS risk profile.

3.4.4 Previous Assessments The NRC in NUREG-1493 (Reference 6) has previously concluded that:

• Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

• Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.

Attachment I to WO 20-0029 Page 32 of 122 The conclusions for WCGS confirm these general conclusions on a plant-specific basis considering the severe accidents evaluated for WCGS, the WCGS containment failure modes, and the local population surrounding WCGS.

Details of the WCGS risk assessment are contained in the Enclosure to this LAR submittal.

3.4.5 RG 1.174, Revision 3, Defense in Depth Evaluation RG 1.174, Revision 3 (Reference 3), describes an approach that is acceptable for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights. One of the considerations included in RG 1.174 is Defense in Depth. Defense in Depth is a safety philosophy that employs successive compensatory measures to prevent accidents or mitigate damage if a malfunction, accident, or naturally caused event occurs at a nuclear facility. The following seven considerations, as presented in RG 1.174, Revision 3, Section C.2.1.1.2, will serve to evaluate the proposed licensing basis change for overall impact on Defense in Depth.

1. Preserve a reasonable balance among the layers of defense.

A reasonable balance of the layers of defense (i.e., minimizing challenges to the plant, preventing any events from progressing to core damage, containing the radioactive source term, and emergency preparedness) helps to ensure an apportionment of the plants capabilities between limiting disturbances to the plant and mitigating their consequences.

The term reasonable balance is not meant to imply an equal apportionment of capabilities.

The NRC recognizes that aspects of a plants design or operation might cause one or more of the layers of defense to be adversely affected. For these situations, the balance between the other layers of defense becomes especially important when evaluating the impact of the proposed licensing basis change and its effect on defense in depth.

Response

Several layers of defense are in place to ensure the WCGS containment structure(s),

penetrations, isolation valves, and mechanical seal systems continue(s) to perform their intended safety function. The purpose of the proposed change is to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and Type C LLRTs for selected components from 60-months to 75-months.

As shown in NUREG-1493, Performance-Based Containment Leak-Test Program (Reference 6), increasing the test frequency of ILRTs up to a 20-year test interval was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing. The study also concluded that extending the frequency of Type B tests is possible with no adverse impact on risk as identified leakage

Attachment I to WO 20-0029 Page 33 of 122 through Type B mechanical penetrations are both infrequent and small. Finally, the study concluded that Types B and C tests could identify the vast majority (greater than 95 percent) of all potential leakage paths.

Several programmatic factors can also be cited as layers of defense ensuring the continued safety function of the WCGS containment pressure boundary. NEI 94-01 Revisions 2-A and 3-A require sites adopting the 15-year extended ILRT interval perform visual examinations of the accessible interior and exterior surfaces of the containment structure for structural degradation that may affect the containment leak-tight integrity at the frequency prescribed by the guidance or, if approved through a Technical Specification (TS) amendment, at the frequencies prescribed by ASME Section XI. Additionally, several measures are put in place to ensure integrity of the Type B and C tested components. NEI 94-01 limits large containment penetrations such as airlocks, hatches, purge and vent valves, to a maximum 30-month testing interval. For those valves that meet the performance standards defined in NEI 94-01, Revision 3-A, and are selected for test intervals greater than 60 months, a leakage understatement penalty is added to the MNPLR prior to the frequency being extended beyond 60-months. Finally, identification of adverse trends in the overall Types B and C leakage rate summations and available margin between the Type B and Type C leakage rate summation and its regulatory limit are required by NEI 94-01, Revision 3-A, to be shown in the WCGS post-outage report(s). Therefore, the proposed change does not challenge or limit the layers of defense available to assess the ability of the WCGS containment structure to perform its safety function.

PRA Response:

The use of the risk metrics of LERF, population dose, and conditional containment failure probability collectively ensures the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. The change in LERF is small with respect to internal events and small when including external events per RG 1.174, and the change in population dose and CCFP are small as defined in this analysis and consistent with NEI 94-01 Revision 3-A.

2. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Nuclear power plant licensees implement a number of programmatic activities, including programs for quality assurance, testing and inspection, maintenance, control of transient combustible material, foreign material exclusion, containment cleanliness, and training. In some cases, activities that are part of these programs are used as compensatory measures; that is, they are measures taken to compensate for some reduced functionality, availability, reliability, redundancy, or other feature of the plants design to ensure safety functions (e.g.,

reactor vessel inspections that provide assurance that reactor vessel failure is unlikely).

NUREG-2122, Glossary of Risk-Related Terms in Support of Risk-Informed Decisionmaking, (Reference 23), defines safety function as those functions needed to

Attachment I to WO 20-0029 Page 34 of 122 shut down the reactor, remove the residual heat, and contain any radioactive material release.

A proposed licensing basis change might involve or require compensatory measures.

Examples include hardware (e.g., skid-mounted temporary power supplies); human actions (e.g., manual system actuation); or some combination of these measures. Such compensatory measures are often associated with temporary plant configurations. The preferred approach for accomplishing safety functions is through engineered systems.

Therefore, when the proposed licensing basis change necessitates reliance on programmatic activities as compensatory measures, the licensee should justify that this reliance is not excessive (i.e., not overly reliant). The intent of this consideration is not to preclude the use of such programs as compensatory measures but to ensure that the use of such measures does not significantly reduce the capability of the design features (e.g.,

hardware).

Response

The purpose of the proposed change is to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and select Type C LLRTs from 60-months to 75-months.

Several programmatic factors were defined in the response to Question 1 above, which are required when adopting NEI 94-01, Revisions 2-A and 3-A. These factors are conservative in nature and are designed to generate corrective actions if the required testing or inspections are deemed unsatisfactory well in advance to ensure the continued safety function of the containment is maintained. The programmatic factors are designed to provide differing ways to test and/or examine the containment pressure boundary in a manner that verifies the WCGS containment pressure boundary will perform its intended safety function. Since the proposed change does not alter the configuration of the WCGS containment pressure boundary, continued performance of the tests and inspections associated with NEI 94-01 will only serve to ensure the continued safety function of the containment without affecting any margin of safety.

PRA Response:

The adequacy of the design feature (the containment boundary subject to Type A testing) is preserved as evidenced by the overall small change in risk associated with the Type A test frequency change.

3. Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

As stated in Section C.2.1.1.1 above, the defense-in-depth philosophy has traditionally been applied in plant design and operation to provide multiple means to accomplish safety functions. System redundancy, independence, and diversity result in high availability and reliability of the function and also help ensure that system functions are not reliant on any single feature of the design. Redundancy provides for duplicate equipment that enables

Attachment I to WO 20-0029 Page 35 of 122 the failure or unavailability of at least one set of equipment to be tolerated without loss of function. Independence of equipment implies that the redundant equipment is separate such that it does not rely on the same supports to function. This independence can sometimes be achieved by the use of physical separation or physical protection. Diversity is accomplished by having equipment that performs the same function rely on different attributes such as different principles of operation, different physical variables, different conditions of operation, or production by different manufacturers which helps reduce common-cause failure (CCF).

A proposed change might reduce the redundancy, independence, or diversity of systems.

The intent of this consideration is to ensure that the ability to provide the system function is commensurate with the risk of scenarios that could be mitigated by that function. The consideration of uncertainty, including the uncertainty inherent in the PRA, implies that the use of redundancy, independence, or diversity provides high reliability and availability and also results in the ability to tolerate failures or unanticipated events.

Response

The proposed change to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and select Type C LLRTs from 60-months to 75-months does not reduce the redundancy, independence or diversity of systems. As shown in NUREG-1493, increasing the test frequency of ILRTs up to a 20-year test interval was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing. The study also concluded that extending the frequency of Type B tests is possible with no adverse impact on risk as identified leakage through Type B mechanical penetrations is both infrequent and small. Additionally, the study concluded that Types B and C tests could identify the vast majority (greater than 95 percent) of all potential leakage paths.

Despite the change in test interval, containment isolation diversity remains unaffected and will continue to provide the inherent isolation, as designed. In addition, NEI 94-01, Revisions 2-A and 3-A, Section 11.3.2 requires a schedule of tests be developed, for components on a test interval greater than 60 months, such that unanticipated random failures and unexpected common-mode failures are avoided. This is typically accomplished by implementing test intervals at approximately evenly distributed intervals. Therefore, the proposed change preserves system redundancy, independence, and diversity and ensures a high reliability and availability of the containment structure to perform its safety function in the event of unanticipated events.

PRA Response:

The redundancy, independence, and diversity of the containment subject to the Type A test is preserved, commensurate with the expected frequency and consequences of challenges to the system, as evidenced by the overall small change in risk associated with the Type A test frequency change.

Attachment I to WO 20-0029 Page 36 of 122

4. Preserve adequate defense against potential CCFs.

An important aspect of ensuring defense in depth is to guard against CCF. Multiple components may fail to function because of a single specific cause or event that could simultaneously affect several components important to risk. The cause or event may include an installation or construction deficiency, accidental human action, extreme external environment, or an unintended cascading effect from any other operation or failure within the plant. CCFs can also result from poor design, manufacturing, or maintenance practices.

Defenses can prevent the occurrence of failures from the causes and events that could allow simultaneous multiple component failures. Another aspect of guarding against CCF is to ensure that an existing defense put in place to minimize the impact of CCF is not significantly reduced; however, a reduction in one defense can be compensated for by adding another.

Response

As part of the proposed change, WCGS will be required to adopt the performance-based testing standards outlined in NEI 94-01, Revisions 2-A and 3-A, along with ANSI/ANS 56.8-2002. NEI 94-01, Revisions 2-A and 3-A, Section 11.3.2 requires a schedule of tests be developed, for components on test intervals greater than 60 months, such that unanticipated random failures and unexpected common-mode failures are avoided. This is typically accomplished by implementing test intervals at approximately evenly distributed intervals. In addition, components considered to be risk-significant from a PRA standpoint are required to be limited to a testing interval less than the maximum allowable limit of 75-months. For those components that have demonstrated satisfactory performance and have had their testing limits extended, administrative testing limits are assigned on a component-by-component basis and are used to identify potential valve or penetration degradation.

Administrative limits are established at a value low enough to identify and should allow early correction in advance of total valve failure. Should a component exceed its administrative limit during testing, NEI 94-01, Revisions 2-A and 3-A, require cause determinations be performed, which are designed to reinforce achieving acceptable performance. The cause determination is designed to identify and address common-mode failure mechanisms through appropriate corrective actions. The proposed change also imposes a requirement to address margin management (i.e., margin between the current containment leakage rate and its pre-established limit). As a result, adoption of the performance-based testing standards proposed by this change ensures adequate barriers exist to preclude failure of the containment pressure boundary due to common-mode failures and therefore continues to guard against CCF.

PRA Response:

Adequate defense against CCFs is preserved. The Type A test detects problems in the containment which may or may not be the result of a CCF; such a CCF may affect failure of another portion of containment (i.e., local penetrations) due to the same phenomena.

Adequate defense against CCFs is preserved via the continued performance of the Type B

Attachment I to WO 20-0029 Page 37 of 122 and C tests and the performance of inspections. The change to the Type A test, which bounds the risk associated with containment failure modes including those involving CCFs, does not degrade adequate defense as evidenced by the overall small change in risk associated with the Type A test frequency change.

5. Maintain multiple fission product barriers.

Fission product barriers include the physical barriers themselves (e.g., the fuel cladding, reactor coolant system pressure boundary, and containment) and any equipment relied on to protect the barriers (e.g., containment spray). In general, these barriers are designed to perform independently so that a complete failure of one barrier does not disable the next subsequent barrier. For example, one barrier, the containment, is designed to withstand a double-ended guillotine break of the largest pipe in the reactor coolant system, another barrier.

A plants licensing basis might contain events that, by their very nature, challenge multiple barriers simultaneously. Examples include interfacing-system loss-of-coolant accidents, steam generator tube rupture, or crediting containment accident pressure. Therefore, complete independence of barriers, while a goal, might not be achievable for all possible scenarios.

Response

The purpose of the proposed change is to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and select Type C LLRTs from 60-months to 75-months.

As part of the proposed change, WCGS will be required to adopt the performance-based testing standards outlined in NEI 94-01, Revisions 2-A and 3-A, along with ANSI/ANS 56.8-2002. The overall containment leakage rate calculations associated with the testing standards contain inherent conservatisms through the use of margin. Plant TS require the overall primary containment leakage rate to be less than or equal to 1.0 L a. NEI 94-01 requires the As-Found Type A test leakage rate must be less than the acceptance criterion of 1.0 La given in the plant TS. Prior to entering a mode where containment integrity is required, the As-Left Type A leakage rate shall not exceed 0.75 La. The As-Found and As-Left values are as determined by the appropriate testing methodology specifically described in ANSI/ANS-56.8-2002. Additionally, the combined leakage rate for all Type B and Type C tested penetrations shall be less than or equal to 0.6 La, determined on a maximum pathway basis from the As-Left LLRT results prior to entering a mode where containment integrity is required. This regulatory approach results in a 25% and 40% margin, respectively, to the 1.0 La requirements. For those local leak rate tested components that have demonstrated satisfactory performance and have had their testing limits extended, administrative testing limits are assigned on a component by component basis and are used to identify potential valve or penetration degradation. Administrative limits are established at a value low enough to identify and allow early correction in advance of total valve failure.

Should a component exceed its administrative limit during testing, NEI 94-01, Revisions 2-

Attachment I to WO 20-0029 Page 38 of 122 A and 3-A, require cause determinations be performed designed to reinforce achieving acceptable performance. The cause determination is designed to identify and address common-mode failure mechanisms through appropriate corrective actions. Therefore, the proposed change adopts requirements with inherent conservatisms to ensure the margin to safety limit is maintained, thereby, preserving the containment fission product barrier.

PRA Response:

Multiple Fission Product barriers are maintained. The portion of the containment affected by the Type A test extension is still maintained as an independent fission product barrier, albeit with an overall small change in the reliability of the barrier.

6. Preserve sufficient defense against human errors.

Human errors include the failure of operators to correctly and promptly perform the actions necessary to operate the plant or respond to off-normal conditions and accidents, errors committed during test and maintenance, and incorrect actions by other plant staff. Human errors can result in the degradation or failure of a system to perform its function, thereby significantly reducing the effectiveness of one of the layers of defense or one of the fission product barriers. The plant design and operation include defenses to prevent the occurrence of such errors and events. These defenses generally involve the use of procedures, training, and human engineering; however, other considerations (e.g.,

communication protocols) might also be important.

Response

Sufficient defense against human errors is preserved. Errors committed during testing and maintenance may be reduced by the less frequent performance of the Type A, Type B, and Type C tests (less opportunity for errors to occur).

PRA Response:

Sufficient defense against human errors is preserved. The probability of a human error to operate the plant, or to respond to off-normal conditions and accidents is not significantly affected by the change to the Type A testing frequency. Errors committed during test and maintenance may be reduced by the less frequent performance of the Type A test (less opportunity for errors to occur).

7. Continue to meet the intent of the plants design criteria.

For plants licensed under 10 CFR Part 50 or 10 CFR Part 52, the plants design criteria are set forth in the current licensing basis of the plant. The plants design criteria define minimum requirements that achieve aspects of the defense-in-depth philosophy; as a consequence, even a compromise of the intent of those design criteria can directly result in a significant reduction in the effectiveness of one or more of the layers of defense. When

Attachment I to WO 20-0029 Page 39 of 122 evaluating the effect of the proposed licensing basis change, the licensee should demonstrate that it continues to meet the intent of the plants design criteria.

Response

The purpose of the proposed change is to extend the testing frequencies of the Type A ILRT from 10 years to 15 years and select Type C LLRTs from 60-months to 75-months.

The proposed extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. As part of the proposed change, WCGS will be required to adopt the performance-based testing standards outlined in NEI 94-01, Revisions 2-A and 3-A, along with ANSI/ANS 56.8-2002. The leakage limits imposed by plant TS remain unchanged when adopting the performance-based testing standards outlined in NEI 94-01, Revision 3-A, and ANSI/ANS 56.8-2002. Plant design limits imposed by the Updated Safety Analysis Report (USAR) also remain unchanged as a result of the proposed change. Therefore, the proposed change continues to meet the intent of the plants design criteria to ensure the integrity of the WCGS containment pressure boundary.

PRA Response:

The intent of the plants design criteria continues to be met. The extension of the Type A test does not change the configuration of the plant or the way the plant is operated.

==

Conclusion:==

The responses to the seven Defense in Depth questions above concludes that the existing Defense in Depth for WCGS has not been diminished, rather, in some instances has been increased.

Therefore, the proposed change does not comprise a reduction in safety.

3.5 Non-Risk Based Assessment Consistent with the defense-in-depth philosophy discussed in RG 1.174, WCGS has assessed other non-risk-based considerations relevant to the proposed amendment. WCGS has multiple inspection and testing programs that ensure the containment structure continues to remain capable of meeting its design functions and is designed to identify any degrading conditions that might affect that capability. These programs are discussed in the following paragraphs of this section.

3.5.1 Nuclear Coatings Program The WCGS Containment Coating Condition Assessment Program is implemented by procedures intended to provide condition monitoring inspection guidance of coatings applied to systems, structures and components within the containment structure.

Coatings assessed are those that have been classified as safety related (Service Level 1) Coatings in the primary containment system. This includes coatings applied to walls, floors, structural steel, exposed piping, and other equipment inside primary containment.

Attachment I to WO 20-0029 Page 40 of 122 Coating Condition Monitoring is performed during each Refueling Outage on all safety-related coatings in containment in order to assure that the coatings are still performing their intended design function and are still considered qualified.

Coating Condition Monitoring is performed in accordance with the requirements established in RG 1.54 (Service Level I, II, and III Protective Coatings Applied to Nuclear Power Plants) as supplemented by;

• 10 CFR 50.65, Maintenance Rule;

• ASTM-D5163-05A, "Standard Guide for Establishing Procedures to Monitor the Performance of Coating Service Level 1 Coating Systems in an Operating Nuclear Power Plant"; and

• EPRI Document "Guideline on Nuclear Safety-Related Coatings" (EPRI 1019157 formerly TR-109937 and 1003102), with the exception of the approval of the ANSI Certified Inspector. The inspector shall meet the qualification as outlined below.

Qualifications of Coatings Surveillance Personnel are at a minimum of a Level II Coatings Inspector as per ANSI N45.2.6, "Qualification of Inspection, Examination and Testing Personnel for Nuclear Power Plants".

WCGS plant procedure, Containment Coatings Condition Assessment Program, provides condition monitoring inspection guidance of coatings applied to systems, structures, and components within the containment structure. This procedure provides details of the containment coating inspections performed.

Each refueling outage a general walkdown inspection of the containment coating systems is performed. In addition, a detailed visual inspection is performed every refuel outage of approximately one third of the containment coating system area. A review of previous inspections is conducted to determine the areas for subsequent outage inspections. Inspections are performed for signs of peeling, cracking, blistering, or delamination.

If degraded or nonconforming conditions are identified, Work Requests (WRs) are generated detailing the location and extent of condition and corrective actions performed.

3.5.1.1 Unqualified / Degraded Coatings in Containment An engineering analyses was performed on unqualified or degraded coatings on components and equipment located inside containment. This coating analysis is documented in WCGS Engineering Calculation, Containment Un-Qualified Coatings That Could Reach the Sump Strainers. If coating inspections determine during an inspection or if equipment changes are made which are

Attachment I to WO 20-0029 Page 41 of 122 unresolved or uncorrected before returning to service, an evaluation is performed by Engineering and documented in the unqualified/degraded Engineering Calculation. This document is maintained as a quality record. Areas added or dispositioned and removed from the Containment Un-Qualified Coating Calculation are documented with the reason for removal along with an associated Condition Report (CR) or Work Order (WO).

The current status of WCGS containment is documented in Engineering Calculation Qualified Coatings that Could Reach the Sump Strainers. The engineering calculation was revised at the end of RF23 (Fall 2019). The revision temporarily updated the amount of unqualified coating inside containment. The total margin allowance for all coating that could fail and reach the sump strainers was reduced to 4,995 ft2.

3.5.2 Containment Inservice Inspection (CISI) Program The WCGS Containment Inservice Inspection Program Plan details the requirements for the examination and testing of Class MC and Class CC components in accordance with the ASME BPVC,Section XI, Division 1, 2013 Edition, and the Code of Federal Regulations, Title 10, Part 50, Paragraph 55a, Codes and standards (10 CFR 50.55a).

The CISI Program Plan includes components subject to examination per ASME BPVC Section XI, Subsection IWE (Class MC components and the metal liner of concrete containments) and Subsection IWL (Class CC concrete components) (References 30 and 31, respectively).

This CISI Program Plan addresses the current 10-year interval that started on September 10, 2018 and ends September 9, 2028 for both the IWE and IWL activities. This Plan is for the third interval and was developed in accordance with the 2013 Edition of ASME Section XI, Subsection IWA (General Requirements), Subsection IWE (Requirements for ISI Class MC and Metallic Liners of ISI Class CC Components), and Subsection IWL (Requirements for ISI Class CC Concrete Components), along with the appropriate conditions of 10 CFR 50.55a. The requirement to implement the 2013 Edition of ASME Section XI is included in 10 CFR 50.55a(g)(4)(ii) and is based on an effective date of this edition of the Code of September 10, 2017, which is 12 months before the start of the WCGS third interval for the CISI Program.

Basis for Containment ISI Program The WCGS CISI Program basis is comprised of meeting three applicable source requirements documents, they are; ASME Section XI (Subsections IWA, IWE, and IWL), 2013 Edition rules and requirements; 10 CFR 50.55a conditions and requirements; and License Renewal containment applicable commitments.

These requirements are each further described in further detail, as follows:

Attachment I to WO 20-0029 Page 42 of 122

1. ASME Section XI 2013 Edition Requirements The ASME Section XI requirements applicable to the WCGS Containment ISI Program Plan are included in the 2013 Edition of the ASME Section XI, Subsections IWA, IWE, and IWL. WCGS utilizes the Inspection Program of IWE-2411 for Subsection IWE and IWL-2400 for Subsection IWL examinations.

ASME Section XI Code Case Utilized:

The alternative requirements included in the following ASME Section XI Code Case will be utilized:

• N-532-5, Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000,Section XI, Division 1 (Listed as approved in RG 1.147). Code Case N-532-5 requires an Owners Activity Report Form OAR-1 to be prepared and certified upon completion of each refueling outage. The OAR-1 forms must be submitted to the NRC within 90 days of the completion of the refueling outage.

2. 10 CFR 50.55a Conditions and Requirements WCGS CISI Program meets the requirements of ASME Section XI, as well as any conditions listed in 10 CFR 50.55a affecting the Program. The specific conditions are included in 10 CFR 50.55a(g)(4) and references the following supplementary conditions;

• 10 CFR 50.55a(b)(2)(viii),Section XI condition: Concrete containment Examinations,

• 10 CFR 50.55a(b)(2)(ix),Section XI condition: Metal containment examinations, and

• 10 CFR 50.55a(g)(6)(ii)(B), Augmented ISI requirements: Submitting Containment ISI Programs.

These 10 CFR 50.55a requirements are as follows:

• 10 CFR 50.55a(b)(2)(viii)(H) requires each inaccessible Class CC area identified for evaluation under IWL-2512(a), or identified as susceptible to deterioration under IWL-2512(b) must be evaluated as specified in paragraph 10 CFR 50.55a(b)(2)(viii)(E).

WCGS Implementation:

If the conditions described in (2)(viii)(H) described above, are identified at WCGS, this requirement creates an additional scope of components [above the scope described in IWL-1000, which exempts inaccessible concrete surface areas from examination in IWL-1220(b)]. The components that meet the conditions of IWL-2512(a) or IWL-2512(b) will be added as, Examination Category L-A, Inaccessible Below-Grade Areas, Item Number L1.13, components.

Attachment I to WO 20-0029 Page 43 of 122 Currently, no suspect areas have been identified for examination. During future inspections, if areas are determined to be suspect by the Responsible Engineer, these areas will be listed in Appendix C of the CISI Program Procedure and a detailed visual examination will be performed in accordance with the guidance in IWL-2510.

• 10 CFR 50.55a(b)(2)(viii)(I) requires the performance of a technical evaluation under IWL-2512(b) of inaccessible below-grade Class CC surfaces exposed to soil, backfill, or groundwater at periodic intervals not to exceed 5 years during the period of extended operation of a renewed license under Part 54 of 10 CFR 50. 10 CFR 50.55a(b)(2)(viii)(I) also requires that representative samples must be examined when portions of the below-grade concrete is exposed by excavation for any reason.

WCGS Implementation:

The technical evaluation of inaccessible below-grade Class CC areas will not exceed a 5-year period before being reevaluated during the period of extended operation for WCGS [the periodic interval specified by IWL-2512(b) is 10 years].

(Note: WCGS will enter the period of extended operation under License Renewal on March 12, 2025; therefore, 10 CFR 50.55a(b)(2)(viii)(I) will be applied to the CISI Program at that time.)

• 10 CFR 50.55a(b)(2)(ix)(A)(2) requires licensees to provide certain information in the ISI Summary Report regarding inaccessible Class MC areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.

WCGS Implementation:

If the conditions described in (ix)(A)(2) above are identified at WCGS, they will be included in the ISI Summary Report. Information provided will be; (ix)(A)(2)(i) - A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (ix)(A)(2)(ii) - An evaluation of each area, and the result of the evaluation; and, (ix)(A)(2)(iii) - A description of necessary corrective actions.

• 10 CFR 50.55a(b)(2)(ix)(B) allows licensees, when performing remotely the visual examinations required by Subsection IWE, to extend the maximum direct examination distance specified in Table IWA-2211-1, Visual Examinations, and to decrease the minimum illumination requirements specified in Table IWA-2211-1 provided that the conditions or indications for which the visual examination is performed can be detected at

Attachment I to WO 20-0029 Page 44 of 122 the chosen distance and illumination. Note that Table IWA-2211-1 includes the requirements for the VT-1 and VT-3 visual examinations.

WCGS Implementation:

This modification of Table IWA-2211-1 requirements for remote Subsection IWE visual examinations is incorporated in WCGS procedure IWE/IWL Visual Examination.

• 10 CFR 50.55a(b)(2)(ix)(J) requires a 10 CFR 50, Appendix J, Type A test be performed after any major containment modification or repair/replacement prior to plant start-up in order to ensure structural and leak-tight integrity. If a Type A, B, or C test is performed, when using IWE-5000, the test pressure and acceptance standards must be in accordance with 10 CFR 50, Appendix J.

WCGS Implementation:

If WCGS performs a major modification or repair/replacement of the containment structure, a 10 CFR 50, Appendix J, Type A test will be performed prior to start-up. If a Type A, B, or C Test is performed under IWE-5000, the test pressure and acceptance standards will be in accordance with 10 CFR 50, Appendix J.

• 10 CFR 50.55a(g)(6)(ii)(B), Augmented ISI requirements: Submitting containment ISI programs, states licensees do not have to submit to the NRC for approval of their containment inservice inspection program that was developed to satisfy the requirements of Subsection IWE and Subsection IWL with specified conditions. The program elements and the required documentation shall be maintained on site for audit.

WCGS Implementation:

The WCGS Containment Inservice Inspection Program and associated documents are maintained on site using the site document control processes. These documents are available for audit at any time.

3. License Renewal Requirements Applicable to Containment An operating license renewal application was submitted to the NRC for WCGS on September 27, 2006, pursuant to 10 CFR 54. The operating license renewal application proposed to extend the life of WCGS for an additional 20 years. A Safety Evaluation Report (SER) was issued by the NRC on November 20, 2008, and the renewed facility operating license was issued for WCGS. The WCGS period of extended operation is effective from March 12, 2025 through March 11, 2045.

(NOTE: Section 3.6 License Renewal Aging Management, of this LAR provides a full description of the License Renewal activities, commitments and tracking for containment items at WCGS.)

Attachment I to WO 20-0029 Page 45 of 122 Examination Frequency Frequency of examinations of ASME Class CC and MC components is specified in the WCGS CISI Program. The inspection program, as defined in IWE-2411, is utilized for Class MC components.

Concrete accessible surface areas shall be examined every 5 years in accordance with IWL-2410.

The unbonded post-tensioned tendon system shall be examined every 5 years in accordance with IWL-2420.

IWL-2410 allows for deferral of concrete visual examinations to the next regularly scheduled plant outage for portions of the concrete surface, which cannot be examined within the IWL-2410 stated time-frames. If such deferral is necessary, the visual examination may only be credited to the interval in effect at the time the deferral was necessary.

Responsibility Overall Responsibility The Manager Support Engineering is responsible for developing and maintaining the Containment Inservice Inspection (CISI) Program Plan, Wolf Creek Generating Station Interval 3.

Program Procedure Containment Inservice Inspection Program, is the administrative oversight procedure for the Program Plan that describes the overall development and control of the Containment Inspection Program Plan.

Responsible Individual The implementation of the WCGS Containment Inservice Inspection Program, defines the responsibility of the Responsible Individual as follows:

ASME Section XI, Subsection IWE, requires that a Responsible Individual be identified to provide oversight of the Class MC and metallic liner of Class CC examination program. The Responsible Individual shall be knowledgeable in the design, inservice inspection, and testing of Class MC and metallic liners of Class CC components. Additionally, the WCGS Program, describes the responsibilities of the Responsible Individual referencing Paragraph IWE-2320 (b) items (1) through (4), which include:

• development of plans and procedures for general visual examination of containment surfaces,

• instruction, training, and approval of general visual examination personnel,

• performance or direction of general visual examinations, and

• evaluation of general visual examination results and documentation.

Attachment I to WO 20-0029 Page 46 of 122 Responsible Engineer The implementation of the WCGS Containment Inservice Inspection Program, defines the responsibility of the Responsible Engineer as follows:

ASME Section XI, Subsection IWL, requires that a Responsible Engineer be identified to provide oversight of the concrete examination program. The Responsible Engineer shall be a registered professional engineer and shall have knowledge of the design and construction codes and other criteria used in design and construction of concrete containments in nuclear power plants. The Responsible Engineer shall also be experienced in evaluating the condition of structural concrete.

Additionally, the WCGS Program, describes the responsibilities of the Responsible Engineer referencing Paragraph IWL-2330 Items (a) through (f), which include:

• development of plans and procedures for examination of concrete surfaces,

• approval, instruction, and training of personnel performing general and detailed visual examination,

• evaluation of examination results,

• preparation or review of Repair/Replacement Plans and procedures,

• review of procedures for pressure tests following repair/replacement activities, and

• Submittal of a report to the Owner documenting results of examinations, repair/replacement activities, and pressure tests.

Examination Methods Examination methods used to satisfy Code examination requirements are listed for nonexempt Class CC components, and Class MC systems and components in the CISI Program Plan.

Personnel performing nondestructive examinations will be qualified in accordance with plant Containment Inservice Inspection Program. This Program specifies that personnel performing visual examinations meet the applicable requirements of ASME Section XI, paragraphs IWA- 2300, and 10 CFR 50.55a(b)(2)(xviii).

Additionally, the Program requires that personnel performing CISI examinations use approved NDE procedures. These examination procedures are also reviewed and approved by the Authorized Nuclear Inservice Inspector (ANII). Discrepancies identified during CISI Examinations are corrected using the Condition Report process controlled per the Corrective Action Program.

Visual Examinations Method for IWE and IWL Containment Inservice Inspection Program provides direction that for IWE/IWL General Visual examinations are to be performed in accordance with Section XI paragraphs IWE-2311 and IWL-

Attachment I to WO 20-0029 Page 47 of 122 2310(a) and that the Detailed Visual exams are to be performed in accordance with IWL-2310(b).

In addition, it provides direction that for VT-3 and VT-1 Visual Examinations, are to be performed in accordance with IWA-2213 and IWA-2211 respectively.

Plant procedure for IWE/IWL Visual Examination, provides detail instruction in performing General, Detailed, VT-3 and VT-1 examinations.

As previously discussed above in the Section for 10 CFR 50.55a, Conditions and Requirements, 10 CFR 50.55a(b)(2)(ix)(B) allows licensees, when performing remotely the visual examinations required by Subsection IWE, to extend the maximum direct examination distance specified in Table IWA-2211-1 and to decrease the minimum illumination requirements specified in Table IWA-2211-1 provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination. Note that Table IWA-2211-1 includes the requirements for the VT-1 and VT-3 visual examinations. This modification of Table IWA-2211-1 requirements for remote Subsection IWE visual examinations is incorporated in the WCGS procedure for IWE/IWL Visual Examination.

Qualification of Examiners Personnel performing IWE/IWL visual examinations according to the Plant Procedure for IWE/IWL Visual Examination, are qualified to a minimum Level II in accordance with procedure Qualification and Certification of Examination Personnel, meeting the requirements of ASME Section XI, 2013 edition.

Additionally, as described in the procedure for IWE/IWL Visual Examination:

For IWE examinations; general visual examinations are performed using VT-3 qualified personnel meeting ASME Section XI 2013 edition.

For IWL examinations; general visual examinations are performed using VT-3 qualified personnel meeting ASME Section XI, 2013 Edition, IWL-2320, and detailed examinations are performed using the VT-1 method and are required to be performed by VT-1 qualified personnel meeting ASME Section XI, 2013 Edition IWL-2320.

A summary of certification of personnel performing CISI examinations, certified in accordance with WCGS procedure controlling Qualification and Certification of Examination Personnel, is provided below:

The qualification program for personnel performing the visual examinations meet the applicable requirements of ASME Section XI, IWA-2300 and the applicable provisions of 10 CFR 50.55a(b)(2)(xviii).

Attachment I to WO 20-0029 Page 48 of 122 The qualification program for personnel performing augmented ultrasonic thickness (UT) examinations meet the applicable requirements of ASME Section XI, IWA-2300.

Personnel performing the general and detailed visual examinations are certified to a minimum Level II VT-3 and Level II VT-1 respectively. Personnel performing all other examinations are a minimum Level II in the respective method. Personnel performing the concrete general and detailed visual examinations meet the IWL-2320 qualifications.

Acceptance Criteria - IWE Acceptance of components subject to IWE examination requirements are in accordance with Article IWE-3000. When the examination of a component detects flaws or area of degradation that does not meet the acceptance standards of IWE-3500 for the respective examination category, that component is unacceptable for continued service until the condition is corrected by a repair/replacement activity or accepted by an engineering evaluation.

• A component containing flaws or areas of degradation is acceptable for continued service if the flaw or degradation are corrected by a repair/replacement activity or by corrective measures to the extent necessary to meet the acceptance standards.

• A component whose examination detects flaws or areas of degradation that do not meet the acceptance stands of IWE-3500, is acceptable for continued service without a repair/replacement activity if the CISI engineering evaluation indicates that the flaw or degradation is nonstructural in nature or has no unacceptable effect on the structural integrity of the containment.

• When flaws or areas of degradation are accepted by engineering evaluation, the area containing the flaw or degradation shall be reexamined in accordance with IWE-2420(b),

(c), and (d).

WCGS has developed inspection guidance and detailed acceptance criteria for IWE visual examinations.

Acceptance Criteria - IWL Acceptance of components subject to IWL examination is acceptable if the Responsible Engineer determines that there is no evidence of damage or degradation, corrosion protection medium leakage, or end cap deformation requiring further evaluation or performance of repair/replacement activities.

The Responsible Engineer shall approve all evaluations of indications identified during IWL examinations.

Attachment I to WO 20-0029 Page 49 of 122 WCGS has developed inspection guidance and detailed acceptance criteria for IWL visual examinations.

Acceptance Standards, Evaluations and Reports The 2013 Edition of ASME Section XI applies the criteria in IWE-3522 to Class MC pressure retaining components and to metallic shell and penetration liners of Class CC components. WCGS applied the ultrasonic examination criteria in IWE-3522 to both Class MC components and the metallic liners of Class CC components. The WCGS concrete containment was not designed as Class CC; however, WCGS applies the acceptance standards specified in IWE-3522 to the metallic liner of the concrete containment and to the acceptance standards for the nondestructive examinations specified in IWL-3000.

The following examination standards for visual examination of containment surfaces are defined in Program Procedure, "Containment Inservice Inspection Program":

Standards for Examination Category (E-A), Containment Surfaces:

• General Visual of Coated and Non-coated areas is acceptable if the Responsible Individual determines that there is no evidence of damage or degradation requiring further evaluation or performance of a repair/replacement activity.

• General Visual of Moisture Barriers that identifies wear, damage, erosion, tear, surface cracks, or other defects that permit intrusion of moisture against inaccessible areas of the pressure retaining surfaces of the metal containment liner shall be corrected. The corrective measures may be deferred until the next regularly schedule outage when an engineering evaluation is performed that demonstrates the requirements of IWE-3512.

Standards for Examination Category (E-C), Containment Surfaces Requiring Augmented Examination:

1. The following VT-1 Visual Examinations relevant conditions shall require correction or evaluation prior to continued service:

• pressure retaining component corrosion or erosion that exceeds 10% of the nominal wall thickness,

• loose, missing, cracked, or fractured parts,

• bolting or fastener relevant conditions listed in IWB-3517.1,

Attachment I to WO 20-0029 Page 50 of 122

• structural distortion or displacement of parts to the extent that the components function is impaired, and

• moisture barrier conditions that fail to meet the acceptance standards of IWE-3512.

2. Ultrasonic Examination of Class MC pressure retaining components that detect material loss in a local area exceeding 10% of the nominal wall thickness, or material loss in a local area projected to exceed 10% of the nominal wall thickness prior to the next examination, shall be documented. Such areas shall be accepted by engineering evaluation or repair/replacement activities. Supplemental examination in accordance with IWE-3200 is performed when specified as a result of the engineering evaluation.

Standards for Examination Category (E-G), Pressure Retaining Bolting:

Relevant conditions listed in IWB-3517.1 requires correction or evaluation to meet the requirements of IWE-3122 prior to return to service.

For Class MC components, when performing Engineering Evaluations to accept flaws or areas of degradation, if portions of later editions of the Construction Code or Section III are used, all related portions of that Code shall be met.

Concrete containment items with examination results that do not meet the acceptance standards of IWL-3000 shall be evaluated by Engineering. Engineering shall prepare an Engineering Evaluation Report stating:

• The cause of the condition that does not meet the acceptance standards,

• The acceptability of the concrete containment without repair to the item,

• Whether or not repair/replacement activity is required, and if required, the extent, method, and completion date for the repair/replacement activity, and

• The extent, nature, and frequency of additional examinations.

A Pressure Test Report is prepared following Repair/Replacement activities on the concrete containment. The report is prepared as detailed below:

A pressure test report is prepared under the direction of the Responsible Engineer. This report may be an addition to a previously prepared Engineering Evaluation Report (IWL-3310). The report shall:

• Describe the pressure test procedures,

• Summarize examination results, and

Attachment I to WO 20-0029 Page 51 of 122

• Shall state whether or not the repair/replacement activity is acceptable. If the repair/replacement activity is not acceptable, the report shall specify corrective measures.

Following each refueling outage, the CISI Coordinator includes relevant IWE/IWL examination information on a Form OAR-1 in accordance with the requirements of Code Case N-532-5, 10 CFR 50.55a(b)(2)(viii)(H), and 10 CFR 50.55a(b)(2)(ix)(A)(2).

Ultrasonic Thickness Measurements for Augmented Examinations IWE-2500(b) provides the requirements for examination of Augmented IWE components. Surface areas requiring augmented examination that are not accessible for visual examination on the side requiring augmented examination shall be examined for wall thinning using an ultrasonic thickness measurement method in accordance with Appendix I of ASME Section XI, IWE-2500(b)(2), IWE-2500(b)(3) and IWE-2500(b)(4).

The ultrasonic thickness examination method is used to conduct the examination of Item E4.12 (Surface Area Grid, Minimum Wall Thickness Locations) of Table IWE-2500-1.

Ultrasonic thickness examinations are performed in accordance with ASME Mandatory Appendix I, Ultrasonic Examinations.

When ultrasonic thickness (UT) measurements are performed, grids are used in accordance with the guidance found in IWE-2500. UT thickness measurements utilize WCGS procedure QCP 503, Ultrasonic Examination for Component Wall Thinning.

Inspection Intervals The original WCGS rulemaking for a CISI Program resulted in the first IWE/IWL 10-year interval beginning on September 10, 1998, and ended on September 9, 2008.

The second 10-year interval CISI Program began on September 10, 2008, and ended on September 9, 2018.

The third and now current 10-year interval CISI Program began on September 10, 2018, and is scheduled to end September 9, 2028.

The examinations and system pressure tests required by IWA and IWE will be completed during each inspection interval for the service lifetime of the plant. In accordance with IWA-2000, WCGS has elected to perform these examinations under the Inspection Program of IWE-2000 (IWE-2400) and IWL-2000 (IWL-2400).

The inspection program includes the following:

Attachment I to WO 20-0029 Page 52 of 122

• The examination required by IWA, IWE and IWL are completed during each inspection interval for the service lifetime of the plant.

• The frequency of IWE examined components is per the requirements of IWE-2400 and is detailed in Appendix B of the CISI Program Plan.

• The frequency of the IWL examinations is every 5 years, which includes concrete containment examinations in accordance with IWL-2410 and for IWL Unbonded Post Tensioning System examinations in accordance with IWL-2420 (Detailed in Appendix C of the CISI Program Plan). The plant is beyond ten years of commercial operation; thus, a five-year frequency of examination, plus or minus one year, is utilized for WCGS.

Reportability Summary Records and reports associated with Subsection IWE and IWL components will be performed in accordance with the Containment Inservice Inspection Program.

Following each refueling outage, the CISI Coordinator includes relevant IWE/IWL examination information on a Form OAR-1 in accordance with the requirements of Code Case N-532-5, 10 CFR 50.55a(b)(2)(viii)(H), and 10 CFR 50.55a(b)(2)(ix)(A)(2).

These requirements described in more detail include:

• Ref: ASME Section Code Case N-532-5, Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000,Section XI, Division 1, has been included in the CISI Program.

The WCGS Technical Requirements Manual (TRM), Technical Requirement 3.6.1 states (summarized), for Actions A2 and B2 are summarized as follows:

Report the condition in the ASME Form OAR-1 within 90 days from completion of the refueling outage for the operational cycle in which tendon surveillance testing performed or in which the containment degradation was found.

• Ref: 10 CFR 50.55a(b)(2)(viii)(H) requires that each inaccessible Class CC area identified for evaluation under IWL-2512(a), or identified as susceptible to deterioration under IWL-2512(b) must be evaluated and reported as specified in paragraph in 10 CFR 50.55a(b)(2)(viii)(E).

Attachment I to WO 20-0029 Page 53 of 122 WCGS Implementation:

If the conditions in 10 CFR 50.55a(b)( (2)(viii)(H) described above, are identified at WCGS, this requirement creates an additional scope of components [above the scope described in IWL-1000, which exempts inaccessible concrete surface areas from examination in IWL-1220(b)]. The components that meet the conditions of IWL-2512(a) or IWL-2512(b) will be added as, Examination Category L-A, Inaccessible Below-Grade Areas, Item Number L1.13, components.

Currently, no suspect areas have been identified for examination. During future inspections, if areas are determined to be suspect by the Responsible Engineer, these areas will be listed in Appendix C of the CISI Program Plan and a detailed visual examination will be performed in accordance with the guidance in IWL-2510.

If conditions are identified as described in 10 CFR 50.55a(b)(2)(viii)(H) above, then the requirements of 10 CFR 50.55a(b)(2)(viii)(E) are followed. (b)(2)(viii)(E) states that for Class CC applications, the applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or the result in degradation to such inaccessible areas and included in the ISI Summary Report required by IWA-6000. Information provided will include:

(ix)(A)(2)(i) - A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (ix)(A)(2)(ii) - An evaluation of each area, and the result of the evaluation; and, (ix)(A)(2)(iii) - A description of necessary corrective actions.

• Ref: 10 CFR 50.55a(b)(2)(ix)(A)(2) requires licensees to provide certain information in the ISI Summary Report regarding inaccessible Class MC areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.

WCGS Implementation:

If the conditions described are identified at WCGS, they will be included in the ISI Summary Report. Information provided will include:

(ix)(A)(2)(i) - A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (ix)(A)(2)(ii) - An evaluation of each area, and the result of the evaluation; and, (ix)(A)(2)(iii) - A description of necessary corrective actions.

Attachment I to WO 20-0029 Page 54 of 122 IWL examinations are typically not performed during refueling outages. Therefore, any report required by IWL will be included in the ISI Summary Report within 90 days from completion of the refueling outage for the operational cycle of the IWL examinations.

Repair/Replacement Activities/Procedures Repair/Replacement activities to components subject to the requirements of IWE and IWL are performed in accordance with plant procedure, ASME Section XI Repair/Replacement Program.

Components on which repair/replacement activities have been performed are to have a preservice examination performed in accordance with IWE-2200 or IWL-2200.

Repair/Replacement activities involving major modifications, such as replacing a large containment penetration, cutting a large construction opening or other similar modification, to the Class MC pressure-retaining components and the liner plate of the Class CC concrete containment, are to be followed by a Type A test in accordance with 10 CFR 50, Appendix J, Paragraph IV.A.

Leak test following minor repair/replacement activities are performed as required by IWE-5220.

IWE Examinations Examination of ASME Components The Containment ISI Plan includes components subject to examination per Subsection IWE of ASME Section XI (Class MC components and the metal liner of concrete containments).

The scoping and defining of responsibility for Subsection IWE components are performed in accordance with Article IWE-1000 and are contained in the CISI Program Plan.

IWE Components Subject to Examination The IWE nonexempt components and subject to examination are those which do not meet the exemption criteria of IWE-1220. These components have been scoped and are identified on the CISI Drawings listed in the CISI Program Plan.

The Subsection IWE components subject to examination include:

• Containment Floor Liner (Sump & Reactor Cavity Liner only)

• Containment Wall Liner

• Containment Dome Liner

• Containment Piping Penetrations, unless exempted per IWE-1220 (detailed below)

• Containment Electrical Penetrations

Attachment I to WO 20-0029 Page 55 of 122

• Equipment Hatch

• Personnel Access Hatch

• Auxiliary Personnel Access Hatch

• Fuel Transfer Tube

• Moisture Barriers (Concrete to liner interfaces and leak chase test connections)

IWE Components Subject to Exemption Per IWE-1220 the following components (or parts of components) are exempted from the examination requirements of IWE-2000:

• Vessels, parts, and appurtenances that are outside the boundaries of the containment system as defined in the Design Specification

• Embedded or inaccessible portions of containment vessels, parts, and appurtenances that met the requirements of the original Construction Code

• Portions of containment vessels, parts, and appurtenances that become embedded or inaccessible as a result of vessel repair/replacement activities if the conditions of IWE-1232(a) and IWE-1232(b) and IWE-5220 are met

• Piping, pumps, and valves that are part of the containment system, or which penetrate or are attached to the containment vessel IWE Components Subject to Augmentation Augmented Subsection IWE Components are those that meet the criteria in IWE-1241. These components are listed in the Program Plan as Examination Category (E-C) and are inspected in accordance with the guidance in IWE-1242. The components identified in IWE-1241(c) continue to require augmented inspection until the criteria in IWE-2420(d) are met.

When surface areas meet the criteria in IWE-1241(c), a separate component is created within the Program spreadsheet that will identify the specific area requiring augmented examination. The corresponding containment surface (E-A) component is maintained in the spreadsheet to track the overall surface examination requirements. The following Table 3.5.2-1, lists the Subsection IWE components that require augmented examination for the current period:

Attachment I to WO 20-0029 Page 56 of 122 Table 3.5.2-1 Containment Surfaces Requiring Augmented Examination Zones 2-3-1A and Zone 2-3-2A, Liner of both normal sumps Zone 2-3-1D Containment Liner of the Reactor Cavity sump 2-3-1C 2-3-1C-MB 2-3-2 Note:

For these components, only the locations identified will be in the augmented Examination Category E-C, and the rest of the component will remain in Examination Category E-A.

A note for each applicable component is provided in Appendix B of the Program Plan describing the specific location.

During future inspections, if other areas are identified which meet the criteria in IWE-1241, these areas will be: categorized as augmented examination areas, listed in the Containment Surfaces Requiring Augmented Examination table, listed in Appendix B of the Plan, and examined in accordance with IWE-2500(b).

Note that the liners of both normal sumps and the reactor cavity sump are listed in Table 3.5.2-1 as augmented examination surfaces as they meet the criteria in IWE-1241(a).

As described in IWE-2500(d), ASME Section XI requires the identification and evaluation of inaccessible Class MC areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.

Currently, no areas at WCGS meeting these criteria have been identified. If areas are determined to meet these criteria during future inspections, these areas will be listed in the Program Plan, and appropriate evaluations will be performed and documented.

Technical Specification Requirements The CISI program is described in the WCGS "Containment Inservice Inspection Program.". The IWE examination program and plan addresses the programmatic requirements of the following section of the WCGS Technical Specifications:

• TS 5.5.16.a.2, Containment Leakage Rate Testing Program

• TS 3.6.1, Containment Systems

• TSR 3.6.1.3, Containment Vessel Structural Integrity

Attachment I to WO 20-0029 Page 57 of 122 IWE Components Examination and Inspection Subsection IWE components scoped and required to be examined are listed in Appendix B of the CISI Program Plan. The examination and inspection of Subsection IWE components are performed in accordance with Articles IWE-2000 and IWA-2000, incorporating the clarifications of scheduling requirements, personnel qualifications and any augmented examination requirements.

The CISI Program Plan, implementing the requirements of IWE-2500, provides summary tables and examination methods to comply with the Tables in IWE-2500-1. Tables 3.5.2-2, 3.5.2-3 and 3.5.2-4 shown below, provide a summary of the IWE containment inspection program at WCGS and the examinations required per IWE-2411.

Attachment I to WO 20-0029 Page 58 of 122 Table 3.5.2-2 Category (E-A) Third Interval CISI Program Plan Extent & Freq Number of Relief Examinatio Item Description of Parts Exam Method of Component Request Notes n Category No. Examined Examination s No.

Containment Vessel 100% During Pressure Retaining General each E1.11 322 NA 1,5,6 Boundary; Accessible Visual inspection Surface Areas Period Containment Vessel Pressure Retaining E-A E1.12 Boundary; VT-3 100% - NA 1,2,7,8 Wetted Surfaces of Containmen Submerged Areas t Surfaces BWR Vent System E1.20 VT-3 100% - NA 3 Accessible Surface Areas Containment Vessel 100% During Pressure Retaining General each E1.30 29 NA 4,5,6 Boundary; Visual inspection Moisture Barriers period Notes:

(1) Examination shall include all accessible interior and exterior surfaces of Class MC components, parts, and appurtenances, and metallic shell and penetration liners of Class CC components. The following items shall be examined:

Attachment I to WO 20-0029 Page 59 of 122 (a) integral attachments and structures that are parts of reinforcing structure, such as stiffening rings, manhole frames, and reinforcement around openings.

(b) surfaces of attachment welds between structural attachments and the pressure retaining boundary or reinforcing structure, except for nonstructural or temporary attachments as defined in NE-4435 and minor permanent attachments as defined in CC-4543.4.

(c) surfaces of containment structural and pressure boundary welds, including longitudinal welds (Category A), circumferential welds (Category B), flange welds (Category C), and nozzle-to shell welds (Category D) as defined in NE-3351 for Class MC and CC-3840 for Class CC; and surfaces of Flued Head and Bellows Seal Circumferential Welds joined to the Penetration.

(d) pressure-retaining bolted connections, including bolts, studs, nuts, bushings, washers, and threads in base material and flange ligaments between fastener holes. Bolted connections need not be disassembled for performance of examinations.

(2) Examinations may be performed at any time during the interval, provided successive examinations are performed no less frequently than every third period.

(3) Includes flow channeling devices within containment vessels. Not applicable to WCGS.

(4) Examination shall include moisture barrier materials intended to prevent intrusion of moisture against inaccessible areas of the pressure retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces which are not seal-welded. Containment moisture barrier materials include caulking, flashing, and other sealants used for this application.

At WCGS, these include the accessible test connections of the leak chase channels, and the concrete to liner interfaces of the fuel transfer tube, incore tunnel and the 2000' elevation floor. Coating is considered to be a sealant. Reference CR 55822.

(5) General Visual examinations of surfaces and pressure retaining bolting will utilize the listings in the CISI Spreadsheet for IWE surfaces and bolting to implement the Table IWE-2500-1, Examination Category E-A, Note (1) requirement.

(6) ASME Section XI requires the General Visual examination for Item Nos. E1.11 and E1.30 to be performed once each period.

(7) Examinations of Item No. E1.12 components are deferrable to the end of the interval. At WCGS, all of the wetted surfaces of submerged areas (normal and incore sump) are in Examination Category E-C because they meet the criteria of IWE-1241(a).

(8) Number of components reflects start of the Third CISI Interval, First Period status, where Item No. E1.12 has 0 components.

Attachment I to WO 20-0029 Page 60 of 122 Table 3.5.2-3 Category (E-C) Third Interval CISI Program Plan Number of Relief Examinatio Item Description of Parts Exam Extent & Freq of Note Component Request n Category No. Examined Method Examination s s No.

E-C 100% of surface areas Containment Surface Containmen identified by IWE-1242 E4.11 Areas; VT-1 5 NA 1 t Surfaces during each inspection Visible Surfaces Requiring period Augmented Examination 100% of minimum wall Containment Surface Ultrasoni thickness locations during Areas; c each inspection period, E1.12 Surface Area Grid, 12 NA 1 Thicknes established in accordance Minimum Wall Thickness s with IWE-2500(b)(3) and Locations (b)(4)

Notes:

(1) Containment surface areas requiring augmented examination are those identified in IWE-1240.

(2) For Component ID, 2-3-2, Ref Work Order 16-418116-001 tracks the requirement to perform an Ultrasonic Thickness examination of bulges at Azimuth 69 and 72 at 2018 elevation (near Emergency Escape Hatch).

Attachment I to WO 20-0029 Page 61 of 122 Table 3.5.2-4 Category (E-G) Third Interval CISI Program Plan Number of Relief Examination Item Description of Parts Exam Extent & Freq of Component Request Notes Category No. Examined Method Examination s No.

E-G 100% of each bolted Pressure Retaining E8.10 Bolted Connections VT-1 71 NA 1,2 connection Bolting Notes:

(1) Examination shall include bolts, studs, nuts, bushings, washers, and threads in base material and flange ligaments between fastener holes.

(2) Examination may be performed with the connection assembled and bolting in place under tension, provided the connection is not disassembled during the interval. If the bolted connection is disassembled for any reason during the interval, the examination shall be performed with the connection disassembled. (Note that the identifiers in the CISI Spreadsheet for bolted connections (typically -B) are being utilized by WCGS.)

Attachment I to WO 20-0029 Page 62 of 122 IWE Component Examination Schedule The schedule for the IWE examinations at WCGS for Category (E-A), (E-C) and (E-G) components, is contained in the CISI Program Plan, Appendix B. The Appendix B schedule is detailed and componentized by outage and period within the current 10-year interval. This IWE examination schedule is summarized in Table 3.5.2-5 shown below.

Scheduling of individual components per the requirements of IWE-2400 is the responsibility of the CISI Coordinator. The CISI Coordinator ensures that the Responsible Individual is informed of the scheduled examinations.

Table 3.5.2-5 IWE Examination Schedule Interval Period Unit 1 Start Date to Start Date to End Projected Outage Outage Number End Date Date Start Date (Interval and Period)

Scheduled RF23 1st Period Fall 2019 (3-1-1) 9/10/18 to 9/9/211,2 Scheduled RF24 Spring 2021 (3-1-2)

Scheduled RF25 3rd 10-year 2nd Period Fall 2022 (3-2-1)

Interval 9/10/21 to 9/9/251,2,3 Scheduled RF26 9/10/18 to 9/9/28 Spring 2024 (3-2-2)

Scheduled RF27 3rd Period Fall 2025 (3-3-1) 9/10/25 to 9/9/281,2,3,4 Scheduled RF28 Spring 2027 (3-3-2)

Scheduled RF29 1st Period Fall 2029 (4-1-1) 9/10/28 to 9/9/31 Scheduled RF30 Spring 2031 (4-1-2)

Scheduled RF31 4th 10-year Interval 5 2nd Period Fall 2032 (4-2-1) 9/10/31 to 9/9/35 Scheduled RF32 9/10/28 to 9/9/38 Spring 2034 (4-2-2)

Scheduled RF33 3rd Period Fall 2035 (4-3-1) 9/10/35 to 9/9/38 Scheduled RF34 Spring 2037 (4-3-2)

Notes

Attachment I to WO 20-0029 Page 63 of 122 (1) IWA-2430(c)(1) allows the inspection interval to be extended by as much as one year and may be reduced without restriction. IWA-2430(c)(3) allows each period to be extended by as much as one year and may be reduced without restriction to enable an inspection to coincide with a plant outage, within the limitations of IWA-2430(c). IWA-2430(d) allows the inspection interval to be extended when a unit is out of service continuously for six months or more, provided the limitations of IWA-2430(d) are met.

(2) IWE-2411(b)(1) requires when items or welds are added to the inspection program during the first period of an interval, the examinations required by the applicable Examination Category and Item Number shall be performed during each of the second and third periods of that interval. Alternately, if deferral of the examinations is permitted for the Examination Category and Item Number, the required examinations shall be performed during either the second or third period of the interval.

(3) IWE-2411(b)(2) requires when items are added to the inspection program during the second period of an interval, the examinations required by the applicable Examination Category and Item Number shall be performed during the third period of that interval.

(4) IWE-2411(b)(3) requires when items are added to the inspection program during the third period of an interval, examinations shall be scheduled in accordance with IWE-2411(a) for successive intervals.

(5) The schedule for the 4th 10-year interval is proposed as the 4th interval plan has yet to be developed.

IWL Examinations Examination of ASME Components The Containment ISI Plan includes components subject to examination per ASME Section XI Subsection IWL, Requirements for Class CC Concrete Components.

The scoping and defining of responsibility for Subsection IWL components are performed in accordance with Article IWL-1000 and are contained in the Program Plan.

IWL Components Subject to Examination The IWL nonexempt components are those which do not meet the exemption criteria of IWL-1220.

These components have been scoped and identified on the CISI drawings listed in the IWL Program Plan.

The Subsection IWL components subject to examination include the following:

• Concrete Containment Structure (includes the tendon gallery ceiling)

Attachment I to WO 20-0029 Page 64 of 122

• Concrete Dome

• Post Tensioning System IWL Components Subject to Exemption In accordance with IWL-1220, the following items are exempted from the examination requirements of IWL-2000:

• Tendon end anchorages that are inaccessible, subject to the requirements of IWL-2521.1

• Portions of the concrete surface that are covered by liner

• Portions of the concrete surface obstructed by adjacent structures, components, parts, or appurtenances, unless the Responsible Engineer determines that examination is required as a result of conditions identified in accessible areas

• Portions of the concrete surface made inaccessible by foundation material or backfill, subject to the provisions of IWL-2512 Concrete Suspect Areas Concrete suspect areas are those that meet the criteria of IWL-2510 and are subject to the examination requirements of Table IWL-2500-1, Examination Category L-A, Item Number L1.12.

These suspect areas shall be identified by the Responsible Engineer and will require a detailed visual examination to be scheduled. These potential concrete suspect areas are not intended to identify all suspect areas identified during scheduled general visual examinations.

Currently, no suspect areas have been identified for examination at WCGS. During future inspections, if areas are determined to be suspect by the Responsible Engineer, these areas will be listed in Appendix C of the IWL Program Plan and a detailed visual examination will be performed in accordance with the guidance in IWL-2510.

As described in IWL-2512, ASME Section XI requires the Responsible Engineer to evaluate suspect conditions and specify the type and extent of examinations, for the inaccessible Class CC surface areas exempted by IWL-1220(c) and IWL-1220(d).

Concrete surfaces exposed to foundation soil, backfill, or ground water are inaccessible areas and shall be evaluated to determine susceptibility of the concrete to deterioration and the ability to perform the intended design function under conditions anticipated until the structure no longer is required to fulfill its intended design function.

The technical evaluation shall be performed and documented by or under the direction of the Responsible Engineer, at periodic intervals not to exceed 10 years. 10 CFR 50.55a(b)(2)(viii)(I)

Attachment I to WO 20-0029 Page 65 of 122 requires the evaluation be performed at a frequency not to exceed 5 years once the licensee has entered the period of extended operation. It also requires that licensees must examine representative samples of the exposed portions of the below-grade concrete, when such below-grade concrete is excavated for any reason.

(Note: WCGS will enter the period of extended operation under License Renewal on March 12, 2025; therefore, 10 CFR 50.55a(b)(2)(viii)(I) will be applied to the CISI Program at that time.)

Currently, a single component item has been identified in this CISI Program Plan for tracking purposes until the inaccessible areas have been evaluated by the Responsible Engineer and determined the type and extent of examinations required to be performed on inaccessible areas.

(Ref: Table IWL-2500-1, Examination Category L-A, Item Number L1.13)

If areas are determined to meet these criteria during future inspections, these areas will be listed in the IWL Program Plan, and appropriate evaluations will be performed, and then documented accordingly.

Technical Specification Requirements The containment tendon surveillance program is defined in WCGS program procedure, Containment Inservice Inspection Program. This IWL testing/surveillance program and plan addresses the programmatic requirements of the following section of the WCGS Technical Specifications:

• TS 5.5.6, Containment Tendon Surveillance Program

• TS 5.5.16.a.1, Containment Leakage Rate Testing Program

• TS 3.6.1, Containment Systems

• TR 3.6.1, Containment Vessel Structural Integrity IWL Components Examination and Inspection The examination and inspection of Subsection IWL components are performed in accordance with Article IWL-2000 and IWA-2000. IWL components scoped and required to be examined are listed in Appendix C of the CISI Program Plan.

The IWL Program Plan summary tables implement the requirements of IWL-2500 which specifies examination methods for components to comply with the Tables in IWL-2500-1 Category (L-A) and IWL-2500-1 Category (L-B). Tables 3.5.2-6, and 3.5.2-7 shown below, provide a summary of the containment inspection program at WCGS and the examinations required.

Attachment I to WO 20-0029 Page 66 of 122 Table 3.5.2-6 Category (L-A) WCGS Third Interval CISI Program Plan Number of Examination Item Description of Parts Exam Extent & Freq of Relief Component Notes Category No. Examined Method Examination Request No.

s Concrete Surface; IWL-2510 and General L1.11 All Accessible Surface 6 NA 1 Visual IWL-2410 Areas L-A Concrete Surface; Detailed IWL-2510 and L1.12 0 NA -

Concrete Suspect Areas Visual IWL-2410 Concrete Surface; IWL-2512(a) and IWL-L1.13 Inaccessible Below- 1* NA 2,3 2512(c) IWL-2512(c)

Grade Areas NOTES:

(1) Includes concrete surfaces at tendon anchorage areas not selected by IWL-2521 or exempted by IWL-1220(a). The number of components includes six surfaces: the containment wall (Zones A, B, and C), containment dome, tendon gallery ceiling concrete, and tendon anchorage areas.

(2) Concrete surfaces (exposed to foundation soil, backfill, or ground water) susceptible to deterioration as defined in IWL-2512, shall be evaluated for structural integrity.

• Currently, no suspect areas have been identified for examination. If future inspections identify areas, they will be shown in Item L1.13. (Refer CISI Program Plan, Section 4.2.1.3)

(3) Method of examination as defined by the Responsible Engineer, based on IWL-2512(b).

Attachment I to WO 20-0029 Page 67 of 122 Table 3.5.2-7 Category (L-B) WCGS Third Interval CISI Program Plan Number of Relief Examination Item Description of Parts Exam Extent & Freq of Note Component Request Category No. Examined Method Examination s s No.

IWL-2521 and Tendon IWL-2522 L2.10 IWL-2420 251 NA -

IWL-2523.1 and L2.20 Wire or Strand IWL-2523.2 IWL-2420 251 NA -

L-B Anchorage Hardware Unbonded and Detailed IWL-2524.1 and L2.30 Post- Visual Surrounding Concrete IWL-2420 251 NA -

Tensioning System IWL-2525.1(a),

IWL-Corrosion Protection L2.40 2525.2(a), IWL-2526, and 251 NA Medium IWL-2526 IWL-2420 -

IWL- IWL-2525.1(b) and Free Water L2.50 2525.2(b) IWL-2420 251 NA -

Attachment I to WO 20-0029 Page 68 of 122 IWL Component Examination Schedule Implementation of the Subsection IWL Program, schedule is driven by:

• IWL-2400, Inservice Inspection Schedule, which requires examinations to be performed at 1, 3, and 5 years following the completion of the containment Structural Integrity Test, and every 5 years thereafter per IWL-2410 for concrete and IWL-2420 for Unbonded Post-Tensioning Systems.

• IWL-2410(c) and IWL-2420(c) require the 10-year and subsequent examinations of concrete to commence not more than 1 year prior to the specified dates and completed not more than 1 year after such dates.

Scheduling of individual components per the requirements of IWL-2410, for Concrete, and IWL-2420, for the Post Tensioning System, is the responsibility of the CISI Coordinator. The CISI Coordinator ensures that the Responsible Engineer and Responsible Individual are informed of the scheduled examinations.

The Structural Integrity Test (SIT) for WCGS was performed on December 18, 1984. This date is used as the IWL reference date for the IWL tendon examinations.

The date of the initial IWL concrete examination was May 23, 2000 and is used as the IWL reference date for the IWL concrete examinations. The plant is beyond ten years of commercial operations; thus, a five-year frequency of examinations, plus or minus one year, is utilized for WCGS.

The schedule for the IWL examinations at WCGS for concrete Category (L-A) and tendons Category (L-B) is contained in the CISI Program Plan, Appendix C. The Appendix C schedule is detailed and componentized by outage and period within the current 10-year interval. This IWL examination schedule is summarized in Table 3.5.2-8 shown below.

Attachment I to WO 20-0029 Page 69 of 122 Table 3.5.2-8 IWL EXAMINATION SCHEDULE Interval 5-Year Period 5-Year Period Start Date to Rolling Exam # - Date Projected Outage Outage End Date (2 Year Window) Start Date Number 35th Year Containment Testing Scheduled RF23 Tendon - 12/18/2019 Fall 2019 Concrete - 5/23/2020 Scheduled No Section XI Exams RF24 Spring 2021 Scheduled No Section XI Exams RF25 3rd 10-year Interval Fall 2022 9/10/18 to 9/9/28 Scheduled No Section XI Exams RF26 Spring 2024 40th Year Containment Testing Scheduled RF27 Tendon - 12/18/2024 Fall 2025 Concrete - 5/23/2025 Scheduled No Section XI Exams RF28 Spring 2027 45th Year Containment Testing Scheduled RF29 Tendon - 12/18/2029 Fall 2029 Concrete - 5/23/2030 Scheduled 4th 10-year Interval 1 No Section XI Exams RF30 Spring 2031 9/10/28 to 9/9/38 Scheduled No Section XI Exams RF31 Fall 2032 Scheduled No Section XI Exams RF32 Spring 2034 50th Year Containment Scheduled Testing RF33 Fall 2035

Attachment I to WO 20-0029 Page 70 of 122 Tendon - 12/18/2034 Concrete - 5/23/2035 Scheduled No Section XI Exams RF34 Spring 2037 (1) The schedule for the 4th 10-year interval is proposed as the 4th interval plan has yet to be developed.

Requests for Relief Relief may be requested per 10 CFR 50.55a for situations where alternatives to ASME Section XI requirements provide an acceptable level of quality and safety; for situations where compliance with ASME Section XI requirements results in a hardship or an unusual difficulty without a compensating increase in the level of quality and safety; and for situations where ASME Section XI requirements are considered impractical.

Currently, there are no Requests for Relief or Requests for Alternatives included or submitted in the third interval CISI Program at WCGS.

ASME Section XI Code Cases In accordance with 10 CFR 50.55a(b), ASME Section XI Code Cases referenced in RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, (Reference 35), may be incorporated into the WCGS ISI Program.

The alternative requirements included in the following ASME Section XI Code Case will be utilized:

• N-532-5, Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000,Section XI, Division 1.

Note that in the WCGS TRM, Technical Requirement 3.6.1 includes Required Actions A.2 and B.2 to report abnormal containment vessel degradation in ASME Form OAR-1.

3.5.3 Supplemental Inspection Requirements Supplemental Inspections will not be required. Inspections of the exterior containment concrete surfaces and the steel liner plate inside containment will be conducted in accordance with TS 5.5.16 (as modified by TS Amendment 152 (Reference 15)) as follows:

Attachment I to WO 20-0029 Page 71 of 122

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief or alternative has been authorized by the NRC.

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by the ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

The bases of acceptability for continued utilization of the examination requirements stated above is contained in Section 3.3.4 of this LAR.

3.5.4 Results of Recent Containment Examinations Results of Recent Coating Inspections Each refueling outage a general walkdown inspection of the containment coating systems is performed along with a detailed visual inspection of approximately one third of the containment coating system area. A review of previous inspections is conducted to determine the areas for subsequent outage inspections. Inspections are performed for signs of peeling, cracking, blistering, or delamination.

The most recent containment coating examinations were completed during RF23 in October 2019.

A summary report was prepared for each examination performed and is summarized below.

RF23 Coating Performance Monitoring Report As reported in past assessment reports, several observations were made of coatings that had suffered mechanical damage due to contact with traffic in containment. Other coating came in contact with chemicals that are being used in the Containment for either operational or maintenance purposes. The mechanical wear and staining is considered normal wear and the coatings remain able to perform during a design basis accident; (i.e. no effect on Emergency Sump (ECCS) performance as long as the coating is observed to be soundly adhered to the substrate). No softening or erosion of coating observed. Loose coatings that were identified under WOs shown were scraped putting the coating into stress which provides reasonable assurance that the coating will not impact ECCS sump performance.

Attachment I to WO 20-0029 Page 72 of 122 Incore Tunnel Sump area showed damaged coatings with some small area of delaminated coatings found on both horizontal and vertical surfaces. The delaminated coatings were disposed of per plant procedures and associated WOs. This area is considered inactive for DBA; therefore, no impact on ECCS sump performance will occur.

Overall, the coatings inspected were observed to be soundly adhered to the substrate with no signs of delamination (peeling, cracking, blistering, etc.). Localized areas that showed signs of coating delamination and could, potentially, impact the ECCS were noted and addressed in this report.

Therefore, inspected coatings were judged to have no effect on the ECCS performance. No mode hold was required.

Detailed conditions that are noted in the RF23 summary report above, are presented below for the following areas:

Evaluation of Coating Inspection Polar Crane and Rail:

The underside of the crane and rail were inspected using remote visual methods (remote cameras).

While the underside of the crane exhibited areas of light rust, no such areas were identified on the rail. No detrimental issues that could be indicative of coating not soundly adhered to the substrate were identified. No visible changes from pictures taken in the previous report (RF22) could be identified.

Maintenance Truss:

The coating for the Maintenance Truss is deemed to be non-qualified and amount is being carried in the Coating Log as well as in the USAR as non-qualified.

The following questions were asked and responses provide by the QC inspector:

1) Is there any loose flakes in catch netting underneath the truss?

"I did not notice any flakes in the screens beneath the truss, everything seems to be fairly clean. I took the putty knife because we have looked at these coatings so many times it is hard to remember if the problem is getting worse from the pictures.

Some of the coatings appear that they are about to flake off, but when scraped they are tightly adhered."

2) Is there a coating underneath the yellow coating that you were trying to remove?

"As for the primer coat beneath the yellow and tan paint I did not see any of the primer coats beginning to flake, just a few scratches where flaking has previously been removed."

Attachment I to WO 20-0029 Page 73 of 122 Based on responses from the QC department, it can be concluded that the Maintenance Truss coating does not have loose coating capable of interfering with ECCS performance. Conditions were said to be exactly the same as before.

Containment Coating Inspection Inside and Outside Bio-Shield Walls:

This inspection included all coatings in accessible area on elevations 2000, 2026, 2047, and 2068.

All conditions observed were judged not to have any impact on the ECCS performance.

Coating inspection per IWE Program Coating Inspection was performed under WOs and associated sub-work orders. Based on review of associated WOs, cracking, flaking, peeling, blistering or bulging of the coating were either not noted or repaired.

Incore Tunnel Sump QC performed visual inspection of coatings on the Incore Tunnel. The inspection identified discoloration, peeling, and flaking and condition reports (CRs) were initiated and work orders (WOs) were written to address the conditions.

Degradation was determined to be as a result of:

1) Prolonged mechanical damage on the coating.

2) Improper surface preparation during RF22.

3) Localized adhesion/cohesion failure of the zinc primer.

Since a majority of the coating system on the liner plate were intact with no visible signs of degradation (as surveyed and accepted per VT-1 and QC inspection procedures), the coating system is judged to be effective in performing its design function. Additionally, based on walkdowns and pictures it was determined that the steel on the vertical surfaces and liner plate have a zinc-based coating that exhibits good adhesion and remains capable of protecting against external stressors as long as it is not continuously in contact with liquids. Since the majority of the coating system on the liner plate is intact with no visible signs of degradation, the identified defects are considered to be localized. The condition of the Incore Tunnel surfaces is to be monitored in future outages under established programs.

Additionally, the liner plate degradation in the Incore Tunnel was evaluated in this report to allow the plant to operate with the uncoated surface for 1 cycle (until RF24). WOs for removing loose coatings in the Incore Tunnel and associated VT-1 examinations were completed in RF23. All work was completed as planned and there were no indications which did not meet the acceptance criteria per QC inspection criteria. Information was collected from the coatings removed and provided to Coating Engineer for evaluation to help with future repair plans. A CR was generated to document an area on the liner plate which is the primary safety barrier showing signs of general surface

Attachment I to WO 20-0029 Page 74 of 122 corrosion that shall be re-coated during RF24. This area is located just above the moisture barrier close to the in-core sump.

Coating Failure Downstream of Penetration P-83/P-84.

A CR was written to document the cracking of the Carboline 890N epoxy coating downstream and south of penetration P-83/P-84 on a 3/4 pipe (BM61-DBB-3/4) going through the 18 pipe penetration of the liner plate. This pipe is designed as ASME Class 2 piping with a design temperature of 600

°F and a normal operating temperature of 560 °F. The pipe is Schedule 160 and is made of carbon steel. This condition appears to have happened due to excessively thick coating applied during touch-up work on penetrations P-83 and P-84, combined with piping design and operating temperatures exceeding maximum allowable Carboline 890N temperature of 340°F. Additionally, BM61-DBB-3/4 pipe was not intended to receive the coating and, hence, never received a proper surface preparation for coating adhesion. Therefore, the Carboline 890N coating is cracking.

Based on the CR, there are 2 pipes that include the cracked coatings. The total length of pipe exhibiting cracked coatings does not exceed 5 inches. The coating on pipes downstream from penetrations P-83 and P-84 is judged to a significantly small amount, with little or no influence on the margin of the calculation. Thus, leaving this in place until RF24 is acceptable, at which time it will be removed.

Containment Air Cooler C QC identified degradation during coating work on Containment Air Cooler C (asset # SGN01C).

Based on the CR, no more than 5 ft2 of unqualified coating was detected by QC during inspection.

Due to installation activities taking place in the area by other groups, it was determined that no effective coating contaminates traps could be installed and all activities needed to stop to allow the rework to begin. In addition, due to outage timing constraints, Outage Control Center rejected approving 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for coating preparation and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> cure time in addition to succeeding all work activities on the Containment Air Coolers to allow painters to paint. Therefore, the margin in the Unqualified Coating Log was updated through the issuance of a Calculation Change Notice to reduce allowable margin by 5 ft2 temporarily. The CR was converted into a WO to complete rework in a future outage.

Elevation 2068 Grating Support Beams:

A CR was written on coating degradation on the bottom flange of the elevation 2068 floor support beam around B containment cooler (asset # SGN01B). Based on the chips collected from the area, the failure mode is determined to be cohesion failure within the zinc primer possibly caused by application of a second coat of primer. Zinc dust could be observed both on the peeled off chips and on the remaining Carbozinc CZ11 that stayed intact on the beam. No dry film thickness (DFT) could be determined. Carbozinc CZ11 is installed as requirement per specification and overcoating CZ11 with Carboguard 890N is approved. It should be said that there is no short-term concern due to how long the coating stayed intact, but over time this condition has been noted to be a localized concern based on plant and industry operating experience (OE). The CR was converted into a WO, which was statused Field Work Complete the same day, and removed chips were provided to the Coatings Engineer for examination.

Attachment I to WO 20-0029 Page 75 of 122 Additionally, a CR was written to document fracking coating above pipe support GN01-R014. The failure mode is similar to other areas where Carbozinc CZ11 is overcoated with Carboguard 890N.

The CR was converted to a WO. Outage Review Board (ORB) met and decided to work the WO in RF24. Therefore, a Calculation Change Notice was issued to update the available unqualified epoxy coating in containment.

Containment Recirculation Sump Strainer (FEN01A) Basin:

During containment walkdown on September 21, 2019, the SRI identified cracking in the coating of FEN01A. The coating appears to be attached in all locations. The cracking was identified just inside of the berm that goes around the sump but outside of the screens. The coating was evaluated by the Coatings Engineer to determine if further action is needed. It was determined that the coating cracking does penetrate though the entire thickness of the layer and is due to minor concrete surface imperfections. It should be noted that when the coating was put into stress, using force with a dull putty knife for adhesion, and could not be removed. Therefore, this condition is judged to be of cosmetic nature and does not prevent the concrete or Containment Recirculation Sump Strainer from performing its safety related function.

Containment Normal Sump Pump Pit (PLF05C):

During RF22 degradation was identified during IWE inspection inside Containment Normal Sump Pump Pit (PLF05C) on level indicator support plate. The condition was repaired. While in RF23, the coated area was identified as degraded again and rework was requested. At the time of completion of this report the WO was statused as Supervisor Approved. Per the USAR, normal sumps are coated with non-qualified coating and, therefore, accounted for. Hence, removed coating has no impact on Containment Recirculation Sumps ability to perform its safety related function.

A CR was generated for an in-depth look at the failure mechanism for long term investigation and resolution of zinc adhesion/cohesion failure.

Results of Recent IWE Examinations RF22 Refueling Outage IWE examinations, following the schedule in the 2nd 10-year Interval of the WCGS CISI Program, were conducted in RF 22 that comply with the Containment Structure Surface Inspection. This procedure is used to satisfy the Containment Leakage Rate Testing Programs containment surface visual examination in accordance with Technical Specification SR 3.6.1.1 and TS 5.5.16.

The inspection was implemented through six (6) WOs in accordance with IWE/IWL Visual Examination. The RF22 examinations were the last exams conducted in the 3rd period, of the 2nd 10-year Interval of the WCGS CISI Program.

Attachment I to WO 20-0029 Page 76 of 122 Overall, the RF22 IWE examinations were completed satisfactorily with some areas requiring follow-up actions that are described below. In conclusion, the IWE examinations performed have shown that the containment pressure boundary continues to perform its intended function as a leak tight barrier.

The six (6) WOs written for the RF22 IWE examination results are summarized below.

(1) General and Detailed Visual Examination of liner surfaces Category (E-A): General Visual Examination - Item No. E1.11 - One IWE general visual was performed.

Category (E-C), Augmented Visual Examination - Item No. E4.11 - Three detailed VT-1 examinations were performed.

A summary of the three (3) areas (four (4) examinations total) this WO examined are:

(1) Category (E-A): General Visual Examination - (ID: 2-3-1C) - Containment Metallic Liner surface of reactor cavity.

(2) Category (E-C): Augmented VT-1 Examination - (ID: 2-3-1C) - Containment Metallic Liner surface of reactor cavity (& follow-up with VT-1).

(3) Category (E-C): Augmented VT-1 Examination - (ID: 2-3-1D) - Examination of Containment Metallic Liner surface of reactor cavity sump, PLF07A/B.

(4) Category (E-C): Augmented VT-1 Examination - (ID: 2-3-1-C-MB) - Examination of Concrete Floor to Liner interface of surface zone.

Inspection Results (in summary):

Areas (1), (2) and (4) above (ID: 2-3-1C and ID: 2-3-1-C-MB): Examination identified multiple locations of surface discoloration (Light red tightly adhering corrosion) or mechanical damage or coating damage of the liner coating. CR was written.

Disposition:

(2-3-1C) Noted degradation of component does not affect the structural integrity of the containment pressure boundary. The noted damage does not reflect accelerated corrosion that resulted in substantial corrosion or excessive wear. This component will be identified as Examination Category (E-A) and will not be added to the augmented inspections category (E-C). (Follow-up: Additional inspection -

this component was added/maintained within the augmented inspections Category (E-C) Item No E4.11 for RF23, RF25 and RF27.)

Attachment I to WO 20-0029 Page 77 of 122 (2-3-1-C-MB) Noted degradation of component conditions were adjacent to the moisture barrier and the moisture barrier remained tightly adhered to the liner. The noted conditions do not affect the structural integrity of the containment pressure boundary. The Reactor Cavity Moisture Barrier (2-3-1-C-MB) will remain on augmented inspection category (E-C). The Reactor Cavity Metallic Containment Liner coating damage noted shall be repaired.

Area (3) above (ID: 2-3-1D): VT-1 examination performed satisfactorily finding acceptable conditions with no indications.

(2) General Visual and Detailed Examination of liner surfaces.

Category (E-A): General Visual Examination - Item No. E1.11 - Five IWE general visual and one VT-1 examination was performed.

Category (E-C): Augmented Visual Examination - Item No. E4.11 - Three VT-1 examinations were performed.

A summary of the seven (7) areas (eight (8) examinations total) this WO examined are:

(1) Category (E-A): General Visual Examination - (ID: Dome-L-2-01) - Reactor Building Dome Containment Liner surface above spring line.

(2) Category (E-C): Augmented VT-1 Examination - (ID: Dome-L-2-01) - Reactor Building Dome Containment Liner surface above spring line.

(3) Category (E-A): General Visual Examination - (ID: 2-6-1) - Containment Metallic Liner surface Zone 2-6-1.

(4) Category (E-A): General Visual Examination - (ID: 2-6-2) - Containment Metallic Liner surface Zone 2-6-2.

(5) Category (E-A): General Visual Examination - (ID: 2-3-1B-A) - Containment Metallic Liner surface of emergency pump sump A-train.

(6) Category (E-A): General Visual Examination - (ID: 2-3-1B-B) - Containment Metallic Liner surface of emergency pump sump B-train.

(7) Category (E-C): Augmented VT-1 Examination - (ID: 2-3-1A) - Containment El 2000, Examination of Containment Metallic Liner surface of sump @ 270 deg.

(8) Category (E-C): Augmented VT-1 Examination - (ID: 2-3-2A) - Containment El 2000, Examination of Containment Metallic Liner surface of sump @ 90 deg.

Attachment I to WO 20-0029 Page 78 of 122 Inspection Results (in summary):

Areas (1) and (2) above (ID: Dome L-2-01) - Examination consisted of a general and a detailed VT-1 examination that identified 5 locations that had coating liner damage and/or light red tightly adhering corrosion bloom. CR was written.

Disposition: The noted degradation of (DOME-L-2-01) does not affect the structural integrity of the containment pressure boundary. The liner coating damage shall be repaired. The damage does not reflect accelerated corrosion that resulted in substantial corrosion or excessive wear.

Areas (3), (4), (5), and (6) above (ID: 2-6-1, ID: 2-6-2, ID: 2-3-1B-A and ID: 2-3-1B-B) - General Visual Examination of containment liner and the emergency sumps Trains A and B found acceptable conditions with no indications.

Area (7) above (ID: 2-3-1A) - VT-1 Examination of Containment Metallic Liner surface of sump at 270 deg identified a half inch diameter area of lost coatings with discoloration consistent with corrosion. It was located under the level indicator bracket to liner plate interface. Later information provided that the depth corrosion pit was approximately 1/64 (0.0156).

Disposition: Engineering has previously provided an acceptable minimum pit/corrosion spot depth of the liner plate in the normal sumps with:

Any pit with a depth less than 0.150 is acceptable and need not be repaired.

The minimum required thickness of the liner plate is 0.100. Therefore, the pit identified is within acceptance criteria with the identified pit depth of 0.0156and no repairs will be performed. Sump inspection will be required in RF23 to ensure that the pit/corrosion spot has not exceeded the acceptance criteria. The coating (Bio Dur 561) shall also be repaired in RF23. CR was generated to repair the coating in RF23. IWE inspection of the normal sumps will be performed in RF23 and will be part of the program inspection scope.

Area (8) above (ID: 2-3-2A) - VT-1 Examination of Containment Metallic Liner surface of sump at 90 deg found acceptable conditions with no indications.

(3) General Visual Examination of Auxiliary Building penetration surfaces and bolting.

Category (E-A): General Visual Examination - Item No. E1.11 - A total of 217 IWE General Visual Examinations were performed. A summary of the 217 areas this WO examined are:

Attachment I to WO 20-0029 Page 79 of 122 Auxiliary Building Outside Containment - Examinations are Program defined for the electrical penetration surfaces, electrical penetration bolting and of piping penetration surfaces at elevations 2028, 2029, 2033, 2037 and 2041 (Zones 1409, 1410, 1411, 1412, 1419) of the Auxiliary building.

Auxiliary Building Outside Containment - Examinations are Program defined for piping penetration surfaces at elevations 2005, 2009 and 2013, (Zones 1322 and 1323), at elevation 2053 (Zones 3 and 4) and at elevation 2029, (Zones 1409 and 1410) of the Auxiliary building.

Reactor Building Inside Containment - Examinations are Program defined for piping penetration surfaces accessible from one side only, in containment below elevation 2000 outside A/D loop and containment elevation 2005.

Inspection Results:

General Visual Examination of all areas defined in this WO passed satisfactorily finding acceptable conditions with no indications. No Condition Reports were written.

(4) Perform General Visual Examination of Fuel Transfer Tube Surface and Bolting Category (E-A): General Visual Examination - Item No. E1.11 - Three IWE General Visual and one VT-3 examination was performed. A summary of the 3 areas (four examinations total) this WO examined are:

(1) Category (E-A): General Visual Examination - (ID: 2-3-1-FFT-MB) - Concrete to liner interface of fuel transfer tube concrete and surface.

(2) Category (E-A): General Visual Examination - (ID: FTT-B) - Reactor Building fuel transfer tube bolting.

(3) Category (E-A): Item No. E1.11B - VT-3 Examination - (ID: FTT-B) - Reactor Building fuel transfer tube bolting (4) Category (E-A): General Visual Examination - (ID: FFT-S) - Reactor Building fuel transfer tube surface Inspection Results:

Examination of these areas found no unacceptable condition or indications. No Condition Reports were written.

Attachment I to WO 20-0029 Page 80 of 122 (5) Perform General Visual Examination of Test Connections Category (E-A): General Visual Examination - Item No. E1.30 - A total of 44 IWE General Visual Examinations were performed. A summary of the 44 areas this WO examined are:

(1) Category (E-A): General Visual Examination - (ID: TC-01, TC-02 and TC-03) -

Test connections in the Incore Tunnel under head on concrete floor. Test connections for test Zones 1, 2 and 3 and associated test plugs were examined.

(2) Category (E-A): General Visual Examination - (ID: TC-04 through TC-17) -

Containment elevation 2000 - Test connections for test Zones 4 through 17 and associated test plugs were examined.

(3) Category (E-A): General Visual Examination - (ID: TC-36, TC-37, TC-38 and TC-39) - Containment elevation 2025 - Test connections for test Zones 36, 37, 38 and 39 and associated test plugs were examined.

(4) Category (E-A): General Visual Examination - (ID: TC-40, TC-41, TC-43 and TC-44) - Containment elevations 2067, 2045, 2015 and 2022 for test Zones 40, 41, 43 and 44, respectively, and associated test plugs were examined.

Inspection Results (Summary):

During the IWE examination of the containment liner plate test connections, TC-08, TC-11 and TC-14 were noted to have surface cracks, flaking, peeling, or mechanical damage of the liner or coating. Additionally, the moisture barrier had cracking or flaking on these components that is noted per the WO. The degradation of TC-08, TC-11 and TC-14 components do not affect the structural integrity of the containment pressure boundary. The moisture barrier at TC-11 and TC-14 shall be repaired per WO 18-428270 and WO 18-438271. In addition, the flaking, peeling, blistering, or mechanical damage of TC-08, TC-11 and TC-14 should be repaired at the same time. No action is required for TC-07. The moisture barrier is already defined as augmented inspection, category (E-C), so no change is required. The test connection will not be added to augmented inspections category (E-C).

(6) Perform General/Detailed Visual Examination Containment Access Hatch Category (E-A): General Visual Examination - Item No. E1.11 - A total of 14 IWE General Visual Examinations were performed.

Category (E-C): Augmented Visual Examination - Item No. E4.11 - One IWE Detailed Visual Examination VT-1 was performed.

Attachment I to WO 20-0029 Page 81 of 122 A summary of the 15 areas this WO examined are:

(1) Category (E-C): Augmented VT-1 Examination - (ID: PAH-CYL-S) - Personnel Access hatch cylinder surface (2) Category (E-A): General Visual Examination - (ID: L-003-S) - Personnel Access hatch penetration surface (3) Category (E-A): General Visual Examination - (ID: PAH-EAAP-B) - Personnel Access hatch exterior door emergency air penetration bolting (4) Category (E-A): General Visual Examination - (ID: PAH-EDB-S) - Personnel Access hatch exterior door bulkhead surface (5) Category (E-A): General Visual Examination - (ID: PAH-EDEV-B) - Personnel Access hatch exterior door equalizing valve bolting (6) Category (E-A): General Visual Examination - (ID: PAH-EDLH-B) - Personnel Access hatch lower door lower hand wheel bolting (7) Category (E-A): General Visual Examination - (ID: PAH-ED-S) - Personnel Access hatch exterior door surface (8) Category (E-A): General Visual Examination - (ID: PAH-EDUH-B) - Personnel Access hatch exterior door upper hand wheel bolting (9) Category (E-A): General Visual Examination - (ID: PAH-IDB-S) - Personnel Access hatch interior door bulkhead surface (10) Category (E-A): General Visual Examination - (ID: PAH-IDEV-B) - Personnel Access hatch interior door equalizing valve bolting (11) Category (E-A): General Visual Examination - (ID: PAH-IDLH-B) - Personnel Access hatch interior door lower hand wheel bolting (12) Category (E-A): General Visual Examination - (ID: PAH-ID-S) - Personnel Access hatch interior door surface (13) Category (E-A): General Visual Examination - (ID: PAH-IDUH-B) - Personnel Access hatch interior door upper hand wheel bolting (14) Category (E-A): General Visual Examination - (ID: PAH-LLTP-B) - Personnel Access hatch exterior door left lock test penetration bolting (15) Category (E-A): General Visual Examination - (ID: PAH-RLTP-B) - Personnel Access hatch exterior door right lock test penetration bolting

Attachment I to WO 20-0029 Page 82 of 122 Inspection Results:

Overall, all 15 examinations were performed satisfactory with no reportable conditions. No CRs were written.

Results of Recent IWL Examinations In October of 2015, the most recent IWL examination was completed. This exam was the 30th Year IWL Tendon Surveillance of the WCGS containment buildings post-tensioning system and concrete structure. The surveillance was per Containment Tendon Inspection." This inspection is an implementation of IWL examination requirements and is completed once every 5 years per the program schedule. This examination is a systematic means of assessing the quality and structural performance of the post-tensioning system. Note: The 30th year, 2015 surveillance was performed in the latter part of the 2nd CISI Interval.

For the tendon selection for the 30th Year Surveillance, a total of seven (7) tendons were selected for inspection, of which four (4) were from hoop tendons, and three (3) were from vertical tendon groups. Tendons were examined appropriately per IWL Examination Category (L-B) requirements Also, as a part of this surveillance, a concrete IWL examination, per IWL Examination Category (L-A), requirements was performed.

The results of the most recent (2015) IWL examinations for both the tendons and concrete components examined are summarized below:

(1) Grease Cap Removal - Acceptable grease coatings were found on all tendon ends inspected, and no unusual conditions were reported at any tendon end.

(2) Free Water Inspection - All inspected tendon ends were found with no evidence of free water and are acceptable per IWL-3221.3(e).

(3) Sheathing Filler Analysis - Soluble ion concentrations, moisture contents, and base numbers of the tested sheathing filler samples were within acceptable limits for all inspected tendons, per IWL-3221.4.

(4) Anchorage Inspection (Buttonhead Count) - Hoop tendon 40BA was found with one (1) buttonhead protruding 1/2 on the field end that had not been previously reported. Non-conformance report was written by vendor performing examination and WCGS Engineering was notified. Vertical tendon V65 had two (2) missing buttonheads on the shop end - a condition that had previously been reported. All other buttonheads at all inspected tendon ends were found in acceptable condition, per IWL-3221.3(c).

WO requested an engineering evaluation of the as found visual inspection nonconformance for tendon 40BA found during the 30th year tendon surveillance work. The nonconformance was identified during the as-found visual inspection of tendon 40BA at A

Attachment I to WO 20-0029 Page 83 of 122 buttress end (field end); the button head of 1 of 170 tendon wires was found to be protruding from the anchor by approximately one-half inch.

The affected wire is the middle wire in the top row.

Engineering Disposition: Engineering approved the use-as-is of Reactor Buildings tendon 40BA due to ~0.5 elongation of 1 out of 170 stands.

(5) Anchorage Inspection (Corrosion Levels / Cracks) - No anchorage component cracks were observed on any inspected tendon end, and all inspected tendon anchorage components were found with an acceptable corrosion level of 1, per IWL-3221.3(a) and IWL-3221.3(b) respectively.

(6) Bearing Plate Concrete Inspection - All cracks observed within the 24 perimeter of the concrete surrounding each inspected bearing plate were less than 0.01 in width, per IWL-3221.3(d).

(7) Anchorhead Thread Measurements - Thread measurements taken at anchorheads of all inspected tendon ends were within the acceptable ranges with respect to the stressing ram adaptor that was used for performing lift-offs at that tendon end.

(8) Hydraulic Jack (Stressing Ram) Calibrations - The hydraulic jacks used for tendon lift-offs were calibrated before and after the surveillance period and were found to be within an acceptable variation of +/- 1.5%.

(9) Tendon Lift-Off Forces - The average value of all normalized tendon lift-off forces for a given tendon group exceeded the minimum required prestress value for their respective tendon group, per IWL-3221.1(a). All normalized individual tendon lift-off forces exceeded 95% of the predicted lower limit value for their respective tendon group, per IWL-3221.1(b).

(10) Comparison with Original Installation Data - In examining the average group percentage loss from the original lock-off force vales to the As-Found lift-off force values during the 30th year surveillance, no abnormal losses were observed. Additionally, a regression analysis was done to verify that the prestress force losses did not result in values below the minimum design value at the next surveillance period. Both of these analyses indicate that no abnormal average force differences were observed during this surveillance period. The average prestress force for each tendon group is expected to exceed the minimum design prestress force at the next scheduled examination, and the tendons are deemed acceptable per IWL-3221.1(c).

(11) Wire Visual Inspection & Tensile Testing - All test wires removed and tested were found to have acceptable corrosion levels of 1, diameter values within the acceptable range, and acceptably high yield stress, ultimate stress, and elongation percent values, per IWL-3221.2.

Attachment I to WO 20-0029 Page 84 of 122 (12) Tendon Restressing - All detensioned tendons were retensioned to acceptable forces and had acceptable elongations, per IWL-2523.3 and IWL-3221.1(d) respectively.

(13) Grease Cap & Grease Replacement - All tendon grease caps were properly installed to their respective tendon ends, and all tendon caps were refilled to acceptable levels, per IWL-3221.4.

(14) Exterior Concrete Surface Examination - There were no changes observed from the last inspection, and the noted observations, including small bug holes and shrinkage cracks, were recorded as IO (information only) items. The condition of the exterior concrete surface was deemed acceptable, per IWL-3211.

(15) Grease Cap Inspections - All inspected grease caps for both the Hoop and Vertical tendons were found in acceptable condition, per IWL-3211.

Summary:

In summary, the Final Report for the 30th Year Tendon Surveillance at Wolf Creek Generating Station has concluded that the functional integrity of the selected post-tensioning system has met the applicable code requirements, unless noted otherwise with non-conformance items, which were recorded, identified, and dispositioned as required.

3.5.5 Containment Leakage Rate Testing Program - Type B and Type C Testing Program The WCGS Types B and C LLRT testing program requires testing of electrical penetrations, airlocks, hatches, flanges and containment isolation valves (CIVs) in accordance with 10 CFR 50, Appendix J, Option B and RG 1.163. The results of the test program are used to demonstrate that proper maintenance and repairs are made on these components throughout their service life. The Types B and C testing program provides a means to protect the health and safety of plant personnel and the public by maintaining leakage from these components below appropriate limits. In accordance with the Program Plan for Containment Leakage Measurement, established to implement TS 5.5.16, the leakage rate acceptance criteria is < 0.60 La for the Types B and C tests.

(Note: Per the WCGS Program Plan, La is defined as 420,000 sccm and 0.60 La is 250,000 sccm).

As discussed, in NUREG-1493, Type B and Type C tests can identify the vast majority of all potential containment leakage paths. Types B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

WCGS performs As-Found LLRTs based on either the schedule per Option B requirements or due to activities/maintenance performed on a Type B or Type C component. Whenever an As-Found LLRT is performed the test results are tabulated into the LLRT spreadsheet and calculated into a running total for a combined as-left leakage for primary containment. Although As-Found tests are performed, a running summation of only the As-Found test data is not calculated as a part of the LLRT test Program. The running summation of most recently performed As-Left test results for

Attachment I to WO 20-0029 Page 85 of 122 both minimum and maximum path leakage, are maintained and monitored as a reflection of real-time condition of overall containment leakage of less than 0.6 La (defined as 250,000 sccm). Table 3.5.5-1 reflects the monitoring of As-Left leakage values. There have been no recorded instances where the As-Found minimum pathway has exceeded 1.0 La for the outages recorded in Table 3.5.5-1 below.

A review of the As-Left (AL) minimum and maximum pathway test values for WCGS Unit 1 can be summarized as:

• As-Left minimum pathway leak rate shows an average of 14.81% of 0.6 La with a high of 17.59% of 0.6 La.

• As-Left maximum pathway leak rate shows an average of 28.48% of 0.6 La with a high of 30.37% of 0.6 La.

Table 3.5.5-1 shown below, provides LLRT data trend summaries for WCGS since 2009 (last ILRT was performed in May 2011).

Table 3.5.5-1 WCGS Unit 1 Types B and C LLRT Combined As-Found/As-Left Trend Summary Outage No. R17 R18 R19 R20 R21 R22 R23 Year 2009 2011 2013 2015 2016 2018 2019 AL Max Path 72,853 67,429 73,050 75,452 73,231 60,385 75,913 (sccm) 1 Fraction of 29.14 26.97 29.22 30.18 29.29 24.15 30.37 0.6La AL Min Path 43,981 35,255 38,011 36,985 35,996 31,860 37,106 (sccm) 1 Fraction of 17.59 14.10 15.20 14.79 14.40 12.74 14.84 0.6La Note (1): Fraction of 0.6La was calculated as a percentage, based on 0.6La defined as 250,000 sccm.

Attachment I to WO 20-0029 Page 86 of 122 The As-left minimum pathway summations represent the high quality of maintenance of Type B and Type C tested components while the As-Left maximum pathway summations represent the effective management of the Containment Leakage Rate Testing Program.

3.5.6 Type B and Type C Local Leak Rate Testing Program Implementation Review 3.5.6.1 LLRT Failure of Program Valves Resulting in Test Interval Reduction Table 3.5.6-1 shown below, identifies WCGS components which were on Appendix J, Option B performance-based extended test intervals, but have not demonstrated acceptable performance during the previous two outages. The test intervals for the shown components encompasses 2 fuel cycles and testing performed in RF22 and RF23. Their testing frequencies have been reduced to a maximum of 30 months per Option B requirements and per plant program procedure. This reduction results in a test interval of one refueling cycle.

Table 3.5.6-1 WCGS Unit 1 Types B and C LLRT Program Implementation Review Admin As- Limit As-left Cause of Corrective Schedule Component Found Alert, sccm Failure Action d Interval sccm Action sccm RF22 - 2018 Disassembled CR written and for Pen 80 inspected. corrective Valve: BG8381 Retested, and maintenan found leakage ce Seat Charging 3800 3840 5800 increased to leakage pump 5800 sccm.

discharge Increase in Frequency check leakage set to one evaluated and refueling accepted cycle

Attachment I to WO 20-0029 Page 87 of 122 Table 3.5.6-1 WCGS Unit 1 Types B and C LLRT Program Implementation Review Admin As- Limit As-left Cause of Corrective Schedule Component Found Alert, sccm Failure Action d Interval sccm Action sccm CR written for (RF23 update) corrective maintenan Pen 80 ce Valve: BG8381 Seat Increase in 3800 8200 8200 leakage Charging leakage evaluated and Frequency pump accepted remains at discharge check one refueling cycle RF23 - 2019 None 3.5.6.2 Repeat LLRT Failures of Program Valves The following list identifies WCGS LLRT components which have shown a performance history of leak tests that repeatedly exceeded their Administrative Limit (Admin Limit) leakage criteria. The test intervals for these components have been reduced to an interval of one refueling cycle.

For the three components listed below, additional details of component performance are provided in the paragraphs that follow:

• Pen 99 - Valve GSHV0037, Containment Atmospheric Monitoring

• Pen 67 - Valve KCV0478, RB Fire Protection Inside Containment Check Valve

• Pen 64 - Valves: SJHV0129 and SJHV0130, Pzr/RCS Liquid Sample Outer Containment Isolation Valves

Attachment I to WO 20-0029 Page 88 of 122 Penetration 99 Valve GSHV0037 failed the As-Found LLRT the past 3 refueling outages and has been on a one-cycle test frequency since Refuel 20 (Fall 2015). In RF22 (Spring 2018) the valve was disassembled and found to have scratches on the disk. The valve was reworked and retested satisfactorily at 80 sccm (Admin Limit 1080 sccm). Because the penetration failed its As-Found LLRT and GSH0037 was rebuilt, the frequency for this penetration will remain at 18 months until two consecutive periodic As-Found test results are within the assigned administrative limits.

The past valve activities and test results for GSHV0037:

RF20 - AF = 1925 sccm (3/18/15), AL = 1925 sccm, set at one refueling cycle frequency.

RF21 - AF = 1125 sccm (10/10/16), AL = 1125 sccm, remains one refueling cycle frequency.

RF22 - AF = 2800 sccm, repair performed, on 4/29/18, AL = 80 sccm, remains on one refueling cycle frequency.

RF23 - AF = 1080 sccm (10/15/19), remains on one refueling cycle frequency.

Penetration 67 Valve KCV0478 had a history of leakage failures and experienced several unsuccessful attempts to repair/replace the valve to achieve acceptable leak tightness. Valve has been on one-cycle test frequency since Refuel 16 after the first replacement of the valve (Spring 2008). The valve was downgraded to Maintenance Rule (a)(1) on May 2, 2013. At end of RF20 (Spring 2015), WO was written to replace valve in RF21. In Refuel 21 (Fall 2016) the valve was replaced from the old swing check to a nozzle check valve and successfully passed the LLRT. Leakage monitoring on a one-cycle basis is in progress and will continue for three successful performances before considering extending test frequency and upgrading the valve back to Maintenance Rule (a)(2).

Throughout the diagnostic process, the Admin Limit leakage limit was initially at 5750 sccm, was raised in RF18 to10,000 sccm and once acceptable performance history is achieved, will be returned to original limit of 5750 sccm.

The past valve activities and test results for KCV0478:

RF16 - AF on 3/20/08 = 18,000 sccm, Valve replaced, AL on 4/5/08 = 3,600 sccm.

RF17 - AF on 10/22/09 = 9000 sccm, AL = 9000 sccm, CR written.

RF18 - AF on 3/28/11 = 7800 sccm, AL on 5/1/11 = 6000 sccm, CR written.

Attachment I to WO 20-0029 Page 89 of 122 RF19 - AF on 2/7/13 = 13,000 sccm, AL = 13,000, CR written.

MC20 - AF on 3/15/14 = 20,000 sccm, AL on 4/2/14 = 14,800 sccm.

RF20 - AF on 3/15/15 = 12,000 sccm, AL on 4/12/15 = 8,500 sccm, to replace in RF21.

RF21 - AF on 10/7/16 = 9752 sccm, Valve replaced, AL on 11/2/16 = 4000 sccm.

RF22 - AF on 4/11/18 = 700 sccm, passed test criteria, AL will be the AF.

RF23 - AF on 9/26/19 = 5,400 sccm. AL passed test criteria on 10/62/19. AL = 5500 sccm. Remains on one refueling cycle frequency.

Initial As-Found testing indicated elevated leakage through KCHV0253 (5500 sccm - failure on 09/26/2019) with KCV0478 passing with 5400 sccm. This test also indicated that KCV0478 was seated and adequately leak tight for the cycle leading into RF23. Boundary valve leakage was suspected as an explanation for the similar leak rates through both valves. A follow-up leakage test was performed revealing boundary valve leakage as the predominant source on September 29, 2019. Boundary valve leakage accounted for total leakage indicating KCV0478 had essentially zero leakage; however this also indicated KCV0478 had become unseated during the As-Found testing performed on September 26, 2019, based on inability to attain test pressure on 09/29/2019.

This constituted a test failure leading to the initiation of a CR identifying KCV0478 as the problem because this test confirmed KCHV0253 to be in excellent condition.

The CR further recommended attempting to flush any particulate, potentially preventing seating of the valve, out of the seating area, then re-perform the local leak rate test on KCV0478. This flush was completed with the follow-up test performed yielding a similar outcome as previously. It was decided that one more flushing attempt would be made. This was again followed by reperforming the LLRT to determine necessity of maintenance. At the onset, test pressure for KCV0478 could not be achieved (internal leak-by was audible) constituting a failed leak rate test. For informational purposes, a dead-blow was used to mechanically agitate the valve - a few light taps enabled seating of the plug allowing test pressure and leak rate to be obtained (this does not reverse test failure). The surveillance was continued by checking boundary valve leakage as well as KCHV0253. All leakage was attributable to boundary valves resulting in zero leakage from KCHV0253. After achieving a successful leak rate of both KCHV0253 and quantifying boundary valve leakage, testing transitioned back to KCV0478 to determine accuracy of the hypothesis that once the valve was seated, it would remain properly seated for the upcoming cycle. To accomplish this, KCV0478 was retested and found unseated (this was expected based on the testing performed 09/26/2019 and 09/29/2019 on KCHV0253 where the act of pressurizing the test volume for KCHV0253 unseated KCV0478) so the dead-blow was employed once again resulting in seating KCV0478 and subsequently test pressure was achieved.

Attachment I to WO 20-0029 Page 90 of 122 To validate the plug would remain seated for the upcoming cycle, the test volume was depressurized and allowed to achieve a non-stressed condition. Air was reintroduced to the test volume resulting in audible excessive leakage through KCV0478. This disproved the hypothesis and validated a problem with KCV0478 thus necessitating maintenance be performed to restore expected valve performance, and a CR was initiated to document this failure.

Maintenance revealed nothing remarkable - all tolerances were within specifications. and no indication of degradation to valve internal components was reported. Final assembly included installation of the resilient seat option with the expectation of this change improving future valve reliability. As-Left testing was acceptable.

Because the As-Found test of KCV0478 was satisfactory, the issue of past operability was eliminated; however, in light of the subsequent failed tests and required maintenance, this valve is expected to remain in Maintenance Rule (a)(1) status with the administrative leakage limit of 10,000 sccm retained until such time as two successful As-Found leak rate tests have been accomplished (thus allowing it to move out of (a)(1)). This may include a design change in favor of a stiffer spring, a different valve, or a change in piping configuration such that gravity helps in closing the currently installed nozzle check valve. An Equipment Performance Evaluation (EPE) was completed because of this failure.

Penetration 64 Valves SJHV0129 and SJHV0130 have a history of leakage failures. In RF20 (Fall 2015), these valves failed their Admin Limit of 2800 sccm and were placed on a one-cycle testing frequency (valves are tested in parallel pairs). Valves were retested in RF21 and exceeded their Admin Limit but showed moderate improvement. Rework of valves was rescheduled for RF22. In RF22 (Spring 2018) valves were replaced and successfully passed their As-left LLRT with a 20 sccm leakage value. Valves were retested in RF23 to initiate a performance history.

The past valve activities and test results for SJHV0129 and SJHV0130:

RF20 - AF on 3/30/15 = 5650 sccm, AL on 4/12/15 = 5430 sccm, WO written to schedule and rework both valves.

RF21 - AF on 9/29/16 = 4000 sccm, AL = AF, Rework was rescheduled for RF22.

RF22 - AF on 4/10/18 = 700 sccm, Valves replaced, AL on 4/27/18 = 20 sccm.

RF23 - AF on 10/02/19 = 20 sccm.

Attachment I to WO 20-0029 Page 91 of 122 In summary for these three penetrations, As-Found LLRTs will be performed during RF24 on each of these valves in order to establish a performance history.

3.5.6.3 Missed Surveillance Test Intervals for Program Valves While performing a review of the WCGS containment valve LLRT testing, it was identified that the performance of As-Found local leakage rate tests on three valves in RF22 had been missed. As-Found LLRTs are programmatically required following poor valve leakage performance or maintenance activities/repair (which occurred in RF21). In addition, the test schedule should have been reset to the base interval to test each successive refueling outage, in accordance with WCGS Program Plan for Containment Leakage Measurement.

A description of the three valves that missed the appropriate surveillance tests include:

• BBV0208 (Penetration 40) - Inner Containment valve for D Reactor Coolant Pump Seal Injection, Surveillance leak test: STS PE-140,

• EGV0204 (Penetration 74) - Component Cooling Water to Reactor Coolant Pumps Check Valve, Surveillance leak test: STS PE-174, and

• SJHV0128 (Penetration 64) - Pressurizer/Reactor Coolant System Liquid Sample Inner Containment Isolation Valve, Surveillance leak test: STS PE-164.

The activities performed on these valves that resulted in the requirement for the LLRT testing interval and the requirement to be set at the base interval are as follows:

• BBV0208, Valve replaced in RF21 (Reference requirement in NEI 94-01 Revision 0, Section 10.2.3.3, third paragraph),

• EGV0204, Valve replaced in RF21 (Reference requirement in NEI 94-01 Revision 0, Section 10.2.3.3, third paragraph),

• SJHV0128, Valve exceeded its Administrative Limit of 1440 sccm with As-Found leakage of 33,400 sccm in RF21, replaced valve. (Reference requirement in NEI 94-01 Revision 0, Sections 10.2.3.3 and 10.2.3.4 2).

Testing in RF21 (Fall 2016) after the above poor performance and valve corrective activities, resulted in the As-left (AL) test values to be acceptable based on their Admin Limit test criteria. The leakage test results recorded for each valve with respect to their Admin Limits were:

• Valve BBV0208, Tested 10/26/16; AL = 1100 sccm (Admin limit 3840 sccm),

• Valve EGV0204, Tested 10/30/16; AL = 1700 sccm (Admin limit 7680 sccm),

Attachment I to WO 20-0029 Page 92 of 122

• Valve SJHV0128, Tested 10/14/16; AL = 0 sccm (Admin limit 1440 sccm).

As-Found testing did not occur on these valves in RF22 (Spring 2018) and should have been, had their test interval been appropriately reset to their base test interval (one cycle). Condition Reports have been written on the missed test intervals to review the correction/extent of condition and impact on the plant. In addition, Work Orders have been written to perform LLRTs in RF23 (Fall 2019). Testing was performed in RF23 in order to initiate a performance history prior to considering extending the test interval.

• Valve BBV0208, Tested 10/02/19; AF = 650 sccm (Admin limit 3840 sccm),

• Valve EGV0204, Tested 09/27/19; AF = 4500 sccm (Admin limit 7680 sccm),

• Valve SJHV0128, Tested 10/02/19; AF = 20 sccm (Admin limit 1440 sccm).

Prior to the beginning of RF23, an Extent of Condition review identified the following additional missed LLRT surveillances. CRs were generated to address the Extent of Condition review and to ensure testing during RF23:

• P-36 FLANGE, Tested 09/26/19; AF = 450 sccm (Admin limit 1000 sccm)

• Valve HBHV7126, Tested 09/29/19; AF = 60 sccm (Admin limit 1920 sccm)

• North Electrical Pen Bank, Tested 09/23/19; AF = 1500 sccm (Admin limit 2000 sccm)

• Valve EMHV8888, Tested 10/12/19; AF = 2300 sccm (Admin limit 1920 sccm)

• Pen 51 Flanges, Tested 10/08/19; AF = 0 sccm (Admin limit 500 sccm)

• Pen 68 Flange, Tested 10/16/19; AF = 0 sccm (Admin limit 1000 sccm) 3.5.6.4 WCGS Type B and C LLRT Performance Summary The percentage of the total number of WCGS Type B surveillances that are on extended performance-based test intervals is 57.0%.

Those Type B surveillances not on an extended test frequency are either:

• Used during RFOs and therefore must be AL tested each RFO subsequent to use, or

• Limited to a test frequency of 30 months per the Primary Containment Leakage Rate Testing Program.

Attachment I to WO 20-0029 Page 93 of 122 The percentage of the total number of WCGS Type C tested components that are on extended performance-based test intervals is 76.6%.

The Type C penetrations not on an extended test frequency for a reason other than exceedance of their administration limit during an As Found Type C test are either:

• On a 30-month frequency (resulting in a test every RFO) following valve replacement or major maintenance to re-establish their performance history of two satisfactory consecutive AF tests,

• Used or removed during RFOs to support outage requirements,

• Tested on an RFO frequency to satisfy IST 24-month test frequency requirements, or

• Have met the performance requirements of two satisfactory consecutive AF tests after the last Type C test but have not yet been moved to an extended test frequency of 60 months.

3.6 OPERATING EXPERIENCE (OE)

During the conduct of the various examinations and tests conducted in support of the containment related programs previously mentioned, issues that do not meet established criteria or that provide indication of degradation, are identified, placed into the site's corrective action program, and corrective actions are planned and performed.

For the WCGS Primary Containment, the following site specific and industry events have been evaluated for impact:

• Information Notice (IN) 1992-20, Inadequate Local Leak Rate Testing

• IN 2010-12, "Containment Liner Corrosion"

• IN 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner

• RIS-2016-07, Containment Shell or Liner Moisture Barrier Inspection

• License Event Report (LER) 2020-001-00 for WCGS Each of these operating experiences are discussed in detail in Sections 3.6.1 through 3.6.5 below.

Attachment I to WO 20-0029 Page 94 of 122 3.6.1 IN 1992-20, Inadequate Local Leak Rate Testing The NRC issued IN 1992-20 to alert licensees of problems with local leak rate testing of two-ply stainless steel bellows used on piping penetrations at four different plants: Quad Cities, Dresden Nuclear Station, Perry Nuclear Plant, and the Clinton Station. Specifically, LLRTs could not be relied upon to accurately measure the leakage rate that would occur under accident conditions, because, during testing, the two plies in the bellows were in contact with each other, restricting the flow of the test medium to the crack locations. Any two-ply bellows of similar construction may be susceptible to the problem. The common issue in the four events was the failure to adequately perform local leak rate testing on different penetration configurations leading to problems that were discovered during ILRT tests in the first three cases.

In the event at Quad Cities, the two-ply bellows design was not properly subjected to LLRT pressure and the conclusion of the utility was that the two-ply bellows design could not be Type B LLRT tested as configured.

In the events at both Dresden and Perry, flanges were not considered a leakage path when the Type C LLRT test was designed. This omission led to a leakage path that was not discovered until the plant performed an ILRT test.

In the event at Clinton, relief valve discharge lines that were assumed to terminate below the suppression pool minimum drawdown level were discovered to terminate at a level above that datum. These lines needed to be reconfigured and the valves should have been Type C LLRT tested.

WCGS Discussion At WCGS there are containment bellows on instrument sensing lines at penetrations 103 and 104.

As described in the USAR, these sensing line transmitters are located immediately adjacent to the outside of the containment wall and are connected to a sealed bellows, located immediately adjacent to the inside containment wall by means of a sealed fluid filled tube. Because of this sealed fluid filled system, a postulated severance of the line during either normal operation or accident conditions will not result in any release from the containment. Both the bellows and tubing inside the containment are enclosed in protective shielding. This shielding (box, channel or guard pipe, etc.) prevent mechanical damage to the component from missiles, water jets, dropped tools, etc. The bellows are not tested by an LLRT. Leak testing is performed when an overall Type A ILRT test is conducted.

Additionally, at WCGS, there are bellows on the fuel transfer tube that spans between the reactor building and the fuel handling building. The USAR discusses this penetration and depicts the bellows as installed between the 20-inch diameter stainless steel pipe and a 26-inch diameter carbon steel sleeve that overall, makes up fuel transfer penetration, No. P-17. The fuel transfer tube contains a closure flange with gasketed seals and is subject to Type B LLRT pressure testing.

Attachment I to WO 20-0029 Page 95 of 122 The bellows associated with the fuel transfer tube are not part of the containment boundary.

The event described in this IN for Dresden and Clinton does not apply to WCGS, as it applies to a BWR plant and WCGS is a PWR.

The event described in this IN for Perry was reviewed and concluded that appropriate procedures are in place for testing all flanged penetrations at WCGS.

3.6.2 IN 2010-12, "Containment Liner Corrosion" IN 2010-12 was issued to alert plant operators to three events that occurred where the steel liner of the containment building was corroded and degraded. At Beaver Valley and Brunswick plants, material was found in the concrete, which trapped moisture against the liner plant and corroded the steel. In one case, it was material intentionally placed in the building and in the other case, it was foreign material which had inadvertently been left in the concrete form when the wall was poured.

But the result in both cases was that the material trapped moisture against the steel liner plate leading to corrosion. In the third case, an insulating material placed between the concrete floor and the steel liner plate at Salem adsorbed moisture and led to corrosion of the liner plate.

Subsequent to IN 2010-12, the NRC issued Technical Letter Report - Revision 1, Containment Liner Corrosion Operating Experience Summary, (Reference 19), on August 2, 2011, that summarized this topic across the nuclear industry. The Technical Letter addresses operating plants that have containment buildings constructed with carbon steel liners in contact with concrete.

In the United States, there are 55 pressurized water reactors (PWRs) and 11 boiling-water reactors (BWRs) with carbon steel liners in contact with concrete. The focus of the Technical Letter was to evaluate steel containment liner corrosion initiated at the liner/concrete interface.

WCGS Discussion The CISI Program requires 100% visual examination completed of the accessible area of the liner each period (every 40 months). Relevant inspections are listed in the Program Category (E-A) and (E-C) Tables. Specifically addressing IN 2010-12 concerns, Category (E-A), Items No. E1.11, Containment Vessel Pressure Retaining Boundary; Accessible Surface Areas and Items No.

E1.30, Containment Vessel Pressure Retaining Boundary; Moisture Barriers describe inspections perform on the containment. The subject table in the WCGS CISI Program follows the guideline provided in ASME Section XI, Subsection IWE, Table IWE-2500-1 (E-A) and (E-C).

The Beaver Valley degradation (an example of organics imbedded in concrete) was found by noting a blister on the liner coating. This attribute is listed on the examination checklist used when performing the visual examinations for the WCGS CISI program.

Although not discussed in the IN, it is noted that regulation 10 CFR 50.55a (b)(2)(ix)(A) requires that licensees evaluate the acceptability of inaccessible areas when conditions exist in accessible

Attachment I to WO 20-0029 Page 96 of 122 areas that could indicate the presence of or result in degradation of such inaccessible areas and to report each inaccessible area in the ISI summary report. This requirement is captured in the WCGS CISI program plan. Currently, no areas meeting these criteria, have been identified at WCGS. It is noted that the area of the WCGS containment liner with the moisture barrier between the containment floor and the vertical liner wall is accessible for examination. Minor corrosion has been identified on areas of the liner above the containment floor. These areas have been evaluated and listed in the WCGS CISI program plan to require augmented examination in accordance with examination category (E-C). Additionally, the liners of the containment normal sumps are covered with a metal cover. While these areas might have been considered inaccessible, the covers of the sumps are removed periodically, and the liners examined.

Note: The areas of the WCGS containment liner that are most susceptible to degradation are the liners of the incore sump and the two normal sumps (submerged areas). In every IWE examination of these sump liners, coating failure and subsequent pitting of the liner has been found, with some of the pitting being severe. Funding was approved for a study to determine the best, most cost-effective solution for the degradation of the sump liners. Initially it was thought that replacement of the liner with stainless steel plate would be the best solution. However, the study determined that a replacement coating would provide the best solution. This replacement coating was installed during RF18. These areas are listed as Examination Category (EC) items in the CISI Program and are scheduled to remain in the augmented examination category because of the conditions in the sumps. (In the current 3rd 10-year Interval, sumps are still in augmented Category (EC) 3.6.3 IN 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner IN 2014-07 addresses concerns identified by the NRC for degradation of floor weld leak-chase channel systems of steel containment shell and concrete containment metallic liners that could affect leak-tightness and aging management of containment structures. This IN also explicitly states the NRCs interpretation that leak chase channels should be considered a moisture barrier as defined in ASME Section XI, Subsection IWE. Although no specific actions or written responses were required by the IN, it was evaluated under WCGS corrective action program in technical evaluation CR 86026.

WCGS Discussion The IN identified floor leak-chase test connections degradation as a result of water intrusion through the floor level test connection cover plates. WCGS does not have leak-chase test connections that are below the concrete floor level that would have water intrusion as noted. The WCGS leak-chase test connections are above floor grade. However, because of the increased awareness of water intrusion into the leak-chase system, the inspection description of test point locations were revised in the 2nd 10-year Interval of the WCGS CISI Program, Revision 6 to include the test plug.

Attachment I to WO 20-0029 Page 97 of 122 The subject leak chase channel inspections are inspected per the 3rd 10-year Interval updated CISI IWE Program found in Category (E-A), Containment Surfaces, Item E1.30, Containment Vessel Pressure Retaining Boundary; Moisture Barriers.

3.6.4 RIS-2016-07 Containment Shell or Liner Moisture Barrier Inspection The NRC issued RIS 2016-07, to reiterate their position on inservice inspection requirements for moisture barrier materials, as noted in ASME Section XI, Subsection IWE. The Code requires examination to include moisture barrier material intended to prevent intrusion of moisture against inaccessible areas of the pressure retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces which are not seal-welded. Containment moisture barrier materials include caulking, flashing, and sealant used for the applications.

WCGS Discussion The 2nd Interval, and current 3rd 10-year Interval, CISI Program Plan, detail the requirements for examination and testing requirements per ASME Section XI and 10CFR 50.55a at Wolf Creek. The CISI program plan identifies moisture barrier locations at the containment floor elevation 2000, the reactor cavity area and the fuel transfer tube concrete to liner interfaces. The program plan also identifies test connections for the leak chase system for examination. To address this RIS, a plant equipment location drawing review was performed and determined that the concrete and the grating steel do not come in contact with the liner. Lastly a walkdown of containment was performed in RF21 (October 11, 2016) to determine if there are any additional locations where the containment liner is in contact with concrete or steel and has some type of moisture barrier installed that is not presently identified in the CISI Program Plan. A walkdown of the 2000, 2026, 2047 and 2068 elevation levels of containment were performed, and no additional moisture barrier locations were identified. No additional action was necessary, and the action was closed.

3.6.5 LER 2020-001-00, Plant Shutdown Due to Inoperable Containment Purge Isolation Valves (Reference 38)

Abstract:

On February 1, 2020, while in Mode 1 at 100% power, surveillance testing of containment isolation valves associated with the containment shutdown purge supply piping was being conducted. It was discovered that the leakage rate through the penetration was greater than that allowed by TS.

Two containment isolation valves in series were determined to be inoperable. This led to entry into TS Limiting Condition for Operation (LCO) 3.6.3 Condition E, which requires the plant be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Due to the high leakage rate, containment was also declared inoperable so WCGS entered TS LCO 3.6.1 Condition A which requires restoration of containment to operable status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This was not possible, so TS LCO 3.6.1 Condition B was entered which also requires the plant to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Attachment I to WO 20-0029 Page 98 of 122 At 2154 Central Standard Time (CST) on February 1, 2020, WCGS completed a shutdown required by TS. Therefore, this is being reported in accordance with 10 CFR 50.73(a)(2)(i)(A). In addition, because containment was declared inoperable, this is also being reported in accordance with 10 CFR 50.73(a)(2)(v) as a condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material as well as to mitigate the consequences of an accident.

Both valves were returned to service the following day, and WCGS subsequently returned to Mode 1 on February 3, 2020.

Event

Description:

WCGS Refueling Outage 23 (RF23) began on September 21, 2019. During RF23, leak rate testing was performed on GTHZ0006 and GTHZ0007. The administrative leakage rate limit is 12,000 sccm for each of these valves. The TS surveillance acceptance criteria for the penetration (which includes the mini-purge valve leakage) is 21,000 sccm. Prior to RF23, previous tests had shown both valves had measured leak rates which were within their administrative limits. The As-Left leakage discovered during RF23 for both valves exceeded the TS limit. Maintenance work was performed on GTHZ0007, and prior to entering Mode 4 coming out of RF23 (which occurred at 0850 Central Standard Time (CST) on November 2, 2019), measured leakage through penetration V-161 was 10,500 sccm. This measurement was the summation of leakage through GTHZ0004, GTHZ0005, and GTHZV0007 because the blind flange was installed on GTHZ0006 which eliminated this leak path. This flow rate is below the administrative limit for this penetration. The decision was made to postpone maintenance work on GTHZ0006 as it is the outside containment isolation valve and could be performed online.

In the case of one containment purge isolation valve not within leakage limits, TS LCO 3.6.3 Condition D is entered. TS LCO 3.6.3 Required Action D.1 requires that the affected penetration flow path be isolated by the use of at least one closed and de-activated automatic valve or closed manual valve. To meet TS LCO 3.6.3 Condition D prior to entering Mode 4 (the first mode of applicability), GTHZ0007 and GTHZ0005 (the minipurge supply inside containment isolation valve) were both verified closed and deactivated. In addition, the blind flange associated with GTHZ0006 was installed to enable leakage measurement of penetration V-161 (GTHZ0004, GTHZ0005, and GTHZ0007).

Required Action D.3 requires that leak rate testing be performed every 92 days on those purge valves that have resilient seals that are closed to comply with Required Action D.1. On 2/1/2020, surveillance procedure Containment Purge Valve Leakage Test was performed to meet Required Action D.3. The as-found leakage rate at the time of this performance was greater than 250,000 sccm which is the allowed TS limit leakage for the containment building. As a result, at 1845 CST on February 1,2020, LCO 3.6.3 Required Action D.3 could not be met. This required entry into LCO 3.6.3 Condition E which directs the plant to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Because the leak rate was greater than the TS limit, containment was declared inoperable at the same time, so TS LCO 3.6.1 Condition A was entered for an inoperable containment.

Required Action A.1 is to restore containment to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. If Required Action A.1 cannot be completed, then Condition B requires the plant to be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

Attachment I to WO 20-0029 Page 99 of 122 and Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The plant entered Mode 3 at 2154 CST on February 1, 2020, within the Required Completion Time for the applicable Conditions in both LCO 3.6.1 and 3.6.3.

Troubleshooting determined that GTHZ0007 valve was leaking, in addition to the previously known leak through GTHZ0006. Repairs were first completed on GTHZ0007 and at 1009 CST on February 2, 2020, GTHZ0007 was declared operable. Because this allowed isolation of the containment penetration flowpath, containment was declared operable and TS LCO 3.6.1 Condition A was exited. GTHZ0006 was then also repaired. At 1638 CST on 2/2/2020, GTHZ0006 was declared operable. At this time TS LCO 3.6.3 Condition D was exited. With LCOs 3.6.1 and 3.6.3 met, WCGS subsequently began preparations to return to power operations and reached Mode 1 at 0656 CST on February 3, 2020.

3.7 LICENSE RENEWAL AGING MANAGEMENT On November 20, 2008, the NRC issued the renewed facility operating license No. NPF-42 for WCGS. The NRC issued the renewed facility operating license based on the staffs review of the WCGS License Renewal Application dated September 27, 2006, as supplemented by letters submitted to the NRC through August 1, 2008. The renewed facility operating license No. NPF-42 expires at midnight March 11, 2045. (Reference 21)

The NRC completed their evaluation and issued in October 2008, the WCGS Unit 1 LRA Safety Evaluation Report (SER) via NUREG-1915, Safety Evaluation Report Related to the License Renewal of Wolf Creek Generating Station Docket No. 50-482.

License renewal commitments are currently tracked at WCGS in USAR, Chapter 18, Appendix A, Appendix A Introduction and License Renewal Commitments, as required per 10 CFR 54.21(d),

and describe enhancements to the ISI Programs beyond the requirements of ASME Section XI.

This USAR Appendix A includes references to the ASME Section XI Containment ISI Program in Sections A1.27, A1.28, A1.30, A2.3, A3.4, and A3.5.

A summary of the applicable ASME Section XI and the 10 CFR 50 Appendix J commitments are shown in Table 3.7-1 below, License Renewal Commitments Supplementing ASME Section XI Requirements. Information included in Table 3.6-1 can also be found in the CISI Program Plan.

The table details the USAR source, the action tracking items, and the implementation schedule.

Additional paragraphs following the Table below provides an explanation of the programs/activities credited with aging management.

As part of the license renewal effort, it had to be demonstrated that the aging effects applicable for the components and structures within the scope of license renewal would be adequately managed during the period of extended operation.

Attachment I to WO 20-0029 Page 100 of 122 In many cases, existing activities were found adequate for managing aging effects during the period of extended operation. In some cases, aging management reviews revealed that existing activities required enhancement to adequately manage applicable aging effects. In a few cases, new activities were developed to provide added assurance that aging effects are adequately managed.

Changes to these credited activities may be made without prior NRC approval provided an evaluation of the change meets the criteria set forth in 10 CFR 50.59, Changes, tests and experiments.

WCGS plant procedure, License Renewal Implementation, describes the basic elements of license renewal implementation and identifies organizational responsibilities at WCGS. The procedure applies to the activities that are to be taken pursuant to the requirements of 10 CFR Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants.

Attachment I to WO 20-0029 Page 101 of 122 Table 3.7-1 License Renewal Commitments Supplementing ASME Section XI Requirements USAR Chapter 18, Program or Topic Summary Implementation Schedule Appendix A The ASME Section XI, Subsection ASME Section XI ISI The existing ASME Section XI ISI Subsection IWE CISI IWE containment inservice A1.27 Subsections IWE Program will manage this commitment inspection program is already in place at WCGS.

The ASME Section XI, Subsection ASME Section XI ISI The existing ASME Section XI ISI Subsection IWL CISI IWL containment inservice A1.28 Subsections IWL Program will manage this commitment. inspection program is already in place at WCGS.

The 10 CFR 50, Appendix J 10 CFR 50, Appendix The existing 10 CFR 50, Appendix J Program will manage A1.30 program is already in place at J this commitment.

WCGS The ASME Section XI, Subsection Concrete A2.3 and The existing ASME Section XI ISI Subsection IWL CISI IWL containment inservice Containment Tendon A3.4 Program will manage this commitment. inspection program is already in Prestress place at WCGS.

Containment Liner Fatigue load cycles of metallic components were No additional examinations or A3.5 and Plate, Polar Crane addressed during the license renewal process. At WCGS, commitments were created from A3.5.1 Bracket, And the only metallic components of the containment pressure this evaluation.

Attachment I to WO 20-0029 Page 102 of 122 Penetration Load boundary that are designed for a specific number of load Cycles cycles in a design lifetime are the main steam penetrations.

Design Cycles for the An evaluation was performed to determine the fatigue Main Steam Line lifetime of the main steam penetrations. It was determined Penetrations there was more than sufficient margin in the design for any possible increase in operating cycles above the original estimate. The design of the main steam penetrations is therefore valid for the period of extended operation

Attachment I to WO 20-0029 Page 103 of 122

• Appendix A1.27 - Chapter 18, USAR Supplement, License Renewal commitments, -

ASME Section XI, Subsection IWE The ASME Section XI, Subsection IWE containment inservice inspection (CISI) program provides aging management of the steel liner of the concrete containment building, including the containment liner plate, piping and electrical penetrations, access hatches, and the fuel transfer tube. Inspections are performed to identify and manage any containment liner aging effects that could result in loss of intended function. Acceptance criteria for components subject to Subsection IWE exam requirements are specified in Article IWE-3000.

In conformance with 10 CFR 50.55a(g)(4)(ii), the WCGS CISI Program is updated during each successive 120-month inspection interval to comply with the requirements of the latest edition and addenda of the Code specified twelve months before the start of the inspection interval.

• Appendix A1.28 - Chapter 18, USAR Supplement, License Renewal commitments, -

ASME Section XI, Subsection IWL The ASME Section XI, Subsection IWL CISI Program manages aging of the concrete containment structure (including the tendon gallery ceiling), the concrete dome, and the post-tensioning system. Inspections are performed to identify and manage any containment concrete aging effects that could result in loss of intended function.

In conformance with 10 CFR 50.55a(g)(4)(ii), the WCGS ISI Program is updated during each successive 120-month inspection interval to comply with the requirements of the latest edition and addenda of the Code specified twelve months before the start of the inspection interval.

Prior to the period of extended operation, procedures will be enhanced to include two new provisions regarding inspection of repair/replacement activities.

(Note: WCGS will enter the period of extended operation under License Renewal on March 12, 2025; therefore, 10 CFR 50.55a(b)(2)(viii)(I) will be applied to the CISI Program at that time.)

• Appendix A1.30 - Chapter 18, USAR Supplement, License Renewal commitments, -

10CFR50, Appendix J

Attachment I to WO 20-0029 Page 104 of 122 The 10 CFR Part 50, Appendix J Program monitors leakage rates through the containment pressure boundary, including the penetrations and access openings, in order to detect degradation of containment pressure boundary. Seals, gaskets, and bolted connections are also monitored under the program.

Containment leak rate tests are performed in accordance with 10 CFR 50 Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors" Option B; Regulatory Guide 1.163, Revision 0, "Performance-Based Containment Leak-Testing Program," NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50 Appendix J; and ANSI/ANS 56.8, "Containment System Leakage Testing Requirements."

Containment leak rate tests are performed to assure that leakage through the primary containment, and systems and components penetrating primary containment does not exceed allowable leakage limits specified in the TS. Corrective actions are taken if leakage rates exceed established administrative limits for individual penetrations or the overall containment pressure boundary.

• Appendix A2.3 - Chapter 18, USAR Supplement, License Renewal commitments, -

Concrete Containment Tendon Prestress The Concrete Containment Tendon Prestress Program, within the WCGS Creek ASME Section XI Subsection IWL Program, manages the loss of tendon prestress in the post-tensioning system. The WCGS post-tensioning system consists of inverted U-shaped tendons, extending up through the basemat, through the full height of the cylindrical walls and over the dome; and horizontal circumferential (hoop) tendons, at intervals from the basemat to about the 45-degree elevation of the dome. The basemat is conventionally reinforced. The tendons are ungrouted, in grease-filled glands.

Prior to the period of extended operation, procedures will be revised to extend the list of surveillance tendons to include random samples for the 40, 45, 50, and 55-year surveillances, to explicitly require a regression analysis for each tendon group after every surveillance; and to invoke and describe regression analysis methods used to construct the lift-off trend lines. Surveillance program predicted force lines for the vertical and hoop tendon groups will be extended to 60 years. Procedure descriptions of acceptance criteria action levels will be revised to conform to the ASME Code, Subsection IWL 3221 descriptions.

(Note: WCGS will enter the period of extended operation under License Renewal on March 12, 2025; therefore, 10 CFR 50.55a(b)(2)(viii)(I) will be applied to the CISI Program at that time.)

Attachment I to WO 20-0029 Page 105 of 122

• Appendix A3.4 - Chapter 18, USAR Supplement, License Renewal commitments, -

Concrete Containment Tendon Prestress The WCGS containment is a prestressed concrete, hemispherical-dome-on- a-cylinder structure, with a steel membrane liner. Post-tensioned tendons compress the concrete and permit the structure to withstand design basis accident internal pressures. The reinforced concrete basemat is conventionally reinforced.

To ensure the integrity of the containment pressure boundary under design basis accident loads, design predictions of loss of prestress demonstrate that prestress should remain adequate for the design life. An inspection program confirms that the tendon prestress remains within design limits throughout the life of the plant [USAR Section 3.8.1, TS SR 3.6.1.2].

Continuing the existing ASME Subsection IWL tendon surveillance program ensures that loss of prestress aging effects will be managed and that the containment tendons will continue to perform their intended functions for the period of extended operation.

• Appendix A3.5 - Chapter 18, USAR Supplement, License Renewal commitments, -

Containment Liner Plate, Polar Crane Bracket, And Penetration Load Cycles Design of the polar crane for a finite number of loads is a time limited aging analyses (TLAA) at WCGS (Ref. USAR Section A3.6.1). At some plants, though not at WCGS, the supporting crane rail brackets or other supporting structural elements may also have been designed for these cyclic loads. NUREG-1800 Section 4.6.1 notes that in some designs Fatigue of the liner plates or metal containments may be considered in the design based on an assumed number of loading cycles for the current operating term.

At WCGS however, the only metallic components of the containment pressure boundary that are designed for a specific number of load cycles in a design lifetime are the main steam penetrations. The containment liner and the remaining penetrations were designed to stress limit criteria, independent of the number of load cycles, and with no fatigue analyses.

• Appendix A3.5.1 - Chapter 18, USAR Supplement, License Renewal commitments, -

Design Cycles for the Main Steam Line Penetrations The BC-TOP-1, Containment Building Liner Plate Design Report, Part II Section 1.1, describes the main steam penetration design for cyclic loads. The design basis includes:

Attachment I to WO 20-0029 Page 106 of 122

• 100 lifetime steady state operating thermal gradient plus normal operating cyclic loads (Loading Condition V), and

• 10 steady state operating thermal gradient plus steam pipe rupture cyclic loads (Loading Condition IV).

The operating history to date indicates that the original design basis 100 operating cycles assumed for main steam penetrations will be adequate for the 60-year extended operating period. For license renewal an evaluation found that the penetrations could withstand as many as 2500 lifetime full-range thermal cycles, plus the fatigue effects of an end-of-life main steam rupture inside containment with a cumulative usage factor of less than 1.0.

Therefore, there is more than sufficient margin in the design for any possible increase in operating cycles above the original estimate. The design of the main steam penetrations is therefore valid for the period of extended operation.

3.8 NRC SER LIMITATIONS AND CONDITIONS 3.8.1 Limitations and Conditions Applicable to NEI 94-01, Revision 2-A The NRC staff found that the use of NEI TR 94-01, Revision 2, was acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT surveillance interval to 15 years, provided the following conditions as listed in Table 3.8.1-1 were satisfied:

Table 3.8.1-1 NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition WCGS Response (From Section 4.0 of SE)

For calculating the Type A leakage rate, the WCGS will utilize the definition in NEI 94-01 licensee should use the definition in the NEI Revision 3-A, Section 5.0. This definition has TR 94-01, Revision 2, in lieu of that in remained unchanged from Revision 2-A to ANSI/ANS-56.8-2002. (Refer to SE Section Revision 3-A of NEI 94-01.

3.1.1.1.)

Attachment I to WO 20-0029 Page 107 of 122 Table 3.8.1-1 NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition WCGS Response (From Section 4.0 of SE)

The licensee submits a schedule of Reference Section 3.5.2 (Tables 3.5.2-5 and containment inspections to be performed 3.5.2-8) of this LAR submittal.

prior to and between Type A tests. (Refer to SE Section 3.1.1.3.)

The licensee addresses the areas of the Reference Section 3.5.2 (Tables 3.5.2-2, containment structure potentially subjected to 3.5.2-3, 3.5.2-4, 3.5.2-6, and 3.5.2-7) of this degradation. (Refer to SE Section 3.1.3.) LAR submittal.

The licensee addresses any tests and There have been no major or minor inspections performed following major containment repairs or modifications modifications to the containment structure, as performed nor are any repairs or applicable. (Refer to SE Section 3.1.4.) modifications planned for the containment structure.

The normal Type A test interval should be WCGS will follow the requirements of NEI 94-less than 15 years. If a licensee has to utilize 01 Revision 3-A, Section 9.1. This the provision of Section 9.1 of NEI TR 94-01, requirement has remained unchanged from Revision 2, related to extending the ILRT Revision 2-A to Revision 3-A of NEI 94-01.

interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. (Refer to SE In accordance with the requirements of NEI Section 3.1.1.2.) 94-01, Revision 2-A, SER Section 3.1.1.2, WCGS will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

Attachment I to WO 20-0029 Page 108 of 122 Table 3.8.1-1 NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition WCGS Response (From Section 4.0 of SE)

For plants licensed under 10 CFR Part 52, Not applicable. WCGS was not licensed applications requesting a permanent under 10 CFR Part 52.

extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No.

1009325, Revision 2, including the use of past containment ILRT data.

3.8.2 Limitations and Conditions Applicable to NEI 94-01, Revision 3-A The NRC staff found that the guidance in NEI TR 94-01, Revision 3, was acceptable for referencing by licensees in the implementation of the optional performance-based requirements of Option B to 10 CFR 50, Appendix J. However, the NRC staff identified two conditions on the use of NEI TR 94-01, Revision 3 (Reference NEI 94-01 Revision 3-A, NRC SER 4.0, Limitations and Conditions), which are discussed as follows:

Topical Report Identified Condition 1 NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.

The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI TR 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g.,

BWR MSIVs), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

Attachment I to WO 20-0029 Page 109 of 122 Response to Identified Condition 1 Topical Report Condition 1 presents the following three (3) separate issues that are required to be addressed:

• ISSUE 1 - The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.

• ISSUE 2 - In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.

• ISSUE 3 - Use of the allowed 9-month extension for eligible Type C valves is only authorized for non-routine emergent conditions with exceptions as detailed in NEI 94-01, Revision 3-A, Section 10.1.

Response to Condition 1, ISSUE 1 The post-outage report shall include the margin between the Type B and Type C MNPLR summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.60 La.

Response to Condition 1, ISSUE 2 When the potential leakage understatement adjusted Types B and C MNPLR total is greater than the WCGS administrative leakage summation limit of 0.5 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the WCGS leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.

Response to Condition 1, ISSUE 3 WCGS will apply the 9-month allowable interval extension period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.

Attachment I to WO 20-0029 Page 110 of 122 Topical Report Identified Condition 2 The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending overall containment leakage potential, the As-Found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 refueling outages. Type C tests involve valves, which in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total is used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for.

Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.

When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Types B and C total leakage and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Response to Identified Condition 2 Topical Report Condition 2 presents the following two (2) separate issues that are required to be addressed:

• ISSUE 1 - Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.

Attachment I to WO 20-0029 Page 111 of 122

• ISSUE 2 - When routinely scheduling any LLRT valve interval beyond 60 months and up to 75 months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Types B and C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Response to Condition 2, ISSUE 1 The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25% in the LLRT periodicity. As such, WCGS will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate, which will increase the As-Left leakage total for each Type C component currently on greater than a 60-month test interval up to the 75-month extended test interval. This will result in a combined conservative Type C total for all 75-month LLRTs being "carried forward" and will be included whenever the total leakage summation is required to be updated (either while on-line or following an outage).

When the potential leakage understatement adjusted leak rate total for those Type C components being tested on greater than a 60-month test interval up to the 75-month extended test interval is summed with the non-adjusted total of those Type C components being tested at less than or equal to a 60-month test interval (Reference 36), and the total of the Type B tested components, results in the MNPLR being greater than the WCGS administrative leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the WCGS leakage limit. The corrective action plan should focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

Response to Condition 2, ISSUE 2 If the potential leakage understatement adjusted leak rate MNPLR is less than the WCGS administrative leakage summation limit of 0.50 La, then the acceptability of the greater than a 60-month test interval up to the 75-month LLRT extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

In addition to Condition 1, ISSUES 1 and 2, which deal with the MNPLR Types B and C summation margin, NEI 94-01, Revision 3-A, also has a margin-related requirement as contained in Section 12.1, Report Requirements.

Attachment I to WO 20-0029 Page 112 of 122 A post-outage report shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B and Type C tests, if performed during that outage. The technical contents of the report are generally described in ANSI/ANS-56.8-2002 and shall be available on-site for NRC review. The report shall show that the applicable performance criteria are met and serve as a record that continuing performance is acceptable. The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit. Adverse trends in the Type B and Type C leakage rate summation shall be identified in the report and a corrective action plan developed to restore the margin to an acceptable level.

At WCGS, in the event an adverse trend in the aforementioned potential leakage understatement adjusted Types B and C summation is identified, then an analysis and determination of a corrective action plan shall be prepared to restore the trend and associated margin to an acceptable level. The corrective action plan shall focus on those components which have contributed the most to the adverse trend in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

At WCGS, an adverse trend is defined as three (3) consecutive increases in the final pre-mode change Types B and C MNPLR leakage summation values, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.

3.9 Conclusion NEI 94-01, Revision 3-A, dated July 2012, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008, describe an NRC-accepted approach for implementing the performance-based requirements of 10 CFR 50, Appendix J, Option B. It incorporated the regulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals to 15 years and Type C test intervals to 75 months. NEI 94-01, Revision 3-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies. WCGS is adopting the guidance of NEI 94-01, Revision 3-A, and the limitations and conditions specified in NEI 94-01, Revision 2-A, for the WCGS Unit 1, 10 CFR 50, Appendix J testing program plan.

Based on the previous ILRTs conducted at WCGS Unit 1, WCNOC concludes that the permanent extension of the containment ILRT interval from 10 to 15 years represents minimal risk to increased leakage. The risk is minimized by continued Type B and Type C testing performed in accordance with Option B of 10 CFR 50, Appendix J, and the overlapping inspection activities performed as part of the following WCGS inspection programs:

• Containment Inservice Inspection Program (IWE)

• Containment Inservice Inspection Program (IWL)

• Nuclear Coatings Program

Attachment I to WO 20-0029 Page 113 of 122

• Containment Inspections per TS SR 3.6.1.2 (Tendon Surveillance)

This experience is supplemented by risk analysis studies, including the WCGS risk analysis provided in the Enclosure. The risk assessment concludes that increasing the ILRT interval on a permanent basis to a one-in-fifteen-year frequency is not considered to be significant because it represents only a small change in the WCGS risk profile.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.

10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR 50, Leakage Rate Testing of Containment of Water-Cooled Nuclear Power Plants. Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.

The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B and Type C testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of 10 CFR 50, Appendix J, the test frequency is based upon an evaluation that reviewed "As-Found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type C test frequencies will not directly result in an increase in containment leakage.

EPRI TR-1009325, Revision 2-A (Reference 11), provided a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance. NEI 94-01, Revision 3-A, Section 9.2.3.1, states that Type A ILRT intervals of up to 15 years are allowed by this guideline. The Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI Report 1018243 (formerly TR-1009325, Revision 2-A), indicates that, in general, the risk impact associated with ILRT interval extensions for intervals up to 15 years is small. However, plant-specific confirmatory analyses are required.

The NRC staff reviewed both NEI TR 94-01, Revision 2 (Reference 8), and EPRI Report No.

1009325, Revision 2. For NEI TR 94-01, Revision 2, the NRC staff determined that it described an acceptable approach for implementing the optional performance-based requirements of Option

Attachment I to WO 20-0029 Page 114 of 122 B to 10 CFR 50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals up to 15 years and incorporates the regulatory positions stated in RG 1.163. The NRC staff finds that the Type A testing methodology, as described in ANSI/ANS-56.8-2002, and the modified testing frequencies recommended by NEI TR 94-01, Revision 2, serve to ensure continued leakage integrity of the containment structure. Type B and Type C testing ensures that individual penetrations are essentially leak tight. In addition, aggregate Type B and Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.

For EPRI Report No. 1009325, Revision 2, a risk-informed methodology using plant-specific risk insights and industry ILRT performance data to revise ILRT surveillance frequencies, the NRC staff finds that the proposed methodology satisfies the key principles of risk-informed decision making applied to changes to TS as delineated in RG 1.174 and RG 1.177, An Approach for Plant-Specific, Risk-Informed Decision making: Technical Specifications (Reference 37). The NRC staff, therefore, found that this guidance was acceptable for referencing by licensees proposing to amend their TS in regard to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.2 of the SE (Reference 9).

The NRC staff reviewed NEI TR 94-01, Revision 3, and determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, as modified by the limitations and conditions summarized in Section 4.0 of the associated SE. This guidance included provisions for extending Type C LLRT intervals up to 75 months. Type C testing ensures that individual CIVs are essentially leak tight. In addition, aggregate Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The NRC staff, therefore, found that this guidance, as modified to include two limitations and conditions, was acceptable for referencing by licensees proposing to amend their TS in regard to containment leakage rate testing. Any applicant may reference NEI TR 94-01, Revision 3, as modified by the associated SER and approved by the NRC, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008, in a licensing action to satisfy the requirements of Option B to 10 CFR 50, Appendix J.

4.2 Precedent This LAR is similar in nature to the following license amendments to extend the Type A Test Frequency to 15 years and the Type C test frequency to 75 months as previously authorized by the NRC in the associated referenced SERs:

• Comanche Peak Nuclear Power Plant, Units 1 and 2, issued December 30, 2015 (Reference 17 - ML15309A073)

Attachment I to WO 20-0029 Page 115 of 122

• McGuire Nuclear Station, Units 1 and 2, issued January 31, 2018 (Reference 18 - ML18009A842)

• Vogtle Electric Generating Plant, Units 1 and 2, issued October 29, 2018 (Reference 22 - ML18263A039) 4.3 No Significant Hazards Consideration Wolf Creek Nuclear Operating Corporation (WCNOC) proposes to amend the Technical Specifications (TS) for Wolf Creek Generating Station (WCGS) Unit 1 to allow extension of the Type A and Type C leakage test intervals. The extension is based on the adoption of the Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, and conditions and limitations set forth in Revision 2-A.

Specifically, the proposed change revises WCGS TS 5.5.16, Containment Leakage Rate Testing Program, by replacing the reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, with reference to NEI 94-01, Revision 3-A and the limitations and conditions set forth in Revision 2-A.

WCNOC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed activity involves the revision of Wolf Creek Generating Station (WCGS),

Unit 1 Technical Specifications (TS) Section 5.5.16, Containment Leakage Rate Testing Program, to allow the extension of the Type A integrated leakage rate test (ILRT) containment test interval to 15 years, and the extension of the Type C local leakage rate test (LLRT) interval to 75 months. Per the guidance provided in Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, Revision 3-A, the current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months.

Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions.

The proposed interval extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of

Attachment I to WO 20-0029 Page 116 of 122 radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.

The change in Type A test frequency to once-per-fifteen years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, based on the internal events (IE) probabilistic risk analysis (PRA), is 0.645 person-rem/year for WCGS. Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, states that a very small population is defined as an increase of 1.0 person-rem per year or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This is consistent with the Nuclear Regulatory Commission (NRC) Final Safety Evaluation for Nuclear Energy Institute (NEI) 94-01 and EPRI Report No. 1009325. Moreover, the risk impact when compared to other severe accident risks is negligible. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

In addition, as documented in NUREG-1493, Performance-Based Containment Leak-Test Program, dated September 1995, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The WCGS Unit 1 Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) activity based, and (2) time based. Activity-based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. The LLRT requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed test interval extensions do not significantly increase the consequences of an accident previously evaluated.

Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated.

Attachment I to WO 20-0029 Page 117 of 122

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment to the WCGSTS 5.5.16, Containment Leakage Rate Testing Program, involves the extension of the WCGS Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plants ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment to WCGS TS 5.5.16, Containment Leakage Rate Testing Program, involves the extension of the WCGS Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components.

This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leaktightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves only the extension of the interval between Type A containment leak rate tests and Type C tests for WCGS. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusions that Types B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME B&PV Code Section Xl and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance

Attachment I to WO 20-0029 Page 118 of 122 criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, WCNOC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, September 1995 (ML003740058)

2. NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 2012 (ML12221A202)

Attachment I to WO 20-0029 Page 119 of 122

3. RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (ML17317A256)

4. RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009 (ML090410014)

5. NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 21, 1995 (ML11327A025)

6. NUREG-1493, Performance-Based Containment Leak-Test Program - Final Report, September 1995 (9510200161)

7. Electric Power Research Institute (EPRI) Topical Report No. 104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, Palo Alto, California, August 1994

8. NEI 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, October 2008 (ML100620847)

9. Letter from NRC (M. J. Maxin) to NEI (J. C. Butler), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (TAC No.

MC9663), dated June 25, 2008 (ML081140105)

10. Letter from NRC (S. Bahadur) to NEI (B. Bradley), Final Safety Evaluation of Nuclear Energy Institute (NEI) Report, 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (TAC No. ME2164), dated June 8, 2012 (ML121030286)

11. EPRI TR-1018243, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: (Formerly TR-1009325, Revision 2-A dated August 2007), October 2008 (ML14024A045)

12. Letter from NRC (W. D. Reckley) to Wolf Creek Nuclear Operating Corporation (N. S.

Carns), Wolf Creek Generating Station - Amendment No. 69 to Facility Operating License No. NPF-42 (TAC No. M85311) dated November 10, 1993, (ML022030519)

13. Letters from NRC (W. D. Reckley) to Wolf Creek Nuclear Operating Corporation (N. S.

Carns), one-time Exemption to 10 CFR 50 Appendix J, Wolf Creek Generating Station -

Attachment I to WO 20-0029 Page 120 of 122 Amendment No. 76 to Facility Operating License No. NPF-42 (TAC No. M88129), dated August 12, 1994 (ML022040380) and Wolf Creek Generating Station: Exemption to 10CFR Part 50, Appendix J (TAC No. M88130), dated August 10, 1994 (ML022040361)

14. Letter from NRC (J. C. Stone) to Wolf Creek Nuclear Operating Corporation (N.S. Carns),

Wolf Creek Generating Station - Amendment No. 97 to Facility Operating License No.

NPF-42 (TAC No. M94316), dated March 1, 1996 (ML022040701)

15. Letter from NRC (M. Khanna) to Wolf Creek Nuclear Operating Corporation (R. A.

Muench), Issuance of Amendment No. 152 and SER, Wolf Creek Generating Station -

Issuance of Amendment RE: Containment Tendon Surveillance Program and Containment Leakage Rate Testing Program (TAC No. MC1068), dated March 17, 2004 (ML040820934) ) and associated TS changes for Amendment No. 152 (ML040830434)

16. Letter from NRC (T. Tran) to Wolf Creek Nuclear Operating Corporation (T.J. Garrett),

Issuance of Renewed Facility Operating License No. NPF-42 for the Wolf Creek Generating Station, Unit 1, dated November 20, 2008 (ML082880647); and NUREG-1915, Safety Evaluation Report Related to the License Renewal of Wolf Creek Generating Station, Unit 1, Docket No. 50-482, Published October 2008 (ML083090483)

17. Letter from NRC (B. K. Singal) to Luminant Generation Co. (R. Flores), Comanche Peak Nuclear Power Plant, Units 1 and 2 - Issuance of Amendments Re: Technical Specification Change for Extension of the Integrated Leak Rate Test Frequency from 10 to 15 Years (CAC Nos. MF5621 and MF5622), dated December 30, 2015 (ML15309A073)

18. Letter from NRC (M. Mahoney) to Duke Energy (T. D. Ray), "McGuire Nuclear Station, Units 1 and 2 - Issuance of Amendments to Extend the Containment Type A Leak Rate Test Frequency to 15 Years and Type C Leak Rate Test Frequency to 75 Months (CAC Nos. MF9020 and MF9021; EPID L-2016-LLA-0032)," dated January 31, 2018 (ML18009A842)

19. Containment Liner Corrosion Operating Experience Summary, Technical Letter Report -

Revision 1, by D. S. Dunn, A. L. Pulvirenti, and M. A. Hiser (Office of Nuclear Regulatory Research - NRC), dated August 2, 2011 (ML112070867)

20. Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Rev. 4, Developed for NEI by EPRI and Data Systems and Solutions, November 2001

21. Letter from Wolf Creek Nuclear Operating Corporation (T. J. Garrett) to NRC (Document Control Desk), Docket No. 50-482: Application for Renewed Operating License, (ET 06-0038), dated September 27, 2006 (ML062770301) and associated Wolf Creek Generating

Attachment I to WO 20-0029 Page 121 of 122 Station Applicants Environmental Report; Operating License Renewal Stage (ML062770305)

22. Letter from NRC (M. Orenak) to Southern Nuclear Operating Company, Inc. (C. A.

Gayheart), "Vogtle Electric Generating Plant, Units 1 and 2, Issuance of Amendments to Extend the Containment Type A Leak Rate Test Frequency to 15 Years and Type C Leak Rate Test Frequency to 75 Months (CAC Nos. MG0240 and MG0241; EPID L-2017-LLA-0295)," dated October 29, 2018 (ML18263A039)

23. NRC, NUREG-2122, Glossary of Risk-Related Terms in Support of Risk-Informed Decisionmaking, Washington, DC, November 2013 (ML13311A353)

24. WCNOC-PSA-022, Accident Sequence Quantification, Revision 5, February 2020

25. Enercon Report WCNOCPES029-REPT-001, Revision 0, Wolf Creek Internal Events Probabilistic Risk Assessment Peer Review, September 2019

26. PWR Owners Group Report PWROG-19038-P, Independent Assessment of Facts &

Observations Closure of the Wolf Creek Probabilistic Risk Assessment, Revision 0, March 25, 2020

27. NEI 05-04, Revision 3, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Nuclear Energy Institute, November 2009

28. NRC Memorandum to NRR (S. L. Rosenberg), "US Nuclear Regulatory Commission Staff Expectations for an Industry Facts and Observations Independent Assessment Process,"

dated May 1st, 2017 (ML17121A271)

29. American Nuclear Society, ANSI/ANS 56.8-2002, Containment System Leakage Testing Requirements, LaGrange Park, Illinois, November 2002

30. ASME Boiler & Pressure Vessel Code,Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, Division 1, Subsection IWE, Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants

31. ASME B&PV Code,Section XI, Division 1, Subsection IWL, Requirements for Class CC Concrete Components of Light-Water Cooled Plants

32. Letter from Constellation Nuclear (C. H. Cruse) to NRC (Document Control Desk), Calvert Cliffs Nuclear Power Plant, Unit No. 1; Docket No. 50-317 - Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, dated March 27, 2002 (ML020920100)

Attachment I to WO 20-0029 Page 122 of 122

33. Letter from Wolf Creek Nuclear Operating Corporation to the NRC, Wolf Creek Generating Station - Staff Assessment of Flooding Focused Evaluation (CAC No.

MF9964; EPID L-2017-JLD-0019), dated November 8, 2017, ML17241A251

34. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA, 1018243, October 2008

35. RG 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, Revision 17, August 2014 (ML13339A689)

36. Letter from Entergy Operations, Inc. (K. Mulligan) to NRC (Document Control Desk),

Grand Gulf Nuclear Station Response to Request for Additional Information Regarding License Amendment Request to Revise Technical Specifications for Containment Leak Rate Testing, Grand Gulf Nuclear Station, Unit 1, Docket No. 50-416, License No. NPF-29, (GNRO-2015/00063), dated October 28, 2015 (ML15302A042)

37. RG 1.177, An Approach for Plant-Specific, Risk-Informed Decision making: Technical Specifications, Revision 1, May 2011 (ML100910008)

38. Wolf Creek Generating Station LER 2020-001-00, Plant Shutdown Due to Inoperable Containment Purge Isolation Valves, report date April 1, 2020

39. Letter from NRC (B. K. Singal) to Wolf Creek Nuclear Operating Corporation (C.

Reasoner), Wolf Creek Generating Station - Amendment No. 223 to Facility Operating License No. NPF-42, dated January 6, 2020 (EPID L-2019-LLA-0036) (ML19311C643)

Attachment II to WO 20-0029 Page 1 of 2 ATTACHMENT II PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP) PAGES

Attachment II to WO 20-0029 Programs and Manuals Page 2 of 2 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.16 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of Nuclear Energy the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Institute (NEI) Appendix J, Option B, as modified by approved exemptions. This Topical Report (TR) program shall be in accordance with the guidelines contained in NEI 94-01, "Industry Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Guideline for Program, dated September 1995, as modified by the following Implementing exceptions:

Performance-Based

1. The visual examination of containment concrete surfaces intended Option of 10 CFR to fulfill the requirements of 10 CFR 50, Appendix J, Option B 50, Appendix J,"

testing, will be performed in accordance with the requirements of Revision 3-A, dated and frequency specified by ASME Section XI Code, Subsection July 2012, and the IWL, except where relief has been authorized by the NRC.

conditions and limitations specified 2. The visual examination of the steel liner plate inside containment in NEI 94-01, intended to fulfill the requirements of 10 CFR 50, Appendix J, Revision 2-A, dated Option B testing, will be performed in accordance with the October 2008 requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

b. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48 psig.

c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests; (continued)

Wolf Creek - Unit 1 5.0-20 Amendment No. 123, 142, 152, 164

Attachment III to WO 20-0029 Page 1 of 2 ATTACHMENT III RETYPED TECHNICAL SPECIFICATION PAGES

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.5.16 Containment Leakage Rate Testing Program

a. A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) Topical Report (TR) NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J,"

Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as modified by the following exceptions:

1. The visual examination of containment concrete surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWL, except where relief has been authorized by the NRC.

2. The visual examination of the steel liner plate inside containment intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified by ASME Section XI Code, Subsection IWE, except where relief has been authorized by the NRC.

b. The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 48 psig.

c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.20% of containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests and 0.75 La for Type A tests; (continued)

Wolf Creek - Unit 1 5.0-20 Amendment No. 123, 142, 152, 164

ENCLOSURE TO WO 20-0029 EVALUATION OF RISK SIGNIFICANCE OF PERMANENT ILRT EXTENSION (45 pages)

Wolf Creek Generating Station:

Evaluation of Risk Significance of Permanent ILRT Extension 54016-CALC-01 Prepared for:

Project Number: 1MRJ54016 Project

Title:

Permanent ILRT Extension Revision: 1 Name and Date Preparer: Justin Sattler Reviewer: Kelly Wright Review Method Design Review Alternate Calculation Approved by: Matt Johnson Revision 1 Page 1 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension REVISION RECORD

SUMMARY

Revision Revision Summary 0 Initial Issue Revised Appendix A to include discussion of the focused scope peer review and the disposition 1

of F&O AS-B3-01 for impact on the ILRT extension application.

Revision 1 Page 2 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension TABLE OF CONTENTS 1.0 PURPOSE ...................................................................................................................... 4 2.0 SCOPE........................................................................................................................... 4

3.0 REFERENCES

............................................................................................................... 6 4.0 ASSUMPTIONS AND LIMITATIONS .............................................................................. 9 5.0 METHODOLOGY AND ANALYSIS ...............................................................................10 5.1 Inputs ..........................................................................................................................10 5.1.1 General Resources Available ...............................................................................10 5.1.2 Plant Specific Inputs ............................................................................................13 5.1.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large) ..............................................................................................14 5.2 Analysis ......................................................................................................................15 5.2.1 Step 1 - Quantify the Baseline Risk in Terms of Frequency per Reactor Year .....16 5.2.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose)................20 5.2.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Years .................................................................................................................22 5.2.4 Step 4 - Determine the Change in Risk in Terms of Internal Events LERF ..........24 5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability

..........................................................................................................................25 5.2.6 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage

..........................................................................................................................26 5.2.7 Impact from External Events Contribution ............................................................29 5.2.7.1 Other External Hazards .............................................................................30 5.2.8 Defense-In-Depth Impact .....................................................................................31 5.3 Sensitivities .................................................................................................................32 5.3.1 Potential Impact from Steel Liner Corrosion Likelihood ........................................32 5.3.2 Expert Elicitation Sensitivity .................................................................................34 6.0 RESULTS ......................................................................................................................36

7.0 CONCLUSION

S AND RECOMMENDATIONS ..............................................................37 A. PRA Acceptability ..........................................................................................................39 Revision 1 Page 3 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension 1.0 PURPOSE The purpose of this analysis is to provide a risk assessment of permanently extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) from ten years to fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for the Wolf Creek Generating Station (WCGS). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1], the NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals from November 2001

[Reference 3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 Revision 2 as applied to ILRT interval extensions, risk insights in support of a request for a plants licensing basis as outlined in Regulatory Guide (RG) 1.174 [Reference 4], the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval [Reference 5], and the methodology used in EPRI 1018243, Revision 2-A of EPRI 1009325 [Reference 24].

2.0 SCOPE Revisions to 10 CFR 50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing frequency requirement from three in ten years to at least once in ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage rate was less than the limiting containment leakage rate of 1La.

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, September 1995 [Reference 6], provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals

[Reference 2].

The NRC report on performance-based leak testing, NUREG-1493, analyzed the effects of containment leakage on the health and safety of the public and the benefits realized from the containment leak rate testing. In that analysis, it was determined that for a representative PWR plant (i.e., Surry), containment isolation failures contribute less than 0.1% to the latent risks from reactor accidents. Consequently, it is desirable to show that extending the ILRT interval will not lead to a substantial increase in risk from containment isolation failures for WCGS.

NEI 94-01 Revision 3-A supports using EPRI Report No. 1009325 Revision 2-A (EPRI 1018243), Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, for performing risk impact assessments in support of ILRT extensions [Reference 24]. The Guidance provided in Appendix H of EPRI Report No. 1009325 Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI TR-104285 [Reference 2]. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

Revision 1 Page 4 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 requires that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements are not changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.

The acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 10-6 per reactor year. RG 1.174 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the Conditional Containment Failure Probability (CCFP), which helps ensure the defense-in-depth philosophy is maintained, is also calculated.

Regarding CCFP, changes of up to 1.1% have been accepted by the NRC for the one-time requests for extension of ILRT intervals. In context, it is noted that a CCFP of 1/10 (10%) has been approved for application to evolutionary light water designs. Given these perspectives, a change in the CCFP of up to 1.5% is assumed to be small [Reference 1].

In addition, the total annual risk (person-rem/yr population dose) is examined to demonstrate the relative change in this parameter. While no acceptance guidelines for these additional figures of merit are published, examinations of NUREG-1493 [Reference 6] and Safety Evaluations (SEs) for one-time interval extension (summarized in Appendix G of Reference 24) indicate a range of incremental increases in population dose that have been accepted by the NRC. The range of incremental population dose increases is from 0.01 to 0.2 person-rem/yr and/or 0.002% to 0.46% of the total accident dose [Reference 24]. The total doses for the spectrum of all accidents (NUREG-1493 [Reference 6], Figure 7-2) result in health effects that are at least two orders of magnitude less than the NRC Safety Goal Risk. Given these perspectives, a small population dose is defined as an increase from the baseline interval (3 tests per 10 years) dose of 1.0 person-rem per year or 1% of the total baseline dose, whichever is less restrictive for the risk impact assessment of the proposed extended ILRT interval [Reference 1].

Revision 1 Page 5 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension

3.0 REFERENCES

The following references were used in this calculation:

1. Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, NEI 94-01, July 2012.

2. Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, EPRI, Palo Alto, CA, EPRI TR-104285, August 1994.

3. Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Revision 4, developed for NEI by EPRI and Data Systems and Solutions, November 2001.

4. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018.

5. Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C. H.

Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, Docket No. 50-317, March 27, 2002.

6. Performance-Based Containment Leak-Test Program, NUREG-1493, September 1995.

7. Evaluation of Severe Accident Risks: Surry Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 3, Revision 1, Part 1, October 1990.

8. Letter from R. J. Barrett (Entergy) to U. S. Nuclear Regulatory Commission, IPN-01-007, January 18, 2001.

9. United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3

- Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB0178), April 17, 2001.

10. Impact of Containment Building Leakage on LWR Accident Risk, Oak Ridge National Laboratory, NUREG/CR-3539, ORNL/TM-8964, April 1984.

11. Reliability Analysis of Containment Isolation Systems, Pacific Northwest Laboratory, NUREG/CR-4220, PNL-5432, June 1985.

12. Technical Findings and Regulatory Analysis for Generic Safety Issue II.E.4.3 Containment Integrity Check, NUREG-1273, April 1988.

13. Review of Light Water Reactor Regulatory Requirements, Pacific Northwest Laboratory, NUREG/CR-4330, PNL-5809, Volume 2, June 1986.

14. Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAM', EPRI, Palo Alto, CA, TR-105189, Final Report, May 1995.

15. Severe Accident Risks: An Assessment for Five U. S. Nuclear Power Plants, NUREG-1150, December 1990.

16. United States Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.

17. WCNOC-PSA-022, Accident Sequence Quantification, Revision 5, February 2020.

18. Performance Improvement Request 97-1055, March 1998.

Revision 1 Page 6 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension

19. Wolf Creek Generating Station, Application for Renewal of Operating License, Appendix F -

Environmental Report, August 2006.

20. Anthony R. Pietrangelo, One-time extensions of containment integrated leak rate test interval - additional information, NEI letter to Administrative Points of Contact, November 30, 2001.

21. Letter from J. A. Hutton (Exelon, Peach Bottom) to U. S. Nuclear Regulatory Commission, Docket No. 50-278, License No. DPR-56, LAR-01-00430, dated May 30, 2001.

22. Risk Assessment for Joseph M. Farley Nuclear Plant Regarding ILRT (Type A) Extension Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, P0293010002-1929-030602, March 2002.

23. Letter from D. E. Young (Florida Power, Crystal River) to U. S. Nuclear Regulatory Commission, 3F0401-11, dated April 25, 2001.

24. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA, 1018243, October 2008.

25. Risk Assessment for Vogtle Electric Generating Plant Regarding the ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, February 2003.

26. Perspectives Gained from the IPEEE Program, USNRC, NUREG-1742, April 2002.

27. Wolf Creek Generating Station, Updated Safety Analysis Report (USAR).

28. Generic Issue 199 (GI-199), Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment, ML100270582, September 2010.

29. The Nuclear Energy Institute - Seismic Risk Evaluations for Plants in the Central and Eastern United States, ML14083A596, March 2014.

30. Surveillance Test Routing Sheet STS PE-018, Wolf Creek Nuclear Operating Corporation, Containment Integrated Leak Rate Test, May 27, 2011.

31. Letter from Wolf Creek Nuclear Operating Corporation to the NRC, Wolf Creek Generating Station - Staff Assessment of Flooding Focused Evaluation (CAC No. MF9964; EPID L-2017-JLD-0019), dated November 8, 2017, ML17241A251.

32. Technical Report TR-95-0015, Individual Plant Examination for External Events Summary Report, July 1995.

33. NEI 05-04, Revision 3, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, Nuclear Energy Institute, November 2009.

34. Technical Letter Report ML112070867, Containment Liner Corrosion Operating Experience Summary, Revision 1, August 2011.

35. Enercon Report WCNOCPES029-REPT-001, Revision 0, Wolf Creek Internal Events Probabilistic Risk Assessment Peer Review, September 2019.

36. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, March 2009.

37. USNRC Memorandum, "US Nuclear Regulatory Commission Staff Expectations for an Industry Facts and Observations Independent Assessment Process," May 1st, 2017 (ADAMS Access ML17121A271).

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension

38. USNRC Memorandum, United States Nuclear Regulatory Commission Audit Report on Observation of Industry Independent Assessment Team Close-Out of Facts and Observations (F&Os), May 1st, 2017 (ADAMS Access ML17095A252).

39. PWR Owners Group Report PWROG-19038-P, Independent Assessment of Facts &

Observations Closure of the Wolf Creek Probabilistic Risk Assessment, Revision 0, March 2020.

40. WCNOC-PSA-035, Internal Flooding PRA Modeling and Quantification Notebook, Revision 3, February 2020.

41. WCNOC-PSA-001, At-Power Internal Events PRA, Initiating Events Analysis, Revision 6, February 2020.

Revision 1 Page 8 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension 4.0 ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculation:

The acceptability (i.e., technical adequacy) of the WCGS PRA [Reference 17] is consistent with the requirements of Regulatory Guide 1.200, as detailed in Appendix A.

The WCGS Level 1 and 2 internal events PRA models [Reference 17] provide representative results.

It is appropriate to use the WCGS internal events PRA model to effectively describe the risk change attributable to the ILRT extension. An analysis is performed in Section 5.2.7 to show the effect of including external event models for the ILRT extension. The Seismic risk from GI-199 [Reference 28] and the updated IPEEE Fire analysis

[Reference 18] are used for this analysis.

Accident classes describing radionuclide release end states are defined consistent with EPRI methodology [Reference 24].

The representative containment leakage for Class 1 sequences is 1La. Class 3 accounts for increased leakage due to Type A inspection failures [Reference 24].

The representative containment leakage for Class 3a sequences is 10La based on the previously approved methodology performed for Indian Point Unit 3 [Reference 8, Reference 9].

The representative containment leakage for Class 3b sequences is 100La based on the guidance provided in EPRI Report No. 1009325, Revision 2-A (EPRI 1018243)

[Reference 24].

The Class 3b can be very conservatively categorized as LERF based on the previously approved methodology [Reference 8, Reference 9].

The impact on population doses from containment bypass scenarios is not altered by the proposed ILRT extension but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes in the conclusions from this analysis will result from this separate categorization.

The reduction in ILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal [Reference 24].

While precise numbers are maintained throughout the calculations, some values have been rounded when presented in this report. Therefore, summing individual values within tables may yield a different result than the sum result shown in the tables.

Revision 1 Page 9 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension 5.0 METHODOLOGY AND ANALYSIS 5.1 Inputs This section summarizes the general resources available as input (Section 5.1.1) and the plant specific resources required (Section 5.1.2).

5.1.1 General Resources Available Various industry studies on containment leakage risk assessment are briefly summarized here:

1. NUREG/CR-3539 [Reference 10]

2. NUREG/CR-4220 [Reference 11]

3. NUREG-1273 [Reference 12]

4. NUREG/CR-4330 [Reference 13]

5. EPRI TR-105189 [Reference 14]

6. NUREG-1493 [Reference 6]

7. EPRI TR-104285 [Reference 2]

8. NUREG-1150 [Reference 15] and NUREG/CR-4551 [Reference 7]

9. NEI Interim Guidance [Reference 3, Reference 20]

10. Calvert Cliffs liner corrosion analysis [Reference 5]

11. EPRI Report No. 1009325, Revision 2-A (EPRI 1018243), Appendix H [Reference 24]

This first study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and is to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different containment leakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRCs cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh study is an EPRI study of the impact of extending ILRT and local leak rate test (LLRT) intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the basis for the consequence analysis of the ILRT interval extension for WCGS. The ninth study includes the NEI recommended methodology (promulgated in two letters) for evaluating the risk associated with obtaining a one-time extension of the ILRT interval. The tenth study addresses the impact of age-related degradation of the containment liners on ILRT evaluations. Finally, the eleventh study builds on the previous work and includes a recommended methodology and template for evaluating the risk associated with a permanent 15-year extension of the ILRT interval.

NUREG/CR-3539 [Reference 10]

Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leak rates on public risk in NUREG/CR-3539. This study uses information from WASH-1400

[Reference 16] as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

NUREG/CR-4220 [Reference 11]

NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985.

The study reviewed over two thousand license event reports (LERs), ILRT reports and other Revision 1 Page 10 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension related records to calculate the unavailability of containment due to leakage.

NUREG-1273 [Reference 12]

A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREG/CR-4220 database. This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected. In addition, this study noted that local leak rate tests can detect essentially all potential degradations of the containment isolation system.

NUREG/CR-4330 [Reference 13]

NUREG/CR-4330 is a study that examined the risk impacts associated with increasing the allowable containment leakage rates. The details of this report have no direct impact on the modeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals.

However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies:

the effect of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of containment.

EPRI TR-105189 [Reference 14]

The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessment because it provides insight regarding the impact of containment testing on shutdown risk. This study contains a quantitative evaluation (using the EPRI ORAM software) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk. The conclusion from the study is that a small, but measurable, safety benefit is realized from extending the test intervals.

NUREG-1493 [Reference 6]

NUREG-1493 is the NRCs cost-benefit analysis for proposed alternatives to reduce containment leakage testing intervals and/or relax allowable leakage rates. The NRC conclusions are consistent with other similar containment leakage risk studies:

Reduction in ILRT frequency from 3 per 10 years to 1 per 20 years results in an imperceptible increase in risk.

Given the insensitivity of risk to the containment leak rate and the small fraction of leak paths detected solely by Type A testing, increasing the interval between integrated leak rate tests is possible with minimal impact on public risk.

EPRI TR-104285 [Reference 2]

Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study),

the EPRI TR-104285 study is a quantitative evaluation of the impact of extending ILRT and LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with NUREG-1150 Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-104285 uses a simplified Containment Event Tree to subdivide representative core damage frequencies into eight classes of containment response to a core damage accident:

1. Containment intact and isolated

2. Containment isolation failures dependent upon the core damage accident

3. Type A (ILRT) related containment isolation failures Revision 1 Page 11 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension

4. Type B (LLRT) related containment isolation failures

5. Type C (LLRT) related containment isolation failures

6. Other penetration related containment isolation failures

7. Containment failures due to core damage accident phenomena

8. Containment bypass Consistent with the other containment leakage risk assessment studies, this study concluded:

the proposed CLRT (Containment Leak Rate Tests) frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms. For example, for the PWR analyzed, the change is about 0.02 person-rem per year NUREG-1150 [Reference 15] and NUREG/CR-4551 [Reference 7]

NUREG-1150 and the technical basis, NUREG/CR-4551 [Reference 7], provide an ex-plant consequence analysis for a spectrum of accidents including a severe accident with the containment remaining intact (i.e., Tech Spec Leakage). This ex-plant consequence analysis is calculated for the 50-mile radial area surrounding Surry. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the WCGS Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent WCGS. (The meteorology and site differences other than population are assumed not to play a significant role in this evaluation.)

NEI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals [Reference 3, Reference 20]

The guidance provided in this document builds on the EPRI risk impact assessment methodology [Reference 2] and the NRC performance-based containment leakage test program

[Reference 6], and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

Calvert Cliffs Response to Request for Additional Information Concerning the License Amendment for a One-Time Integrated Leakage Rate Test Extension [Reference 5]

This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting liner corrosion, and the corresponding change in risk. The methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms was factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner.

EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals [Reference 24]

This report provides a generally applicable assessment of the risk involved in extension of ILRT test intervals to permanent 15-year intervals. Appendix H of this document provides guidance for performing plant-specific supplemental risk impact assessments and builds on the previous EPRI risk impact assessment methodology [Reference 2] and the NRC performance-based containment leakage test program [Reference 6], and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

The approach included in this guidance document is used in the WCGS assessment to determine the estimated increase in risk associated with the ILRT extension. This document Revision 1 Page 12 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension includes the bases for the values assigned in determining the probability of leakage for the EPRI Class 3a and 3b scenarios in this analysis, as described in Section 5.2.

5.1.2 Plant Specific Inputs The plant-specific information used to perform the WCGS ILRT Extension Risk Assessment includes the following:

CDF and LERF Model results [Reference 17, Reference 18, Reference 40]

Dose within a 50-mile radius [Reference 19]

ILRT results to demonstrate adequacy of the administrative and hardware issues

[References 30 and 31]

WCGS Model The Internal Events (IE) and Internal Flood (IF) PRA Models that are used for WCGS are characteristic of the as-built plant. The current CDF and LERF model is a linked fault tree model

[Reference 17]. The IE+IF CDF is 1.63E-5/yr; the LERF is 1.10E-7/yr [References 17 and 40].

Table 5-1 and Table 5-2 provide a summary of the IE+IF CDF and LERF results for the WCGS PRA Model.

The total Fire CDF is 8.14E-6/yr [Reference 18]. The seismic CDF is taken from GI-199

[Reference 28]. Other external event risk is screened in Reference 32. Refer to Section 5.2.7 for further details on external events as they pertain to this analysis.

Table 5 Internal Events CDF Internal Events Frequency (per year)

Internal Floods 9.16E-6 Transients 5.37E-6 FLB/SLB 7.66E-7 LOCAs 5.62E-7 LOOP 1.41E-7 SGTR 2.49E-7 ISLOCA 3.29E-8 Total Internal Events CDF 1.63E-5 Revision 1 Page 13 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Internal Events LERF Internal Events Frequency (per year)

Internal Floods 3.65E-8 Transients 8.21E-8 ISLOCA 3.28E-8 FLB/SLB 3.06E-9 SGTR 2.84E-8 LOCAs 3.30E-10 LOOP 2.66E-10 Total Internal Events LERF 1.10E-7 Release Category Definitions Table 5-3 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology [Reference 24]. These containment failure classifications are used in this analysis to determine the risk impact of extending the Containment Type A test interval, as described in Section 5.2 of this report.

Table 5 EPRI Containment Failure Classification [Reference 24]

Class Description Containment remains intact including accident sequences that do not lead to containment failure in the 1 long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant.

Containment isolation failures (as reported in the Individual Plant Examinations) including those accidents 2

in which there is a failure to isolate the containment.

Independent (or random) isolation failures include those accidents in which the pre-existing isolation 3

failure to seal (i.e., provide a leak-tight containment) is not dependent on the sequence in progress.

Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 3 isolation 4

failures, but is applicable to sequences involving Type B tests and their potential failures. These are the Type B-tested components that have isolated, but exhibit excessive leakage.

Independent (or random) isolation failures including those accidents in which the pre-existing isolation 5 failure to seal is not dependent on the sequence in progress. This class is similar to Class 4 isolation failures, but is applicable to sequences involving Type C test and their potential failures.

Containment isolation failures including those leak paths covered in the plant test and maintenance 6

requirements or verified per in-service inspection and testing (ISI/IST) program.

Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J 7

testing requirements do not impact these accidents.

Accidents in which the containment is bypassed (either as an initial condition or induced by phenomena) 8 are included in Class 8. Changes in Appendix J testing requirements do not impact these accidents.

5.1.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large)

The ILRT can detect a number of component failures such as liner breach, failure of certain bellows arrangements, and failure of some sealing surfaces, which can lead to leakage. The proposed ILRT test interval extension may influence the conditional probability of detecting Revision 1 Page 14 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension these types of failures. To ensure that this effect is properly addressed, the EPRI Class 3 accident class, as defined in Table 5-3, is divided into two sub-classes, Class 3a and Class 3b, representing small and large leakage failures respectively.

The probability of the EPRI Class 3a and Class 3b failures is determined consistent with the EPRI Guidance [Reference 24]. For Class 3a, the probability is based on the maximum likelihood estimate of failure (arithmetic average) from the available data (i.e., 2 small failures in 217 tests leads to 2 / 217 = 0.0092). For Class 3b, the probability is based on the Jeffreys non-informative prior for no large failures in 217 tests (i.e., 0.5 / (217+1) = 0.0023).

In a follow-up letter [Reference 20] to their ILRT guidance document [Reference 3], NEI issued additional information concerning the potential that the calculated delta LERF values for several plants may fall above the very small change guidelines of the NRC Regulatory Guide 1.174

[Reference 4]. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for LERF. NEI describes ways to demonstrate that, using plant-specific calculations, the LERF is smaller than that calculated by the simplified method.

The supplemental information states:

The methodology employed for determining LERF (Class 3b frequency) involves conservatively multiplying the CDF by the failure probability for this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, some plant-specific accident classes leading to core damage are likely to include individual sequences that either may already (independently) cause a LERF or could never cause a LERF, and are thus not associated with a postulated large Type A containment leakage path (LERF). These contributors can be removed from Class 3b in the evaluation of LERF by multiplying the Class 3b probability by only that portion of CDF that may be impacted by Type A leakage.

The application of this additional guidance to the analysis for WCGS, as detailed in Section 5.2, involves subtracting LERF risk from the CDF that is applied to Class 3b because this portion of LERF is unaffected by containment integrity. To be consistent, the same change is made to the Class 3a CDF, even though these events are not considered LERF.

Consistent with the NEI Guidance [Reference 3], the change in the leak detection probability can be estimated by comparing the average time that a leak could exist without detection. For example, the average time that a leak could go undetected with a three-year test interval is 1.5 years (3 years / 2), and the average time that a leak could exist without detection for a ten-year interval is 5 years (10 years / 2). This change would lead to a non-detection probability that is a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRT testing.

Correspondingly, an extension of the ILRT interval to 15 years can be estimated to lead to a factor of 5 ((15/2)/1.5) increase in the non-detection probability of a leak.

It should be noted that using the methodology discussed above is very conservative compared to previous submittals (e.g., the IP3 request for a one-time ILRT extension that was approved by the NRC [Reference 9]) because it does not factor in the possibility that the failures could be detected by other tests (e.g., the Type B local leak rate tests that will still occur). Eliminating this possibility conservatively over-estimates the factor increases attributable to the ILRT extension.

5.2 Analysis The application of the approach based on the guidance contained in EPRI 1009325 [Reference 24] and previous risk assessment submittals on this subject [References 5, 8, 21, 22, and 23]

have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report, as described in Table 5-4.

Revision 1 Page 15 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension The analysis performed examined WCGS-specific accident sequences in which the containment remains intact or the containment is impaired. Specifically, the breakdown of the severe accidents, contributing to risk, was considered in the following manner:

Core damage sequences in which the containment remains intact initially and in the long term (EPRI 1009325, Class 1 sequences [Reference 24]).

Core damage sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B or Type C test components. For example, liner breach or bellow leakage (EPRI 1009325, Class 3 sequences [Reference 24]).

Accident sequences involving containment bypassed (EPRI 1009325, Class 8 sequences [Reference 24]), large containment isolation failures (EPRI 1009325, Class 2 sequences [Reference 24]), and small containment isolation failure-to-seal events (EPRI 1009325, Class 4 and 5 sequences [Reference 24]) are accounted for in this evaluation as part of the baseline risk profile. However, they are not affected by the ILRT frequency change.

Class 4 and 5 sequences are impacted by changes in Type B and C test intervals; therefore, changes in the Type A test interval do not impact these sequences.

Table 5 EPRI Accident Class Definitions Accident Classes (Containment Release Type) Description 1 No Containment Failure 2 Large Isolation Failures (Failure to Close) 3a Small Isolation Failures (Liner Breach) 3b Large Isolation Failures (Liner Breach) 4 Small Isolation Failures (Failure to Seal - Type B) 5 Small Isolation Failures (Failure to Seal - Type C) 6 Other Isolation Failures (e.g., Dependent Failures) 7 Failures Induced by Phenomena (Early and Late) 8 Bypass (SGTR and Interfacing System LOCA)

CDF All CET End States (Including Very Low and No Release)

The steps taken to perform this risk assessment evaluation are as follows:

Step 1 - Quantify the baseline risk in terms of frequency per reactor year for each of the accident classes presented in Table 5-4.

Step 2 - Develop plant-specific person-rem dose (population dose) per reactor year for each of the eight accident classes.

Step 3 - Evaluate risk impact of extending Type A test interval from 3 in 10 years to 1 in 15 years and 1 in 10 years to 1 in 15 years.

Step 4 - Determine the change in risk in terms of Large Early Release Frequency (LERF) in accordance with RG 1.174 [Reference 4].

Step 5 - Determine the impact on the Conditional Containment Failure Probability (CCFP).

5.2.1 Step 1 - Quantify the Baseline Risk in Terms of Frequency per Reactor Year As previously described, the extension of the Type A interval does not influence those accident progressions that involve large containment isolation failures, Type B or Type C testing, or Revision 1 Page 16 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension containment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks is included in the model (these events are represented by the Class 3 sequences in EPRI 1009325 [Reference 24]). The question on containment integrity was modified to include the probability of a liner breach or bellows failure (due to excessive leakage) at the time of core damage. Two failure modes were considered for the Class 3 sequences. These are Class 3a (small breach) and Class 3b (large breach).

The frequencies for the severe accident classes defined in Table 5-4 were developed for WCGS by first determining the frequencies for Classes 1, 2, 6, 7, and 8.

Revision 1 Page 17 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-5 presents the grouping of each release category in EPRI Classes based on the associated description. Table 5-6 provides a summary of the accident sequence frequencies that can lead to radionuclide release to the public and have been derived consistent with the NEI Interim Guidance [Reference 3] and the definitions of accident classes and guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24]. Adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion of the steel liner per the methodology described in Section 5.2.6.

Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (e.g., containment liner) exists that can only be detected by performing a Type A ILRT. The probability of leakage detectable by a Type A ILRT is calculated to determine the impact of extending the testing interval. The Class 3 calculation is divided into two classes: Class 3a is defined as a small liner breach (La < leakage

< 10La), and Class 3b is defined as a large liner breach (10La < leakage < 100La).

Data reported in EPRI 1009325, Revision 2-A [Reference 24] states that two events could have been detected only during the performance of an ILRT and thus impact risk due to change in ILRT frequency. There was a total of 217 successful ILRTs during this data collection period.

Therefore, the probability of leakage is determined for Class 3a as shown in the following equation:

2

= = 0.0092 217 Multiplying the CDF by the probability of a Class 3a leak yields the Class 3a frequency contribution in accordance with guidance provided in Reference 24. As described in Section 5.1.3, additional consideration is made to not apply failure probabilities on those cases that are already LERF scenarios. Therefore, these LERF contributions from CDF are removed. The frequency of a Class 3a failure is calculated by the following equation:

= = 1.63E 1.10E-7 = 1.49E-7 In the database of 217 ILRTs, there are zero containment leakage events that could result in a large early release. Therefore, the Jeffreys non-informative prior is used to estimate a failure rate and is illustrated in the following equations:

5678 9: ;<=6 > + 1/2 Jeffreys Failure Probability =

5678 9: A >B> + 1 0 + 1/2 C = = 0.0023 217 + 1 The frequency of a Class 3b failure is calculated by the following equation:

.D C = C = E 1.63E 1.10E-7 = 3.71E-8 For this analysis, the associated containment leakage for Class 3a is 10La and for Class 3b is 100La. These assignments are consistent with the guidance provided in Reference 24.

Class 1 Sequences. This group consists of all core damage accident progression bins for which the containment remains intact (modeled as Technical Specification Leakage). The SAMA

[Reference 19] provides the most recent plant-specific risk profile; since the model [Reference 17] does not calculate Intact frequency, the SAMA Intact frequency is scaled using the CDF, calculated below:

GHI I = JKB;LBMNON / MNON GP = 2.80E-5 / 3.16E-5

• 1.63E-5 = 1.44E-5 Revision 1 Page 18 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension The Intact frequency for internal events is 1.44E-5. The EPRI Accident Class 1 frequency is then adjusted by subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF),

calculated below:

= GHI I C = 1.44E 1.49E 3.71E-8 = 1.42E-5 Class 2 Sequences. This group consists of accident progression bins with large containment isolation failures. The large isolation failure is in internal events cutsets that contribute 0.269 of LERF. Multiplying by the total LERF, the EPRI Accident Class 2 frequency is 2.95E-8, as shown in Table 5-5.

Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis, consistent with approved methodology.

Class 5 Sequences. This group consists of all core damage accident progression bins for which a containment isolation failure-to-seal of Type C test components occurs. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis, consistent with approved methodology.

Class 6 Sequences. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs.

These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution. All other failure modes are bounded by the Class 2 assumptions.

This accident class is also not evaluated further.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure is induced by severe accident phenomena (e.g., overpressure). This frequency is calculated by subtracting the Class 1, 2, and 8 frequencies from the total CDF. For this analysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table 5-5.

Class 8 Sequences. This group consists of all core damage accident progression bins in which containment is bypassed via SGTR or ISLOCA. The SGTR internal events (including induced SGTR) cutsets contribute 0.260 of LERF. The ISLOCA initiators are in internal events cutsets that contribute 0.299 of LERF. Thus, the total EPRI Accident Class 8 frequency is the summation of the SGTR and ISLOCA frequencies, 6.12E-8, as shown in Table 5-5 and Table 5-6.

Revision 1 Page 19 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Accident Class Frequencies (Core Damage)

EPRI Category Unit 1 Frequency (/yr)

Class 1 1.44E-5 Class 2 2.95E-8 Class 7 1.76E-6 Class 8 (SGTR) 2.84E-8 Class 8 (ISLOCA) 3.28E-8 Total (CDF) 1.63E-5 Table 5 Baseline Risk Profile Class Description Frequency (/yr) 1 No containment failure 1.42E-52 2 Large containment isolation failures 2.95E-8 3a Small isolation failures (liner breach) 1.49E-7 3b Large isolation failures (liner breach) 3.71E-8 4 Small isolation failures - failure to seal (type B) 1 5 Small isolation failures - failure to seal (type C) 1 6 Containment isolation failures (dependent failure, personnel errors) 1 7 Severe accident phenomena induced failure (early and late) 1.76E-6 8 Containment bypass 6.12E-8 Total 1.63E-5

1. represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.

2. The Class 3a and 3b frequencies are subtracted from Class 1 to preserve total CDF.

5.2.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose)

Plant-specific release analyses were performed to estimate the person-rem doses to the population within a 50-mile radius from the plant. The population dose was calculated using data in Table F.3-3 of the Severe Accident Mitigation Alternatives (SAMA) analysis [Reference 19], which is presented in Table 5-7. Reference 19 Leakage/NCF (Intact) Release Category corresponds to EPRI Accident Class 1. CIF (Containment Isolation Failure) Release Category corresponds to EPRI Accident Class 2. Since they are not associated with other classes, two containment end-states correspond to EPRI Accident Class 7 (LCF(K) and ECF Release Categories); the EPRI Accident Class 7 dose is calculated via a weighted average using the frequencies provided in Reference 19. The SGTR Release Category and ISLOCA Release Category correspond to EPRI Accident Class 8; dose used in this analysis is weighted via the ISLOCA and SGTR frequencies in this calculation. Class 3a and 3b population dose values are calculated from the Class 1 population dose and are represented as 10La and 100La, respectively, as guidance in Reference 1 dictates. Since population dose (person-rem) is not presented directly in SAMA Table F.3-3, population dose rate (person-rem/yr) is divided by frequency (/yr) to calculate population dose. This methodology yields a population dose for each release category except CIF because its reported population dose rate is 0.00, which does not allow for a calculation of population dose; therefore, Class 2 dose is calculated via a different methodology detailed below.

Revision 1 Page 20 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Baseline Population Doses Leakage/

Release Category LCF(K) SGTR ISLOCA CIF ECF Total NCF Frequency (/yr) 2.80E-05 1.13E-06 1.65E-07 1.92E-06 3.42E-09 4.48E-07 3.16E-05 Population Dose 0.09 0.04 0.04 2.55 0.00 0.14 2.86 Rate (person-rem/yr)

Population Dose 3.21E+3 3.54E+4 2.42E+5 1.33E+6 N/A1 3.13E+5 9.05E+4 (person-rem)

EPRI Class 1 7 8 8 2 7

1. Since the reported population dose rate is 0.00, a population dose is not calculated using SAMA data.

The population dose for Class 2 is calculated using the methodology of scaling Surry population doses to WCGS [Reference 7]. The adjustment factor for reactor power level (AFpower) is defined as the ratio of the power level at WCGS (PLWC) [Reference 19] to that at Surry (PLS)

[Reference 7]. This adjustment factor is calculated as follows:

AFpower = PLWC / PLS = 3565 / 2441 = 1.460 The adjustment factor for population (AFPopulation) is defined as the ratio of the population within 50-mile radius of WC (POPWC) [Reference 19] to that of Surry (POPS) [Reference 7]. The 2040 population surrounding WCGS was projected as 241,803 [Reference 32]. This adjustment factor is calculated as follows:

AFPopulation = POPWC / POPS = 241803 / 1231275 = 0.196 The adjustment factor (AF) is calculated by combining the factors as follows:

AF = AFpower

• AFPopulation = 1.460

• 0.196 = 0.287 The population dose data in NUREG/CR-4551 for Surry [Reference 7] is reported in ten distinct collapsed accident progression bins (CAPBs). For this ILRT extension application, CAPB2 is categorized in EPRI Accident Class 2. Based on the above adjustment factors and the 50-mile population dose (person-rem) for each CAPB considered in the NUREG/CR-4551 Surry study, the WC population doses (WCPD) for Class 2 is calculated as follows:

WCPDClass2 = AF

• PDCAPB2 = 0.287

• 6.46E+5 = 1.85E+5 Table 5-8 presents dose exposures calculated from the methodology described in Reference

24. Table 5-9 presents the baseline risk profile for WCGS.

The population dose for EPRI Accident Classes 3a and 3b were calculated based on the guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24] and the Class 1 dose as follows:

J =;>> 3; 9]6=;B<9K 9> = 10 3.21 +3 = 3.21 +4 J =;>> 38 9]6=;B<9K 9> = 100 3.21 +3 = 3.21 +5 Revision 1 Page 21 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Baseline Population Doses Class Description Population Dose (person-rem) 1 No containment failure 3.21E+03 2 Large containment isolation failures 1.85E+05 3a Small isolation failures (liner breach) 3.21E+041 3b Large isolation failures (liner breach) 3.21E+052 4 Small isolation failures - failure to seal (type B) N/A 5 Small isolation failures - failure to seal (type C) N/A 6 Containment isolation failures (dependent failure, personnel errors) N/A 7 Severe accident phenomena induced failure (early and late) 1.14E+05 8 Containment bypass 8.24E+05

1. 10*La

2. 100*La Table 5 Baseline Risk Profile for ILRT Class Description Frequency Contribution Population Population

(/yr) (%) Dose (person- Dose Rate rem) (person-rem/yr) 1 No containment failure2 1.42E-05 87.46% 3.21E+03 4.58E-02 2 Large containment isolation failures 2.95E-08 0.18% 1.85E+05 5.46E-03 Small isolation failures (liner 3a 1.49E-07 0.92% 3.21E+04 4.79E-03 breach)

Large isolation failures (liner 3b 3.71E-08 0.23% 3.21E+05 1.19E-02 breach)

Small isolation failures - failure to 4 1 1 1 1 seal (type B)

Small isolation failures - failure to 5 1 1 1 1 seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1 1 1 1 errors)

Severe accident phenomena 7 1.76E-06 10.84% 1.14E+05 2.01E-01 induced failure (early and late) 8 Containment bypass 6.12E-08 0.38% 8.24E+05 5.04E-02 Total 1.63E-05 100.00% 3.20E-01

1. represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.

2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b to preserve total CDF.

5.2.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Years The next step is to evaluate the risk impact of extending the test interval from its current 10-year interval to a 15-year interval. To do this, an evaluation must first be made of the risk associated with the 10-year interval, since the base case applies to a 3-year interval (i.e., a simplified representation of a 3-in-10 interval).

Revision 1 Page 22 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Risk Impact Due to 10-Year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and Class 3b sequences is impacted. The risk contribution is changed based on the NEI guidance as described in Section 5.1.3 by a factor of 10/3 compared to the base case values. The Class 3a and 3b frequencies are calculated as follows:

^ _`a = = 1.62E-5 = 4.97E-7

_ .D _ .D

^ C _`a = E

= E 1.62E-5 = 1.24E-7 The results of the calculation for a 10-year interval are presented in Table 5-10.

Table 5 Risk Profile for Once in 10 Year ILRT Class Description Frequency Contribution Population Population

(/yr) (%) Dose (person- Dose Rate rem) (person-rem/yr) 1 No containment failure2 1.38E-05 84.80% 3.21E+03 4.44E-02 2 Large containment isolation failures 2.95E-08 0.18% 1.85E+05 5.46E-03 Small isolation failures (liner 3a 4.97E-07 3.05% 3.21E+04 1.60E-02 breach)

Large isolation failures (liner 3b 1.24E-07 0.76% 3.21E+05 3.97E-02 breach)

Small isolation failures - failure to 4 1 1 1 1 seal (type B)

Small isolation failures - failure to 5 1 1 1 1 seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1 1 1 1 errors)

Severe accident phenomena 7 1.76E-06 10.84% 1.14E+05 2.01E-01 induced failure (early and late) 8 Containment bypass 6.12E-08 0.38% 8.24E+05 5.04E-02 Total 1.63E-05 3.57E-01

1. represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.

2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b to preserve total CDF.

Risk Impact Due to 15-Year Test Interval The risk contribution for a 15-year interval is calculated in a manner similar to the 10-year interval. The difference is in the increase in probability of leakage in Classes 3a and 3b. For this case, the value used in the analysis is a factor of 5 compared to the 3-year interval value, as described in Section 5.1.3. The Class 3a and 3b frequencies are calculated as follows:

D

^ D`a = =5 1.62E-5 = 7.45E-7 D .D .D

^ C D`a = =5 1.62E-5 = 1.85E-7 E E Revision 1 Page 23 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension The results of the calculation for a 15-year interval are presented in Table 5-11.

Table 5 Risk Profile for Once in 15 Year ILRT Class Description Frequency Contribution Population Dose Population

(/yr) (%) (person-rem) Dose Rate (person-rem/yr) 1 No containment failure2 1.35E-05 82.89% 3.21E+03 4.34E-02 Large containment isolation 2 2.95E-08 0.18% 1.85E+05 5.46E-03 failures Small isolation failures (liner 3a 7.45E-07 4.58% 3.21E+04 2.40E-02 breach)

Large isolation failures (liner 3b 1.85E-07 1.14% 3.21E+05 5.96E-02 breach)

Small isolation failures - failure 4 1 1 1 1 to seal (type B)

Small isolation failures - failure 5 1 1 1 1 to seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1 1 1 1 errors)

Severe accident phenomena 7 1.76E-06 10.84% 1.14E+05 2.01E-01 induced failure (early and late) 8 Containment bypass 6.12E-08 0.38% 8.24E+05 5.04E-02 Total 1.63E-05 3.84E-01

1. represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.

2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b to preserve total CDF.

5.2.4 Step 4 - Determine the Change in Risk in Terms of Internal Events LERF The risk increase associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from an intact containment could, in fact, result in a larger release due to the increase in probability of failure to detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.

Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 [Reference 4] defines very small changes in risk as resulting in increases of CDF less than 10-6/yr and increases in LERF less than 10-7/yr, and small changes in LERF as less than 10-6/yr. Since containment overpressure is not required in support of ECCS performance to mitigate design basis accidents and no equipment in the shield building is credited in the CDF model at WCGS, the ILRT extension does not impact CDF. Therefore, the relevant risk-impact metric is LERF.

For WCGS, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a 10-year test interval from Table 5-10, the Class 3b frequency is 1.24E-07/yr; based on a 15-year test interval from Table 5-11, the Class 3b frequency is 1.85E-07/yr. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is 1.48E-7/yr. Similarly, the increase due to increasing the interval from 10 to 15 years is 6.18E-8/yr. Table 5-12 summarizes these results.

Revision 1 Page 24 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Years 15 Years Class 3b (Type A LERF) 3.71E-08 1.24E-07 1.85E-07 LERF (3 year baseline) 8.65E-08 1.48E-07 LERF (10 year baseline) 6.18E-08 As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology), the estimated change in LERF meets the criteria for a very small change when comparing the 15-year results to the current 10-year requirement, as it remains less than 1.0E-7/yr, and meets the criteria for a small change when comparing the 15-year results to the original 3-year requirement, as it exceeds 1.0E-7/yr and remains less than a 1.0E-6/yr change in LERF. For this small change in LERF to be acceptable, total LERF must be less than 1.0E-5/yr. The total LERF is calculated below by adding the model IE+IF LERF and change in LERF shown in Table 5-12:

LERF = LERFinternal + LERFclass3Bincrease LERF = 1.10E-7/yr + 1.48E-7/yr = 2.58E-7/yr As specified in Regulatory Guide 1.174 [Reference 4], since the total LERF is less than 1.0E-05/yr, it is acceptable for the LERF to be between 1.0E-07/yr and 1.0E-06/yr.

NEI 94-01 [Reference 1] states that a small population dose is defined as an increase of 1.0 person-rem/yr, or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. As shown in Table 5-13, the results of this calculation meet the dose rate criteria.

Table 5 Impact on Dose Rate due to Extended Type A Testing Intervals ILRT Inspection Interval 10 Years 15 Years Dose Rate (3 year baseline) 3.760E-02 6.446E-02 Dose Rate (10 year baseline) 2.686E-02

%Dose Rate (3 year baseline) 11.76% 20.17%

%Dose Rate (10 year baseline) 7.519%

1. Dose Rate is the difference in the total dose rate between cases. For instance, Dose Rate (3 year baseline) for the 1 in 15 case is the total dose rate of the 1 in 15 case minus the total dose rate of the 3 in 10 year case.

2. %Dose Rate is the Dose Rate divided by the total baseline dose rate. For instance, %Dose Rate (3 year baseline) for the 1 in 15 case is the Dose Rate (3 year baseline) of the 1 in 15 year case divided by the total dose rate of the 3 in 10 year case.

5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability Another parameter that the NRC guidance in RG 1.174 [Reference 4] states can provide input into the decision-making process is the change in the conditional containment failure probability (CCFP). The CCFP is defined as the probability of containment failure given the occurrence of an accident. This probability can be expressed using the following equation:

KL:

=1 Revision 1 Page 25 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension where f(ncf) is the frequency of those sequences that do not result in containment failure; this frequency is determined by summing the Class 1 and Class 3a results.

Since CCFP is only concerned with a containment failure and not whether the release is small or large, the Class 1 results without containment spray refinement are used to calculate the CCFP. Table 5-14 shows the steps and results of this calculation.

Table 5 Impact on CCFP due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Years 15 Years f(ncf) (/yr) 1.44E-05 1.43E-05 1.42E-05 f(ncf)/CDF 0.884 0.878 0.875 CCFP 0.116 0.122 0.125 CCFP (3 year baseline) 0.532% 0.911%

CCFP (10 year baseline) 0.380%

As stated in Section 2.0, a change in the CCFP of up to 1.5% is assumed to be small. The increase in the CCFP from the 3 in 10 year interval to 1 in 15 year interval is 0.911%. Therefore, this increase is judged to be small.

5.2.6 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage An estimate of the likelihood and risk implications of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is evaluated using a methodology similar to the Calvert Cliffs liner corrosion analysis [Reference 5]. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner.

The following approach is used to determine the change in likelihood, due to extending the ILRT, of detecting corrosion of the containment steel liner. This likelihood is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:

Differences between the containment basemat and the containment cylinder and dome The historical steel liner flaw likelihood due to concealed corrosion The impact of aging The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw Assumptions Consistent with the Calvert Cliffs analysis, a half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures (See Table 5-15, Step 1).

In the 5.5 years following September 1996 when 10 CFR 50.55a started requiring visual inspection, there were three events where a through wall hole in the containment liner was identified. These are Brunswick 2 on 4/27/99, North Anna 2 on 9/23/99, and D. C.

Cook 2 in November 1999. The corrosion associated with the Brunswick event is believed to have started from the coated side of the containment liner. Although WCGS has a different containment type, this event could potentially occur at WCGS (i.e.,

corrosion starting on the coated side of containment). Construction material embedded in the concrete may have contributed to the corrosion. The corrosion at North Anna is believed to have started on the uninspectable side of containment due to wood imbedded in the concrete during construction. The D.C. Cook event is associated with Revision 1 Page 26 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension an inadequate repair of a hole drilled through the liner during construction. Since the hole was created during construction and not caused by corrosion, this event does not apply to this analysis. Based on the above data, there are two corrosion events from the 5.5 years that apply to WCGS.

Consistent with the Calvert Cliffs analysis, the estimated historical flaw probability is also limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date (and have been performed since the time frame of the Calvert Cliffs analysis)

(See Table 5-4, Step 1).

Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages (See Table 5-15, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.

In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere, given that a liner flaw exists, was estimated as 1.1% for the cylinder and dome, and 0.11% (10% of the cylinder failure probability) for the basemat. These values were determined from an assessment of the probability versus containment pressure. For WCGS, the ILRT maximum pressure is 50.1 psig [Reference 30].

Probabilities of 1% for the cylinder and dome, and 0.1% for the basemat are used in this analysis, and sensitivity studies are included in Section 5.3.1 (See Table 5-15, Step 4).

Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the basemat region is considered to be less likely than the containment cylinder and dome region (See Table 5-15, Step 4).

In the Calvert Cliffs analysis, it is noted that approximately 85% of the interior wall surface is accessible for visual inspections. Consistent with the Calvert Cliffs analysis, a 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used. To date, all liner corrosion events have been detected through visual inspection (See Table 5-15, Step 5).

Consistent with the Calvert Cliffs analysis, all non-detectable containment failures are assumed to result in early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

Table 5 Steel Liner Corrosion Base Case Step Description Containment Cylinder and Containment Basemat Dome (85%) (15%)

Historical liner flaw likelihood Events: 2 Events: 0 Failure data: containment location (Brunswick 2 and North Anna 2) Assume a half failure specific 2 / (70 x 5.5) = 5.19E-03 0.5 / (70 x 5.5) = 1.30E-03 1 Success data: based on 70 steel-lined containments and 5.5 years since the 10 CFR 50.55a requirements of periodic visual inspections of containment surfaces Year Failure rate Year Failure rate Aged adjusted liner flaw likelihood 2 During the 15-year interval, assume 1 2.05E-03 1 5.13E-04 failure rate doubles every five years average 5-10 5.19E-03 average 5-10 1.30E-03 (14.9% increase per year). The 15 1.43E-02 15 3.57E-03 Revision 1 Page 27 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Steel Liner Corrosion Base Case Step Description Containment Cylinder and Containment Basemat Dome (85%) (15%)

average for the 5th to 10th year set to the historical failure rate. 15 year average = 6.44E-03 15 year average = 1.61E-03 Increase in flaw likelihood between 3 and 15 years Uses aged adjusted 0.71% (1 to 3 years) 0.18% (1 to 3 years) 3 liner flaw likelihood (Step 2), 4.14% (1 to 10 years) 1.04% (1 to 10 years) assuming failure rate doubles every 9.66% (1 to 15 years) 2.42% (1 to 15 years) five years.

Likelihood of breach in containment 4 1% 0.1%

given liner flaw 10%

5% failure to identify visual flaws plus 5% likelihood that the flaw is not visible (not through-cylinder 100%

Visual inspection detection failure 5 but could be detected by ILRT).

likelihood Cannot be visually inspected All events have been detected through visual inspection. 5%

visible failure detection is a conservative assumption.

0.00071% (3 years) 0.00018% (3 years) 0.71% x 1% x 10% 0.18% x 0.1% x 100%

Likelihood of non-detected 0.00414% (10 years) 0.00104% (10 years) 6 containment leakage (Steps 3 x 4 x

5) 4.14% x 1% x 10% 1.04% x 0.1% x 100%

0.00966% (15 years) 0.00242% (15 years) 9.66% x 1% x 10% 2.42% x 0.1% x 100%

The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the containment cylinder and dome, and the containment basemat, as summarized below for WCGS.

Table 5 Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for WCGS Description At 3 years: 0.00071% + 0.00018% = 0.00089%

At 10 years: 0.00414% + 0.00104% = 0.00517%

At 15 years: 0.00966% + 0.00242% = 0.01207%

The above factors are applied to those core damage accidents that are not already independently LERF or that could never result in LERF.

The two corrosion events that were initiated from the non-visible (backside) portion of the containment liner used to estimate the liner flaw probability in the Calvert Cliffs analysis are assumed to be applicable to this containment analysis. These events, one at North Anna Unit 2 (September 1999) caused by timber embedded in the concrete immediately behind the containment liner, and one at Brunswick Unit 2 (April 1999) caused by a cloth work glove embedded in the concrete next to the liner, were initiated from the nonvisible (backside) portion of the containment liner. A search of the NRC website LER database identified two additional events have occurred since the Calvert Cliffs analysis was performed. In January 2000, a 3/16-inch circular through-liner hole was found at Cook Nuclear Plant Unit 2 caused by a wooden Revision 1 Page 28 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension brush handle embedded immediately behind the containment liner. The other event occurred in April 2009, where a through-liner hole approximately 3/8-inch by 1-inch in size was identified in the Beaver Valley Power Station Unit 1 (BVPS-1) containment liner caused by pitting originating from the concrete side due to a piece of wood that was left behind during the original construction that came in contact with the steel liner [Reference 34]. Two other containment liner through-wall hole events occurred at Turkey Point Units 3 and 4 in October 2010 and November 2006, respectively. However, these events originated from the visible side caused by the failure of the coating system, which was not designed for periodic immersion service, and are not considered to be applicable to this analysis. More recently, in October 2013, some through-wall containment liner holes were identified at BVPS-1, with a combined total area of approximately 0.395 square inches. The cause of these through-wall liner holes was attributed to corrosion originating from the outside concrete surface due to the presence of rayon fiber foreign material that was left behind during the original construction and was contacting the steel liner. For risk evaluation purposes, these five total corrosion events occurring in 66 operating plants with steel containment liners over a 17.1 year period from September 1996 to October 4, 2013 (i.e., 5/(66*17.1) = 4.43E-03) are bounded by the estimated historical flaw probability based on the two events in the 5.5 year period of the Calvert Cliffs analysis (i.e.,

2/(70*5.5) = 5.19E-03) incorporated in the EPRI guidance [Reference 34].

5.2.7 Impact from External Events Contribution An assessment of the impact of external events is performed. The primary purpose for this investigation is the determination of the total LERF following an increase in the ILRT testing interval from 3 in 10 years to 1 in 15 years.

The IPEEE Fire PRA calculated a CDF of 8.14E-6 [Reference 18]. Since no Fire LERF value is calculated, it is assumed the LERF/CDF ratio will be similar for fire risk as for internal events risk. Applying the internal event LERF/CDF ratio to the Fire CDF yields an estimated Fire LERF of 5.48E-8, as shown by the equation below.

LERFFire CDFFire

• LERFIE / CDFIE = 8.14E-6

• 1.10E-7 / 1.63E-5 = 5.48E-8 As described in Section 5.1.3, additional consideration is made to not apply failure probabilities on those cases that are already LERF scenarios. Therefore, these LERF contributions from CDF are removed. To reduce conservatism in the ILRT extension analysis, the methodology of subtracting existing LERF from CDF is also applied to the Fire PRA model. The following shows the calculation for Class 3b:

0.5 C = C = 8.14 -6 5.48 -8 = 1.85E-8 218

_ _ _.D C _`a = C = E 8.14 -6 5.48 -8 = 6.18E-8 D _.D C D`a = C =5 E 8.14 -6 5.48 -8 = 9.27E-8 The 2014 Seismic Reevaluations for operating reactor sites [Reference 29] states the conclusions reached in 2010 by GI-199 [Reference 28] remain valid for estimating Seismic CDF at plants in the Central and Eastern United States, which includes WCGS. EPRI guidance

[Reference 29] on recent seismic evaluations states, "EPRI does not recommend using any very conservative approaches to estimate the SCDF such as use of the maximum SCDFs calculated at any one frequency. This type of bounding approach is overly conservative and judged to not provide realistic risk estimates consistent with SCDFs calculated in actual SPRAs. Therefore, the simple average of 6.45E-06 calculated from the reported CDF values in Table D-1 of GI-199

[Reference 28] is used for the Seismic CDF. Since no Seismic LERF value is calculated, it is Revision 1 Page 29 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension assumed the LERF/CDF ratio will be similar for seismic risk as for internal events risk. Applying the internal event LERF/CDF ratio to the seismic CDF yields an estimated seismic LERF of 4.34E-8, as shown by the equation below.

LERFSeismic CDFSeismic

• LERFIE / CDFIE = 6.45E-6

• 1.10E-7 / 1.63E-5 = 4.34E-8 Subtracting seismic LERF from CDF, the Class 3b frequency can be calculated by the following formulas:

_.D C = C = 6.45 -6 4.34 -8 = 1.47E-8 E

_ _ _.D C _`a = C = E 6.45 -6 4.34 -8 = 4.90E-8 D _.D C D`a = C =5 6.45 -6 4.34 -8 = 7.35E-8 E

The external event contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The change in LERF is calculated for the 1 in 10 year and 1 in 15 year cases and the change defined for the external events in Table 5-17.

Table 5 WCGS External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 per 15 years) 3 per 10 year 1 per 10 year 1 per 15 years External Events 3.32E-08 1.11E-07 1.66E-07 1.33E-07 Internal Events 3.71E-08 1.24E-07 1.85E-07 1.48E-07 Combined 7.03E-08 2.34E-07 3.52E-07 2.81E-07 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined, the increase due to increasing the interval from 10 to 15 years is 1.17E-7; the total change in LERF due to increasing the ILRT interval from 3 to 15 years is 2.81E-7, which meets the guidance for small change in risk, as it exceeds 1.0E-7/yr and remains less than a 1.0E-6 change in LERF. For this change in LERF to be acceptable, total LERF must be less than 1.0E-5. As in Section 5.2.4, the total LERF is calculated below by adding external and internal events LERF and change in LERF:

LERF = LERFIE + LERFfire + LERFseismic + LERFclass3Bincrease LERF = 1.10E-7/yr + 5.48E-8/yr + 4.34E-8/yr + 2.81E-7/yr = 4.89E-7/yr Several conservative assumptions were made in this ILRT analysis, as discussed in Sections 4.0, 5.1.3, 5.2.1, and 5.2.4; therefore, the total change in LERF is considered conservative for this application. As specified in Regulatory Guide 1.174 [Reference 4], since the total LERF is less than 1.0E-5, it is acceptable for the LERF to be between 1.0E-7 and 1.0E-6.

5.2.7.1 Other External Hazards Several other external events, including high winds, external floods, and transportation and nearby facility accidents, were evaluated and screened out in the IPEEE [Reference 32].

Chapter 3 of the USAR [Reference 27] maintains the screening for aircraft hazards, tornado missiles, and nearby explosions or military activities. Reference 31 provides an updated Revision 1 Page 30 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension external flooding evaluation and concludes effective flood protection exists from the reevaluated flood hazards.

5.2.8 Defense-In-Depth Impact Regulatory Guide 1.174, Revision 3 [Reference 4] describes an approach that is acceptable for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights. One of the considerations included in RG 1.174 is Defense in Depth. Defense in Depth is a safety philosophy that employs successive compensatory measures to prevent accidents or mitigate damage if a malfunction, accident, or naturally caused event occurs at a nuclear facility. The following seven considerations as presented in RG 1.174, Revision 3, Section C.2.1.1.2 will serve to evaluate the proposed licensing basis change for overall impact on Defense in Depth.

1. Preserve a reasonable balance among the layers of defense.

The use of the risk metrics of LERF, population dose, and conditional containment failure probability collectively ensures the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. The change in LERF is small with respect to internal events and small when including external events per RG 1.174, and the change in population dose and CCFP are small as defined in this analysis and consistent with NEI 94-01 Revision 3-A.

2. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

The adequacy of the design feature (the containment boundary subject to Type A testing) is preserved as evidenced by the overall small change in risk associated with the Type A test frequency change.

3. Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

The redundancy, independence, and diversity of the containment subject to the Type A test is preserved, commensurate with the expected frequency and consequences of challenges to the system, as evidenced by the overall small change in risk associated with the Type A test frequency change.

4. Preserve adequate defense against potential CCFs.

Adequate defense against CCFs is preserved. The Type A test detects problems in the containment which may or may not be the result of a CCF; such a CCF may affect failure of another portion of containment (i.e., local penetrations) due to the same phenomena. Adequate defense against CCFs is preserved via the continued performance of the Type B and C tests and the performance of inspections. The change to the Type A test, which bounds the risk associated with containment failure modes including those involving CCFs, does not degrade adequate defense as evidenced by the overall small change in risk associated with the Type A test frequency change.

5. Maintain multiple fission product barriers.

Multiple Fission Product barriers are maintained. The portion of the containment affected by the Type A test extension is still maintained as an independent fission product barrier, albeit with an overall small change in the reliability of the barrier.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension

6. Preserve sufficient defense against human errors.

Sufficient defense against human errors is preserved. The probability of a human error to operate the plant, or to respond to off-normal conditions and accidents is not significantly affected by the change to the Type A testing frequency. Errors committed during test and maintenance may be reduced by the less frequent performance of the Type A test (less opportunity for errors to occur).

7. Continue to meet the intent of the plants design criteria.

The intent of the plants design criteria continues to be met. The extension of the Type A test does not change the configuration of the plant or the way the plant is operated.

5.3 Sensitivities 5.3.1 Potential Impact from Steel Liner Corrosion Likelihood A quantitative assessment of the contribution of steel liner corrosion likelihood impact was performed for the risk impact assessment for extended ILRT intervals. As a sensitivity run, the internal event CDF was used to calculate the Class 3b frequency. The impact on the Class 3b frequency due to increases in the ILRT surveillance interval was calculated for steel liner corrosion likelihood using the relationships described in Section 5.2.6. The EPRI Category 3b frequencies for the 3 per 10-year, 10-year, and 15-year ILRT intervals were quantified using the internal events CDF. The change in the LERF, change in CCFP, and change in Annual Dose Rate due to extending the ILRT interval from 3 in 10 years to 1 in 10 years, or to 1 in 15 years are provided in Table 5 Table 5-20. The steel liner corrosion likelihood was increased by a factor of 1000, 10000, and 100000. Except for extreme factors of 10000 and 100000, the corrosion likelihood is relatively insensitive to the results.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Steel Liner Corrosion Sensitivity Case: 3B Contribution 3b 3b 3b LERF LERF LERF Frequency Frequency Frequency Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

Corrosion Likelihood 3.71E-08 1.24E-07 1.85E-07 8.65E-08 1.48E-07 6.18E-08 X1 Corrosion Likelihood 3.74E-08 1.30E-07 2.08E-07 9.26E-08 1.70E-07 7.78E-08 X 1000 Corrosion Likelihood 4.04E-08 1.88E-07 4.09E-07 1.47E-07 3.69E-07 2.22E-07 X 10000 Corrosion Likelihood 7.01E-08 7.63E-07 2.42E-06 6.93E-07 2.35E-06 1.66E-06 X 100000 Table 5 Steel Liner Corrosion Sensitivity: CCFP CCFP CCFP CCFP CCFP CCFP CCFP Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

Corrosion Likelihood 1.16E-01 1.22E-01 1.25E-01 5.32E-03 9.11E-03 3.80E-03 X1 Corrosion Likelihood 1.16E-01 1.22E-01 1.25E-01 5.36E-03 9.19E-03 3.83E-03 X 1000 Corrosion Likelihood 1.16E-01 1.22E-01 1.26E-01 5.79E-03 9.92E-03 4.13E-03 X 10000 Corrosion Likelihood 1.18E-01 1.28E-01 1.35E-01 1.00E-02 1.72E-02 7.17E-03 X 100000 Table 5 Steel Liner Corrosion Sensitivity: Dose Rate Dose Rate Dose Rate Dose Rate Dose Rate Dose Rate Dose Rate Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

Corrosion 1.19E-02 3.97E-02 5.96E-02 2.78E-02 4.77E-02 1.99E-02 Likelihood X 1 Corrosion Likelihood X 1.20E-02 4.18E-02 6.68E-02 2.98E-02 5.48E-02 2.50E-02 1000 Corrosion Likelihood X 1.30E-02 6.03E-02 1.32E-01 4.73E-02 1.19E-01 7.13E-02 10000 Corrosion Likelihood X 2.25E-02 2.45E-01 7.79E-01 2.23E-01 7.57E-01 5.34E-01 100000 Revision 1 Page 33 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension 5.3.2 Expert Elicitation Sensitivity Another sensitivity case on the impacts of assumptions regarding pre-existing containment defect or flaw probabilities of occurrence and magnitude, or size of the flaw, is performed as described in Reference 24. In this sensitivity case, an expert elicitation was conducted to develop probabilities for pre-existing containment defects that would be detected by the ILRT only based on the historical testing data.

Using the expert knowledge, this information was extrapolated into a probability-versus-magnitude relationship for pre-existing containment defects [Reference 24]. The failure mechanism analysis also used the historical ILRT data augmented with expert judgment to develop the results. Details of the expert elicitation process and results are contained in Reference 24. The expert elicitation process has the advantage of considering the available data for small leakage events, which have occurred in the data, and extrapolate those events and probabilities of occurrence to the potential for large magnitude leakage events.

The expert elicitation results are used to develop sensitivity cases for the risk impact assessment. Employing the results requires the application of the ILRT interval methodology using the expert elicitation to change the probability of pre-existing leakage in the containment.

The baseline assessment uses the Jeffreys non-informative prior and the expert elicitation sensitivity study uses the results of the expert elicitation. In addition, given the relationship between leakage magnitude and probability, larger leakage that is more representative of large early release frequency, can be reflected. For the purposes of this sensitivity, the same leakage magnitudes that are used in the basic methodology (i.e., 10 La for small and 100 La for large) are used here. Table 5-21 presents the magnitudes and probabilities associated with the Jeffreys non-informative prior and the expert elicitation used in the base methodology and this sensitivity case.

Table 5 WCGS Summary of ILRT Extension Using Expert Elicitation Values (from Reference 24)

Leakage Size (La) Expert Elicitation Mean Probability of Occurrence Percent Reduction 10 3.88E-03 86%

100 2.47E-04 91%

Taking the baseline analysis and using the values provided in Table 5-10 and Table 5-11 for the expert elicitation sensitivity yields the results in Table 5-22.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 WCGS Summary of ILRT Extension Using Expert Elicitation Values Accident ILRT Interval Class 3 per 10 Years 1 per 10 Years 1 per 15 Years Base Adjusted Dose Dose Frequency Dose Frequency Dose Frequency Base (person- Rate Rate Rate Frequency rem) (person- (person- (person-rem/yr) rem/yr) rem/yr) 1 1.44E-05 1.44E-05 3.21E+03 4.62E-02 1.42E-05 4.57E-02 1.41E-05 4.53E-02 2 2.95E-08 2.95E-08 1.85E+05 5.46E-03 2.95E-08 5.46E-03 2.95E-08 5.46E-03 3a N/A 6.27E-08 3.21E+04 2.02E-03 2.09E-07 6.72E-03 3.14E-07 1.01E-02 3b N/A 3.99E-09 3.21E+05 1.28E-03 1.33E-08 4.28E-03 2.00E-08 6.42E-03 7 1.76E-06 1.76E-06 1.14E+05 2.01E-01 1.76E-06 2.01E-01 1.76E-06 2.01E-01 8 6.12E-08 6.12E-08 8.24E+05 5.04E-02 6.12E-08 5.04E-02 6.12E-08 5.04E-02 Totals 1.63E-05 1.63E-05 1.48E+06 3.07E-01 1.63E-05 3.14E-01 1.63E-05 3.19E-01 LERF N/A 9.32E-09 1.60E-08 (3 per 10 yrs base)

LERF N/A N/A 6.66E-09 (1 per 10 yrs base)

CCFP 11.42% 11.47% 11.52%

The results illustrate how the expert elicitation reduces the overall change in LERF and the overall results are more favorable with regard to the change in risk.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension 6.0 RESULTS The Internal Events results from this ILRT extension risk assessment for WCGS are summarized in Table 6-1.

Table 6 ILRT Extension Summary (Internal Events)

Class Dose Base Case Extend to Extend to (person- 3 in 10 Years 1 in 10 Years 1 in 15 Years rem)

CDF/yr Person- CDF/yr Person- CDF/yr Person-Rem/yr Rem/yr Rem/yr 1 3.21E+03 1.42E-05 4.58E-02 1.38E-05 4.44E-02 1.35E-05 4.34E-02 2 1.85E+05 2.95E-08 5.46E-03 2.95E-08 5.46E-03 2.95E-08 5.46E-03 3a 3.21E+04 1.49E-07 4.79E-03 4.97E-07 1.60E-02 7.45E-07 2.40E-02 3b 3.21E+05 3.71E-08 1.19E-02 1.24E-07 3.97E-02 1.85E-07 5.96E-02 7 1.14E+05 1.76E-06 2.01E-01 1.76E-06 2.01E-01 1.76E-06 2.01E-01 8 8.24E+05 6.12E-08 5.04E-02 6.12E-08 5.04E-02 6.12E-08 5.04E-02 Total 1.63E-05 3.20E-01 1.63E-05 3.57E-01 1.63E-05 3.84E-01 ILRT Dose Rate from 3a and 3b Total From 3 N/A 3.76E-02 6.45E-02 Dose Rate Years (Person- From 10 Rem/yr) N/A N/A 2.69E-02 Years From 3 N/A 11.76% 20.17%

%Dose Years Rate From 10 N/A N/A 7.52%

Years 3b Frequency (LERF/yr)

From 3 N/A 8.65E-08 1.48E-07 Years LERF From 10 N/A N/A 6.18E-08 Years CCFP %

From 3 N/A 0.532% 0.911%

Years CCFP%

From 10 N/A N/A 0.380%

Years Revision 1 Page 36 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension

7.0 CONCLUSION

S AND RECOMMENDATIONS Based on the results from Section 5.2 and the sensitivity calculations presented in Section 5.3, the following conclusions regarding the assessment of the plant risk are associated with extending the Type A ILRT test frequency to 15 years:

Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines small changes in risk as resulting in increases of CDF greater than 1.0E-6/yr and less than 1.0E-5/yr and increases in LERF greater than 1.0E-7/yr and less than 1.0E-6/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 1.48E-7/yr using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Total Internal Events LERF (baseline and change in LERF due to the ILRT extension) is 2.58E-7. Therefore, the estimated change in LERF is determined to be small using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4]. The risk change resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change. Considering the increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 10 years to 1 in 15 years is estimated as 6.18E-8/yr, the risk increase is very small using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].

When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 2.81E-7/yr using the EPRI guidance, and total LERF is 4.89E-7/yr. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4]. The risk change resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change. When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 1 in 10 years to 1 in 15 years is estimated as 1.17E-7/yr, and the total LERF is 3.25E-7/yr. Therefore, the risk increase is small using the acceptance guidelines of Regulatory Guide 1.174

[Reference 4].

The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing is 0.645 person-rem/yr. NEI 94-01 [Reference 1] states that a small population dose is defined as an increase of 1.0 person-rem per year, or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

The increase in the conditional containment failure probability from the 3 in 10 year interval to 1 in 15 year interval is 0.911%. NEI 94-01 [Reference 1] states that increases in CCFP of 1.5% is small. Therefore, this increase is judged to be small.

Therefore, increasing the ILRT interval to 15 years is considered to be small since it represents a small change to the WCGS risk profile.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Previous Assessments The NRC in NUREG-1493 [Reference 6] has previously concluded that:

Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.

The conclusions for WCGS confirm these general conclusions on a plant-specific basis considering the severe accidents evaluated for WCGS, the WCGS containment failure modes, and the local population surrounding WCGS.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension A. PRA ACCEPTABILITY A.1. PRA Quality Statement for Permanent 15-Year ILRT Extension Revision 9 of the WCGS PRA model is the most recent evaluation of internal event risk. The WCGS PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the WCGS PRA is based on the single top fault tree methodology, which is a well-known PRA methodology in the industry.

The internal events PRA model [References 17] has been assessed against RG 1.200

[Reference 36]. The internal events PRA model was subject to a full-scope peer review conducted in accordance with RG 1.200 in June 2019 [Reference 35]. The independent assessment of the F&O closures was performed between November 2019 and March 2020, including a two-day meeting at the Wolf Creek site in Burlington, KS, on December 10 and 11, 2019 [Reference 39].

The F&O closure independent assessment was performed following the guidance of Appendix X of the Nuclear Energy Institute (NEI) peer review guidance document NEI 05-04 [Reference 33]

as well as the process clarifications provided by the United States Nuclear Regulatory Commission (USNRC), which are documented in Reference 37. This independent assessment also considered the lessons learned from the USNRC staff observations of three independent assessment team pilot reviews [Reference 38]. After this F&O closure, three of the original peer review F&Os remained open. During the F&O closure review, one F&O was judged to be closed with a PRA upgrade, which required a focused scope peer review. This triggered a focused scope peer review of the Supporting Requirements associated with the upgrade. Following the focused scope peer review, all the involved SRs have been judged to be met at Capability Category II or higher, but a new F&O (AS-B3-01) has been assigned and remains open. The four open F&Os are detailed in Section A.2. F&Os 3-8 and 6-8 remain open following the closure assessment. Since F&Os 3-8 and 6-8 relate to documentation, they do not affect the PRA acceptability for use in the ILRT extension analysis. F&O 4-10 was not reviewed during this F&O closure. One part of F&O 4-10 pertains to the ISLOCA modeling only meeting Capability Category I, but this does not negatively affect the acceptability of the PRA for the ILRT extension analysis. The other part of F&O 4-10 relates to documentation, so it does not affect the PRA acceptability for use in the ILRT extension analysis. New F&O AS-B3-01 pertains to the disposition of the potential for sump strainer blockage during feed and bleed events with open PORVs. A bounding sensitivity study demonstrates no impact on the conclusions of this analysis. Therefore, the PRA is of sufficient quality and level of detail to support the ILRT extension analysis.

Wolf Creek Nuclear Operating Company (WCNOC) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models. This approach includes both a proceduralized PRA maintenance and update process and the use of self-assessments and independent peer reviews. The following information describes the WCNOC approach to PRA model maintenance.

A.1.1 PRA Maintenance and Update The WCNOC risk management process ensures that the applicable PRA models used in this application continue to reflect the as-built and as-operated plant. The process delineates the responsibilities and guidelines for updating the PRA models, and includes criteria for both regularly scheduled and interim PRA model updates. The process includes provisions for monitoring potential areas affecting the PRA models (e.g., due to changes in the plant, errors or Revision 1 Page 39 of 45

54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension limitations identified in the model, and industry operational experience), for assessing the risk impact of unincorporated changes, and for controlling the model and associated computer files.

The process will assess the impact of these changes on the plant PRA model in a timely manner.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension A.2. Wolf Creek Open Findings F&O Number: 3-8 Applicable Supporting Requirements: AS-C3, HR-I3, IE-D3, SC-C3, SY-C3, QU-F4 Closure Status: Open Upgrade? No New Method: No This constitutes model maintenance. Tracking of uncertainties associated with assumptions in the overall risk profile is not an upgrade as new methods are not normally added and the base risk profile is not changed.

Closure Status: Open F&O

Description:

Identify plant specific sources of uncertainty. This identification can be documented in a manner similar to the tables that characterize the generic sources of model uncertainty and related assumptions.

F&O Basis:

Sources of uncertainty are required to be identified.

F&O Proposed Solution:

Identify plant specific sources of uncertainty. This identification can be documented in a manner similar to the tables that characterize the generic sources of model uncertainty and related assumptions.

Resolution to F&O:

All documents have been reviewed and updated to ensure each assumption and source of uncertainty is properly identified and characterized (conservative, non-conservative, realistic).

Independent Assessment Team Assessment of F&O Resolution:

OPEN - Individual assumptions in the various notebooks have been reviewed with an initial characterization of the associated uncertainties. There is still a gap in the assessment of the individual assumption and the final uncertainty assessment in the quantification notebook. The section discussing sensitivities (which is where any quantitative assessment on the impact of uncertainties associated with assumption should be discussed) does not allow a clear tracking on which assumption in the rest of the documentation is being addressed. Especially for assumptions that are marked as non conservative a closure statement needs to be added somewhere on the importance for the results. It is also recommended that an assessment on some of the most conservative assumptions is made, to ensure that risk insights are not masked.

Impact on ILRT Extension:

Since this F&O does not affect the base risk profile, it does not affect the ILRT extension analysis.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension F&O Number: 4-10 Applicable Supporting Requirement: LE-C13 Closure Status: Open F&O

Description:

The simplified approach to scrubbing of ISLOCA releases meets the CC-I requirements but does not meet the CC-II requirements. The approach to scrubbing of SGTR releases is consistent with the CC-II requirements but does not provide sufficient technical basis to justify the credit taken.

F&O Basis:

The SR is not met at CC-II for ISLOCA events. Additional documentation of SGTR scrubbing is needed.

Possible Resolution:

For ISLOCA events: Identify release locations and assess possible radionuclide scrubbing for each release location.

For SGTR events: Provide additional documentation of the engineering basis by citing appropriate plant-specific or generic analyses.

Impact on ILRT Extension:

A conservative ISLOCA CDF and LERF is conservative for overall risk and overall dose rate.

Since ISLOCA risk is classified as Class 8, a conservative ISLOCA CDF and LERF does not affect the LERF or Dose Rate results; if ISLOCA contribution to LERF were reduced, the LERF, Dose Rate, and CCFP would not change. Since a decrease in ISLOCA LERF decreases the overall Dose Rate, the %Dose Rate increases, but this increase does not impact the conclusions of the ILRT extension evaluation because Dose Rate is less than the acceptance criteria.

For SGTR events, the F&O is only related to documentation.

Therefore, resolution of this F&O would not negatively affect this ILRT extension analysis.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension F&O Number: 6-8 Applicable Supporting Requirement: SY-C2 Closure Status: Open Upgrade? No New Method: No No model update was required - documentation only update.

F&O

Description:

The notebook states that walkdowns and interviews were performed but not documented.

Without the documentation there is no evidence that these tasks were performed and that the walkdown was included the present as built plant.

F&O Basis:

There is no evidence that a walkdown or operator interview was performed and when these tasks were performed.

Possible Resolution:

The results of the walkdowns and interviews should be included in the system analysis documentation.

Impact on ILRT Extension:

A new set of System Engineer Interviews have been completed, and walkdowns were completed during outage and documented through the system health reporting process. The majority of the interviews with System Engineers were performed and documented in the appropriate system notebooks. The results of the interviews and insights are appropriately captured in both the interview notes as well as the text. The task remains incomplete as one interview is still needed to the Electrical Systems engineer [39]. Therefore, the open part of this F&O is only related to documentation.

Since this F&O is only related to documentation, it does not affect the ILRT extension analysis.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension F&O Number: AS-B3-01 Applicable Supporting Requirement: AS-B3 Closure Status: Open Upgrade? N/A New Method: N/A New F&O resulting from a focused scope peer review of original F&O resolution that constituted a PRA upgrade.

F&O

Description:

Feed and bleed scenarios involving open PORVs did not consider the potential for sump strainer blockage. The review identified no model logic or a documented basis that would address open PORV transients including considerations of the complications associated with containment sump blockage with the actuation of containment spray.

F&O Basis:

A review of plant documentation and event tree models did not result in evidence of treatment of feed and bleed scenarios where sump plugging cases with the possibility of spray actuation were considered. As an example, the application or disposition of SUMP-NPSH-NONLOCA to feed and bleed sequences with open PORVs is not addressed.

Possible Resolution:

Add to the model or document the basis for not modeling containment sump blockage for feed and bleed scenarios.

Impact on ILRT Extension:

The Wolf Creek PRA models the potential for sump strainer blockage during various accident scenarios, including pipe break LOCAs, RCP seal LOCA, and stuck open PORV LOCAs. The F&O appears to be concerned with possible sump strainer blockage that could occur during feed and bleed events. Feed and bleed is used as the intentional method of core cooling in transient events in which secondary heat removal is not available or is failed. In these feed and bleed scenarios, containment spray may actuate based on containment conditions induced by feed and bleed operation. In this case, sump strainer blockage modeled by basic event SUMP-NPSH-NONLOCA may be applicable to the scenario. The impact of applying this basic event in the Wolf Creek PRA to all transient and internal flooding sequences and applying the probability of a Class 3b pre-existing leak is judged to have no impact on the conclusion of the ILRT extension analysis.

The probability of basic event SUMP-NPSH-NONLOCA is 1E-5, and the probability of a pre-existing leak for a 1 in 15 year ILRT surveillance interval is 1.15E-2. Multiplying the two probabilities yields 1.15E-7. This probability would be multiplied by the frequency of transient and internal flooding events in which feed and bleed is required for mitigation of core damage, and then the risk of the 3 in 10 year ILRT surveillance interval case would be similarly estimated and subtracted to yield the additional change in LERF for this application. Totaling the internal flood and transient initiating events yields a frequency of ~1.1/yr [Reference 41]; therefore, the resulting LERF estimate would be (1.1

• 1E-5

• 1.15E-2) - (1.1

• 1E-5

• 2.3E-3) = ~1.0E-7, which would be considered small. Adding this to the existing LERF (1.48E-7) and total LERF (2.58E-7) yields ~2.5E-7 LERF and ~3.6E-7 total LERF. Performing similar calculations with the external events results in LERF of ~3.8E-7 and a total LERF of ~5.9E-7.

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54016-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension These bounding estimates demonstrate the conclusion of the ILRT extension analysis is not sensitive to any increase in CDF and LERF that would result from inclusion of the 1E-5 sump strainer failure probability for feed and bleed events in the Wolf Creek PRA. Therefore, the potential resolution to this F&O does not affect the conclusion of the ILRT extension analysis.

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