RS-21-014, License Amendment Request for One-Time Extension of the Containment Type a Integrated Leakage Rate Test Frequency

License Amendment Request for One-Time Extension of the Containment Type a Integrated Leakage Rate Test Frequency

ML21055A822
Person / Time
Site:	Clinton
Issue date:	02/24/2021
From:	Simpson P Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-21-014
Download: ML21055A822 (125)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-21-014 10 CFR 50.90 February 24, 2021 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

License Amendment Request for One-Time Extension of the Containment Type A Integrated Leakage Rate Test Frequency In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License No. NPF-62 for Clinton Power Station (CPS), Unit 1. The proposed change would allow for a one-time extension to the 15-year frequency of the CPS Unit 1 containment leakage rate test (i.e., integrated leakage rate test (ILRT) or Type A test). This test is required by Technical Specifications Section 5.5.13, "Primary Containment Leakage Rate Testing Program." The proposed one-time change would permit the current ILRT interval of 15 years to be extended by eight months.

This request is subdivided as follows.

• Attachment 1 provides a description and evaluation of the proposed change.

• Attachment 2 provides a markup of the affected TS page.

• Attachment 3 provides an assessment of risk associated with the one-time extension.

The proposed change has been reviewed by the Plant Operations Review Committee in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed change by July 30, 2021. Once approved, the amendment will be implemented within 60 days. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), EGC is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

February 24, 2021 U.S. Nuclear Regulatory Commission Page 2 There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 24th day of February 2021.

Respectfully, Patrick R. Simpson Sr. Manager Licensing Attachments:

1. Evaluation of Proposed Change

2. Markup of Technical Specifications Page

3. Risk Assessment for CPS Regarding the ILRT (Type A) One-Time Extension Request cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - Clinton Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Change

SUBJECT:

License Amendment Request for One-Time Extension of the Containment Type A Integrated Leakage Rate Test Frequency 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

3.1 Description of Primary Containment System 3.2 Description of Drywell 3.3 Chronology of Testing Requirements of 10 CFR 50, Appendix J 3.4 Current CPS ILRT Requirements 3.5 Integrated Leakage Rate Testing History 3.6 Plant Specific Confirmatory Analysis 3.7 Non-Risk Based Assessment 3.8 Results of Recent Inspections

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 No Significant Hazards Consideration 4.3 Conclusions

6.0 REFERENCES

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ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License No. NPF-62 for Clinton Power Station (CPS), Unit 1. The proposed change would allow for a one-time extension to the 15-year frequency of the CPS Unit 1 containment leakage rate test (i.e., integrated leakage rate test (ILRT) or Type A test). This test is required by Technical Specifications (TS) Section 5.5.13, "Primary Containment Leakage Rate Testing Program." The proposed one-time change would permit the current ILRT interval of 15 years to be extended by eight months.

The last ILRT at CPS was completed in February 2008. Therefore, based on the 15-year ILRT frequency, the next ILRT must be performed by February 2023. In order to perform the test during a regularly scheduled refueling outage, the next ILRT at CPS would need to be completed no later than startup after the fall 2021 refueling outage.

On January 31, 2020, the U.S. Department of Health and Human Services declared a Coronavirus Disease (COVID-19) public health emergency (PHE) for the United States.

Subsequently, the Centers for Disease Control and Prevention issued recommendations (e.g.,

social distancing, limiting assemblies) to limit the spread of COVID-19.

In response to concerns of a continuation of the COVID-19 public health emergency, in the interest of personnel safety, and to preclude the potential for transmittal and spread of COVID-19, EGC requests a one-time extension of the CPS Type A test interval. This request is part of an overall effort by EGC to reduce the number of outside personnel required on-site, and the overall outage scope, in response to the developing COVID-19 pandemic situation while maintaining the safety and reliability of the plant for the next operating cycle. This effort assures that the overriding priority of nuclear safety is maintained while providing for plant personnel and public safety and health. Performing the Type A test would require vendor personnel from across the United States working alongside plant personnel in close proximity for extended periods of time. Including the ILRT in the fall outage scope would also increase the overall outage duration by approximately three days, increasing the amount of time that supplemental workforce would remain onsite. Therefore, the proposed change to extend the ILRT interval by eight months would allow CPS personnel to more effectively follow recommendations for social distancing.

2.0 DETAILED DESCRIPTION CPS TS 5.5.13, "Primary Containment Leakage Rate Testing Program," currently states, in part:

A program shall be established to implement the leakage rate testing of the primary 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 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 Page 2

ATTACHMENT 1 Evaluation of Proposed Change modified by the following exception: (1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Type A tests.

The proposed change to CPS TS 5.5.13 will add an exception to allow for the performance of the next Type A test no later than October 2023, which represents an extension of eight months.

This will allow performance of the test during the fall 2023 refueling outage, which is currently scheduled to begin in September 2023. In addition, the proposed change adds a second exception to allow the Type A test to be extended indefinitely if the test interval ends while primary containment integrity is not required (i.e., TS 3.6.1.1, "Primary Containment," does not require the primary containment to be operable in Modes 4 and 5). In this case, the second exception requires that the Type A test be performed prior to entering Mode 2. The revised wording states:

A program shall be established to implement the leakage rate testing of the primary 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 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) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Type A tests, (2) the next Type A test performed after the February 2008 Type A test shall be performed no later than October 31, 2023, and (3) if the Type A test has not been performed by October 31, 2023, and the unit is in Mode 4 or 5, the Type A test shall be performed prior to entering Mode 2 A markup of the proposed change is provided in Attachment 2.

3.0 TECHNICAL EVALUATION

3.1 Description of Primary Containment System The containment consists of a right circular cylinder with a hemispherical domed roof and a flat base slab. It is constructed of reinforced concrete and completely lined on the inside of the walls and dome with 1/4-inch stainless steel plate below elevation 735 feet 0 inch and with carbon steel plate of at least 1/4-inch thickness above elevation 735 feet 0 inch.

The principal dimensions of the containment are:

• height above basemat: 215 feet 0 inch;

• inside diameter: 124 feet 0 inch;

• wall thickness: 3 feet 0 inch;

• dome thickness: 2 feet 6 inches; and

• mat thickness: 9 feet 8 inches.

The containment structure supports the polar crane, galleries, and the access ramp to the refueling floor. The lower section of the containment acts as the outer boundary of the Page 3

ATTACHMENT 1 Evaluation of Proposed Change suppression pool. Two double-door personnel locks, one located at the refueling floor and the other located at the grade floor, permit access to the containment. An equipment hatch is located at the grade floor. The equipment hatch is sealed during normal operation, or at other times when primary containment is required.

3.1.1 Pipe Penetrations Pipe penetrations for process pipes which pass through the containment and drywell walls may be classified into three types. Type 1 is used for high-energy lines requiring guard pipes when passing through both the containment and drywell walls. Types 2 and 3 are used for the remainder of process pipes which pass through the containment. CPS Updated Safety Analysis Report (USAR) Figure 3.8-11 (Reference 1) shows the basic design of the three penetration types along with the inclined fuel transfer tube detail.

Type 1 penetrations consist of a guard pipe anchored at the containment wall and welded to the flued head. The flued head is welded to the process pipe using a gradual buildup weld. The process pipe is allowed free axial thermal movement from the flued head through the drywell.

The guard pipe is allowed free axial thermal movement from the containment anchor point through its own sleeve at the drywell wall. Bellows, anchored to the drywell and welded to the guard pipe, will act as a seal for normal drywell environmental conditions. They are designed for thermal guard pipe expansion and relative seismic motion of guard pipe and drywell.

Type 2 penetrations consist of a penetration sleeve anchored in the containment and extending to just inside the liner. Full penetration welds are used to weld the flued head to the process pipe.

Type 3 penetrations consist of the sleeve anchored in the containment wall and extending just beyond the containment liner. Full penetration welds are used to attach the cover plate to the process pipe.

3.1.2 Electrical Penetrations Dual header plate type electrical penetration assemblies are used to extend electrical conductors through the containment structure pressure boundary. These penetration assemblies are designed, fabricated, tested, and installed in accordance with the requirements of IEEE 317, "Standard for Electrical Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations," dated December 1976 (Reference 2).

3.1.3 Personnel and Equipment Access Hatches Two personnel access locks are provided for access to the interior of the containment.

Each personnel lock consists of an interlocked double door of welded steel assembly. Each door is equipped with a valve for equalizing pressure across the door such that the doors are not operable unless the pressure is equalized.

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ATTACHMENT 1 Evaluation of Proposed Change The two doors in each personnel lock are interlocked to prevent both being opened simultaneously and to ensure that one door is completely closed before the opposite door can be opened. An emergency lighting and communication system operating from an external auxiliary energy source is provided within the personnel locks.

The equipment hatch is fabricated from welded steel and furnished with a double-gasketed flange and bolted dished door. The hatch barrel is welded to the containment liner.

Provisions are made to pressure test the space between the double gaskets of the door flanges.

The weld seam test channels at the liner joint and the dished door are provided to monitor any leakage during leak rate testing.

3.1.4 Fuel Transfer Penetration The inclined fuel transfer tube, along with the three types of process pipe penetrations, penetrates the containment wall through the fuel transfer penetration. This is essentially a 3/4-inch-thick carbon steel rolled plate pipe sleeve of 40-inch ID with a 36-inch standard flange on the containment side. The fuel transfer penetration forms a part of the containment boundary.

A containment isolation assembly containing a blind flange and a bellows that connects from the containment isolation assembly to the building containment penetration are provided to make containment isolation. A hand-operated 24-inch gate valve is provided to isolate the reactor building pool water from the transfer tube so that the blind flange can be installed.

Normally, containment isolation is made by the containment isolation assembly and blind flange, containment bellows and the steel containment penetration. Special gaskets and double ply bellows are provided for leak checking to assure containment isolation. Alternatively, the blind flange may be removed for short periods of time during power operation, as allowed by the Technical Specifications. Leak testing of this alternative configuration (including transfer tube, associated drain line isolation valves, bellows, and flange connections) is not required because:

• these periods of time are short with respect to the overall duration of power operations,

• the transfer tube terminates below the fuel building spent fuel pool water level, and

• the configuration of the transfer tube drain line is controlled by the TS.

3.1.5 Containment Liner The containment wall liner is anchored to the wall with structural T sections. When a stiffener is cut to avoid interference with an insert assembly, welded studs are provided to restore anchorage of the liner plate.

Typical spacing of the liner anchors is 15 inches in the containment wall and the dome.

The top of the exposed base slab is lined with 1/2-inch and 1/4-inch stainless steel plate which serves as a leaktight boundary. The drywell wall and the sump floor are anchored through the base liner plate and into the base slab. The spans of liner panels in the basemat area are:

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ATTACHMENT 1 Evaluation of Proposed Change

• pedestal cavity area: 3 feet 0 inch;

• sump floor area: 6 feet 0 inch; and 20 feet 0 inch;

• suppression pool area: 3 feet 0 inch to 4 feet 8 1/4 inches (max.).

Leak test channels are provided at the liner seams in the suppression pool area and in the containment wall up to elevation 757 feet 0 inch. The containment liner in the wet areas of the suppression pool is of stainless steel to minimize corrosion problems.

3.2 Description of Drywell The drywell is a cylindrical reinforced concrete structure which surrounds the reactor pressure vessel and its support structure. The drywell is structurally designed as follows:

• to provide structural support to containment pools, main steam tunnel and reactor water cleanup (RWCU) compartments;

• to channel steam release from a LOCA through the horizontal vents for condensation in the suppression pool;

• to protect the containment vessel from internal missiles and/or pipe whip;

• to provide anchor points for pipes; and

• to provide a support structure for the work platforms, monorails, pipe supports, and restraints that are located in the annulus between the drywell and the containment vessel.

The inside diameter of the drywell cylinder is 69 feet 0 inch, and the wall thickness is 5 feet 0 inch. The top of the drywell consists of a flat annular slab 6 feet 0-inch-thick at elevation 803 feet 3 inches. The drywell wall is rigidly attached to the base slab at elevation 712 feet 0 inch. A steel head which can be removed to allow access to the reactor is located over the opening in the annular slab.

The lower portion of the drywell wall is submerged in the suppression pool. Three rows of circular suppression pool vents, 34 vents per row, penetrate the drywell wall below the normal level of the suppression pool. The surfaces of the drywell wall exposed to the suppression pool are lined with stainless steel clad plate 1-inch-thick, which is designed to act compositely with the drywell wall. Above the level of the suppression pool a carbon steel form plate 1/2-inch-thick is provided on the interior surfaces of the cylinder walls and top slab. Structural T's and headed studs are attached to the form plate to provide mechanical anchorage of the plate to the concrete and to stiffen the liner for construction loads. The form plate provides a surface for forming the drywell walls and ceiling and minimizes bypass leakage, if any, through the drywell wall under accident conditions.

3.2.1 Pipe Penetrations Piping penetrations are of the types used in the containment wall and are discussed in Section 3.1.1 above.

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ATTACHMENT 1 Evaluation of Proposed Change 3.2.2 Electrical Penetrations Electrical penetrations feature an epoxy-based sealing compound qualified for harsh environmental conditions, which surrounds the cables that pass through the penetration. The penetration consists of a rigid steel conduit welded to the drywell liner.

3.2.3 Personnel and Equipment Access Hatches Access to the drywell is provided by the drywell personnel lock, a personnel hatch located in the drywell ceiling, and the drywell equipment hatch. The personnel lock consists of an interlocked, double-door, welded steel assembly. Each door is equipped with a valve for equalizing pressure across the door such that the doors are not operable unless the pressure is equalized.

The two doors in the personnel lock are interlocked to prevent both being opened simultaneously, and to ensure that one door is completely closed before the opposite door can be opened. An emergency lighting and communication system operating from an external auxiliary energy source is provided within the personnel lock interior.

The personnel hatch located in the drywell ceiling consists of a double-gasketed bolted flange.

The equipment hatch is fabricated from welded steel and furnished with a double-gasketed flange and bolted, dished door. Provision is made to pressure test the space between the double gaskets of the door flanges. A shield wall is provided with the same shielding requirements as the drywell wall.

3.2.4 Access for Refueling Operations The drywell head is removed during refueling operations. This head is held in place by bolts and sealed with a double seal. It is opened only when the primary coolant temperature is below 200°F and the core is sub-critical. The double seal provides a method for determining the leak tightness of the seal without pressurizing the drywell.

3.2.5 Suppression Pool Weir Wall The suppression pool weir wall, located inside the drywell, acts as the inner boundary of the suppression pool. It is constructed of reinforced concrete and extends from the outer edge of the drywell sump floor. The weir wall is lined with 1/4-inch stainless steel plate on the suppression pool side to protect the concrete from demineralized water.

The principal dimensions of the weir wall are:

• Inside diameter: 61 feet;

• Wall thickness: 1 foot 10 inches;

• Height above basemat: 23 feet 9 inches; and

• Height above sump floor: 12 feet 7 1/4 inches.

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ATTACHMENT 1 Evaluation of Proposed Change 3.2.6 Process Pipe Tunnel The process pipe tunnel provides shielding for the process piping between the drywell and the containment. It is designed as an integral part of the drywell structure and is constructed of reinforced concrete. The arrangement at the containment wall permits differential movement between the tunnel and the containment. Doorways connect the tunnel to the containment volume.

3.2.7 Drywell Sump Floor The drywell sump floor is a thick slab of reinforced concrete which rests on the basemat and supports the suppression pool weir wall and the reactor pedestal. A stainless steel liner is provided on the suppression pool side to protect the concrete from demineralized water.

The sump floor has the following principal dimensions:

• inside diameter: 18 feet 6 inches;

• outside diameter: 64 feet 8 inches; and

• thickness: 11 feet 1 3/4 inches.

3.3 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. Appendix J also ensures that periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment and 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. Appendix J identifies three types of required tests: (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 leakage rates. Type 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 Type 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.

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ATTACHMENT 1 Evaluation of Proposed Change Also in 1995, RG 1.163 (Reference 3) was issued. The RG endorsed NEI 94-01, Revision 0, (Reference 4) 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 5) and Electric Power Research Institute (EPRI) TR-104285 (Reference 6) 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 was 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 7), 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) 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 (Reference 3). 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.

On December 8, 2008, the NRC issued Regulatory Issue Summary (RIS) 2008-27, "Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50" (Reference 8). The RIS clarifies the NRC position concerning licensee requests to extend Type A test intervals beyond 15 years, stating that a licensee can commence the test no later than the last day of the month in which it becomes due, without seeking NRC approval through a license amendment. The RIS also endorses the statement made in NEI 94-01, Revision 2, that if the test interval ends while primary containment integrity is not required, or is required solely for shutdown activities, the test interval may be extended indefinitely, but a Type A test shall be completed prior to entering the operating mode requiring primary containment integrity. In addition, the RIS states that any extension beyond the end of the month in which the test is due would require a license amendment request. Per RIS 2008-27, the license amendment request should demonstrate:

• a sound technical justification and/or undue hardship or unusual difficulty;

• the requested amendment poses minimal safety risk;

• acceptable plant-specific containment performance, including a plant-specific risk-informed analysis; and

• that containment does not have a history of significant degradation issues.

In 2012, NEI 94-01, Revision 3-A (Reference 9), 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 SEs of June 25, 2008 (Reference 10) and June 8, 2012 (Reference 11) as an acceptable methodology for complying Page 9

ATTACHMENT 1 Evaluation of Proposed Change with the provisions of Option B in 10 CFR 50, Appendix J. The regulatory positions stated in RG 1.163 as modified by References 10 and 11 are incorporated in this document. 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.

The NRC has provided 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 SE Section 3.1.1.2 which states, in part:

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.

The NRC has also provided the following concerning the extension of ILRT intervals to 15 years in NEI 94-01, Revision 3-A, NRC SE Section 4.0, Condition 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.

3.4 Current CPS ILRT Requirements 10 CFR 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." On June 21, 1996, the NRC issued Amendment 105 for CPS (Reference 12) authorizing the implementation of 10 CFR 50, Appendix J, Option B for Type A, B and C tests with the following exemptions from the requirement of 10 CFR 50, Appendix J - Option B, paragraph III.B:

• Exempting the measured leakage rates from the main steam isolation valves from inclusion in the combined leak rate for local leak rate tests; and

• Exempting leakage from the valve packing and the body-to-bonnet seal of valve 1E51-F374 associated with containment penetration 1MC-44 from inclusion in the combined leakage rate for penetrations and valves subject to Type B and C tests.

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ATTACHMENT 1 Evaluation of Proposed Change In addition to the above exemptions, the following, previously approved exemptions from the requirements of 10 CFR 50, Appendix J - Option A were determined to be no longer applicable:

• exemption from paragraph III.D.2(b)(ii) to permit substituting the air lock door seal leakage for the entire primary containment air lock test; and

• exemption from paragraph III.D.1(a) pertaining to the requirement to conduct the third Type A test during the last outage within the 10-year inservice inspection interval.

On September 26, 2017, the NRC issued Amendment 214 for CPS (Reference 13). This amendment approved a permanent extension of the Type A ILRT frequency from once every 10 years to once every 15 years in accordance with NEI 94-01, Revision 3.A, and the conditions and limitations specified in NEI 94-01, Revision 2-A. The amendment also approved an extension of the containment isolation valve leakage rate testing frequency from 60 months to 75 months for Type C leakage rate testing of selected components in accordance with NEI 94-01, Revision 3-A.

With the issuance of Amendment 214, the due date for the CPS Type A test moved from February 2018 to February 2023. The proposed change would defer the Type A test for CPS until no later than October 2023. This will allow performance of the test during the fall 2023 refueling outage, which is currently scheduled to begin in September 2023. This represents an extension of approximately eight months. The intervals for the Type B and Type C tests at CPS would remain unchanged at 120 months and 75 months, respectively.

3.5 Integrated Leakage Rate Testing History Previous CPS ILRT results have confirmed that the containment is acceptable, with considerable margin, with respect to the TS acceptance criterion of 0.65% of primary containment air weight per day at the design basis loss-of-coolant accident pressure. Since the last two Type A test results meet the performance leakage rate criteria from NEI 94-01, Revision 3-A (Reference 9), a test frequency of 15 years would be acceptable. Table 3.5-1 lists the past periodic Type A ILRT results for CPS.

Table 3.5-1, CPS Type A ILRT History Test Date As-Found Leakage Rate As-Left Leakage Rate (Containment air (Containment air weight %/day) weight %/day)

January 1986 0.2930 0.3463 (Preoperational) (1)

November 1986 0.2875 0.2933 (Preoperational)

February 1991 0.2209 0.2291 November 1993 0.2089 0.2204 February 2008 0.2708 0.226 Page 11

ATTACHMENT 1 Evaluation of Proposed Change (1) Subsequent to the ILRT, a hole was discovered through the containment liner plate. This hole was evidently in existence during the ILRT and was created by the removal of temporary attachment CL-J-12-4. After repair, the hole was retested utilizing a leak chase channel and bubbler. The retest showed zero leakage. ILRT was re-performed in November 1986.

No modifications that require a Type A test are planned at CPS prior to the fall 2023 refueling outage, when the next Type A test will be performed in accordance with this proposed change.

Any unplanned modifications to containment prior to the next scheduled Type A test would be subject to the special testing requirements of Section IV.A of 10 CFR 50, Appendix J. There have been no pressure or temperature excursions in the CPS containment which could have adversely affected containment integrity. There is no anticipated addition or removal of plant hardware within containment which could affect leak-tightness.

3.6 Plant Specific Confirmatory Analysis 3.6.1 Methodology An evaluation has been performed to provide an assessment of the risk associated with implementing a one-time extension of the CPS containment Type A ILRT interval by approximately eight months from the currently approved value of 15 years to 15.7 years. The risk assessment follows the guidelines from NEI 94-01, Revision 3-A (Reference 9), the methodology outlined in Electric Power Research Institute (EPRI) Topical Report (TR) 104285 (Reference 6), as updated by the EPRI Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (EPRI TR-1018243) (Reference 14), the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a request for a change to a plant's licensing basis as outlined in Regulatory Guide (RG) 1.174 (Reference 15), and the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage going undetected during the extended test interval (Reference 16).

3.6.2 Summary of Plant-Specific Risk Assessment Results The risk assessment determined that increasing the ILRT interval on a one-time basis to a one-in-15.7-year frequency is not considered to be significant since it represents only a small change in the CPS risk profile. Details of the CPS risk assessment are contained in to this letter. The plant-specific results for a one-time extension of the CPS ILRT surveillance interval from the current 15 years to 15.7 years are summarized below.

• RG 1.174 (Reference 15) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines "very small" changes in risk as resulting in increases of core damage frequency (CDF) below 1E-6/yr and increases in large early release frequency (LERF) below 1E-7/yr. "Small" changes in risk are defined as increases in CDF below 1E-5/yr and increases in LERF below 1E-6/yr. Since the ILRT extension has no impact on CDF for CPS, the relevant criterion is LERF. The increase in internal events LERF resulting from a change in the Type A ILRT interval with corrosion included is 9.80E-10/yr, which falls within the "very small" change region of the acceptance guidelines in RG 1.174.

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ATTACHMENT 1 Evaluation of Proposed Change

• The change in dose risk for changing the Type A ILRT interval from once-per-15 years to once-per-15.7-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.78E-04 person-rem/yr using the EPRI guidance with the corrosion included. NEI 94-01 (Reference 4) 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.

• The increase in the conditional containment failure probability from the 15-year interval to the 15.7-year interval including corrosion effects using the EPRI guidance is 0.03%,

which is less than the acceptance criteria of 1.5%.

• To determine the potential impact from external events, an additional assessment from the risk associated with external events was performed. The total increase in LERF due to internal events and external events is 4.22E-08/yr, which falls within the "very small" change region of the acceptance guidelines in RG 1.174.

3.7 Non-Risk Based Assessment Consistent with the defense-in-depth philosophy discussed in RG 1.174, CPS has assessed other non-risk based considerations relevant to the proposed amendment. CPS has multiple inspections and testing programs that ensure the containment structure remains capable of meeting its design functions and that are designed to identify any degrading conditions that might affect that capability. These programs are discussed below.

3.7.1 CPS Protective Coating Program CPS has committed to follow NRC RG 1.54, "Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," Revision 0. The RG describes a method to comply with requirements of Appendix B to 10 CFR 50, and invokes several ANSI Standards. Standards pertinent to coatings are ANSI N101.2, "Protective Coatings (Paints) for Light Water Nuclear Reactor Containment Facilities," ANSI N101.4, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities," and ANSI N5.12, "Protective Coatings for the Nuclear Industry." (ANSI N5.9, referenced in ANSI N101.2, was replaced by ANSI N5.12 in 1974, prior to CPS obtaining a construction permit).

A program to maintain containment coatings was developed to meet the requirements of Regulatory Guide 1.54, Revision 0. This program is implemented using CPS Procedure 1080.01, "CPS Protective Coating Program." Every refueling outage (i.e., every 24 months), a preventive maintenance activity to inspect the protective coatings in the containment building, including the drywell, is performed.

The most recent inspection was performed during a refueling outage (i.e., C1R19) in September 2019. During outage C1R19, all elevations of the Drywell Liner Plate, Inner Wall and floors, Containment Liner Plate, Inner Wall and Floors and Containment Steam Tunnel were inspected to identify degraded coatings.

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ATTACHMENT 1 Evaluation of Proposed Change There will be no change to the schedule for these inspections as a result of the extended ILRT interval.

Unqualified/Degraded Coatings in Containment CPS 1080.01 requires that a visual inspection be performed to establish baseline condition of Containment and Drywell coatings, and to identify Unqualified and Degraded coatings. The baseline condition is the first issue of the Combined Degraded and Unqualified Coatings list. It was completed on August 19, 1998.

As of September 2019, there are 2,303.07 pounds of combined Degraded and Unqualified coatings. Allowed is 20,000 pounds per Engineering Evaluation EE-00-143, Rev. 0. Therefore, 2,303.07 pounds of combined Degraded and Unqualified coatings would not be of concern during a LOCA.

3.7.2 Containment Inservice Inspection Program (CISI)

CPS performs a comprehensive primary containment inspection to the requirements of American Society of Mechanical Engineers (ASME) Section XI, "Inservice Inspection,"

Subsections IWE, "Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants," and Subsection IWL, "Requirements of Class CC Concrete Components of Light-Water Cooled Power Plants." The CPS Containment Inservice Inspection Program began development in 1996 and the initial inspections were completed in September 2001. The components subject to Subsection IWE and IWL requirements are those which make up the containment structure, its leak-tight barrier (including integral attachments),

and those that contribute to its structural integrity. Specifically included are Class MC pressure retaining components, including metallic shell and penetration liners of Class CC pressure retaining components, and their integral attachments. The ASME Code Inspection Plan was developed in accordance with the requirements of the 1992 Edition with the 1992 Addenda of the ASME Boiler and Pressure Vessel Code, Section XI, Division 1, Subsections IWE and IWL, as modified by NRC final rulemaking to 10 CFR 50.55a published in the Federal Register on August 8, 1996.

The initial inspections of the CPS metal/concrete containment have been completed. Various indications were observed, documented, and evaluated and determined to be acceptable. No areas of the containment liner surfaces require augmented examination. No loss of structural integrity of primary containment was observed.

There will be no change to the schedule for these inspections as a result of the extended ILRT interval. Inspection period dates for the 2nd and 3rd ClSI inspection intervals are displayed in Tables 3.7.2-1 and 3.7.2-2.

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ATTACHMENT 1 Evaluation of Proposed Change Table 3.7.2-1, CPS IWE Examination Schedule 2nd and 3rd Ten-Year Inspection Interval 2nd Interval 9/10/08 to 9/9/11 1st Period C1R12 2nd Interval 9/10/11 to 9/9/15 2nd Period C1R13 C1R14 C1R15 2nd Interval 9/10/15 to 9/9/18 3rd Period C1R16 C1R17 C1R18 3rd Interval 9/10/18 to 9/9/21 1st Period C1R19 3rd Interval 9/10/21 to 9/9/25 2nd Period C1R20 C1R21 3rd Interval 9/10/25 to 9/9/28 3rd Period C1R22 C1R23 Table 3.7.2-2, CPS IWL Examination Schedule 2nd and 3rd Ten-Year Inspection Interval C1R12 - 2010 C1R15 - 2015 C1R19 - 2019 C1R22 - 2025 Edition and Addenda of the ASME Section XI Code CPS is committed to the following Editions and Addenda of the ASME Section XI Code.

• American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Rules for lnservice Inspection of Nuclear Power Plant Component, 2013 Edition with no Addenda (hereafter referred to as the ASME Section XI Code).

• The applicable requirements of Subsection IWA (General Requirements), Subsection IWE (Requirements for Class MC and Metallic Liners of Class CC Components), and Subsection IWL (Requirements for Class CC Concrete Components) of the 2013 Edition with no Addenda and the ASME Section XI Code shall apply to components and items classified as ASME Code Class MC or ASME Code Class CC.

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ATTACHMENT 1 Evaluation of Proposed Change

• In addition to the requirements of the ASME Section XI Code, the applicable modifications and limitations outlined in 10 CFR 50.55a(b)(2)(viii) and 50.55a(b)(2)(ix) shall also be implemented.

Code Cases There are no Code Cases implemented at this time.

Relief Requests There are no Relief Requests implemented at this time.

Identification of Class MC and/or CC Exempt Components Table 3.7.2-3, Class MC and/or CC Exempt Components Exam Category Item Number Description Applicability to CPS E-A E1.30 Moisture Barriers Not Applicable L-B L2.10 Tendon Not Applicable L2.20 Wire or Strand Not Applicable L2.30 Anchorage Hardware and Not Applicable Surrounding Concrete L2.40 Corrosion Protection Medium Not Applicable L2.50 Free Water Not Applicable Augmented Inspections Table 3.7.2-4, Augmented Inspections Exam Category Item Number Description Total Number of Components E-C E4.11 Visible Surfaces 0 Containment E4.12 Surface Area Grid 0 Surfaces Requiring Minimum Wall Augmented Thickness Location Examination L-A L1.12 Suspect Areas 0 Concrete Page 16

ATTACHMENT 1 Evaluation of Proposed Change Inaccessible Areas For Class MC applications, CPS shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, CPS shall provide the following in the ISI summary report, as required by 10 CFR 50.55a(b)(2)(ix)(A):

• A description of the type and estimated extent of degradation, and the conditions that led to the degradation;

• An evaluation of each area, and the result of the evaluation; and

• A description of necessary corrective actions.

In addition, for Class CC applications, for each inaccessible area of concrete identified for evaluation under IWL-2512(a), or identified as susceptible to deterioration under IWL-2512(b),

CPS shall provide the following in the ISI summary report, as required by 10 CFR 50.55a(b)(2)(viii)(E):

• A description of the type and estimated extent of degradation, and the conditions that led to the degradation;

• An evaluation of each area, and the result of the evaluation; and

• A description of necessary corrective actions.

The most recent ISI summary report for CPS was submitted to the NRC in Reference 17. As stated in Reference 17, all Class MC components required to be examined, and all Class CC components, were scheduled and examined in C1R19. No unacceptable indications were identified.

CPS has not needed to implement any new technologies to perform inspections of any inaccessible areas at this time. However, EGC actively participates in various nuclear utility owner's groups and ASME Code committees to maintain cognizance of ongoing developments within the nuclear industry. Industry operating experience is also continuously reviewed to determine its applicably to CPS. Adjustments to inspection plans and availability of new, commercially available technologies for the examination of the inaccessible areas of the containment would be explored and considered as part of these activities.

3.7.3 Supplemental Inspection Requirements TS 5.5.13 requires, in part, that a primary containment leakage rate testing program be established in accordance with the guidelines contained in NEI 94-01, Revision 3-A. This requires that a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity be conducted. This inspection must be conducted prior to each Type A test and during at least three (3) other outages before the next Type A test if the interval for the Type A test has been extended to 15 years in accordance with the following sections of NEI 94-01, Revision 3-A:

• Section 9.2.1, "Pretest Inspection and Test Methodology"

• Section 9.2.3.2, "Supplemental Inspection Requirements" Page 17

ATTACHMENT 1 Evaluation of Proposed Change In addition to the inspections performed by the IWE/IWL Containment Inspection Program, procedure CPS 9861.01, "Integrated Leak Rate Test," requires that the structural integrity of the exposed accessible interior and exterior surfaces of the drywell and the containment, including the liner plate, shall be determined by a visual inspection of those surfaces prior to the Type A Containment Leak Rate Test. This inspection also fulfills the surveillance requirement of TS SR 3.6.1.1.1 and NEI 94-01.

3.7.4 Type B and Type C Testing Programs The CPS Type B and Type C testing programs require testing of electrical penetrations, airlocks, hatches, flanges, and containment isolation valves 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 TS 5.5.13, the allowable maximum pathway total Types B and C leakage is 0.6 La where La equals approximately 361,277 sccm.

As discussed in NUREG-1493 (Reference 5), Type B and Type C tests can identify the vast majority of all potential containment leakage paths. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

A review of the Type B and Type C test results from 2008 through 2019 for CPS has shown substantial margin between the actual As-Found (AF) and As-Left (AL) outage summations and the regulatory requirements. Table 3.7.4-1 provides local leak rate test (LLRT) data trend summaries for CPS Unit 1 inclusive of the CPS 2008 ILRT.

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ATTACHMENT 1 Evaluation of Proposed Change Table 3.7.4-1, CPS Type B and C LLRT Combined As-Found/As-Left Trend Summary As-Found As-Found As-Left As-Left RFO Min Path Min Path Max Path Max Path (sccm) as % of La (sccm) as % of La C1R11 11597.7 3.21 24924.6 6.89 2008 C1R12 18425.3 5.1 36848.3 10.19 2010 C1R13 31551.3 8.73 52689.2 14.58 2011 C1R14 33798.1 9.35 66201.4 18.32 2013 C1R15 35305.4 9.77 66465.1 18.39 2015 C1R16 35389.95 9.79 66867.6 18.51 2016 C1R17 31361.3 8.68 66663.7 18.45 2017 C1R18 32931.69 9.11 66270.9 18.34 2018 C1R19 36982.85 10.24 65937.7 18.25 2019 3.8 Results of Recent Inspections 3.8.1 Containment and Drywell Coatings - C1R19 - 2019 3.8.1.1 Drywell 723' Elevation On the 723' elevation, the condition of the Liner Plate in the Drywell basement is in good condition. The liner plate was originally coated with an inorganic zinc primer and top coated with a phenolic epoxy. Equipment and scaffolding transport during outages have resulted in the most common cause of impact damage to the containment floor and liner plate. This is a normal coating maintenance issue that does not generally impact operability. No identifiable additional areas of mechanical damage or coating defects were observed from what was reported in the previous outage report. Generally, the coatings at the concrete wall, liner plate, and existing steel are in good condition. However, the coating on the concrete drywell floor is in Page 19

ATTACHMENT 1 Evaluation of Proposed Change poor condition in limited areas. The degraded coating has been assessed and determined to support operability. Future repairs are being considered.

737' Elevation The liner plate on this elevation exhibits a few areas of damaged topcoat coating flaking and areas of mechanical damage. During the inspection, areas of light rusting on components and supports were also noted.

768' Elevation On the 768' elevation, the liner plate is in good condition. Less than a dozen areas of coating defects were found on this elevation. These areas are typical mechanical damaged areas caused by scaffolding and equipment movements during refueling outages. Other defects were from testing areas performed and not repaired or from supports and welding during previous outages and not repaired or repaired improperly. The liner coating is tightly adhering and intact throughout. Handrails on this elevation and throughout the drywell, exhibit mechanical damage but the remaining coating that is still in place appears to be tightly adhered.

3.8.1.2 Containment 737' Elevation and 755' Elevation During the inspection on Elevation 737' and Elevation 755' the liner plate, floor, piping and inner wall exhibited numerous areas of minor mechanical damage. The design of the BWR/6 Mark III containment makes it harder to inspect the liner plate closely on the suppression pool wall. The suppression pool wall is made of stainless-steel material. The liner plate on Elevation 737' is approximately 15 feet away. On this elevation, there are areas of cracked and flaked coating on the liner plate. On Elevation 755', flaking and cracking are present on the liner plate and concrete inner wall. The floor at this elevation has mechanical damage due to wear and traffic.

Plating on the floor at this elevation exhibits surface rust.

778' Elevation and 803' Elevation On Elevation 778' and 803' the liner plate, supports and valve bodies and inner wall exhibit numerous areas of minor mechanical damage. On elevation 778', areas of mechanical damage, flaking and cracking are present on the liner plate and concrete inner wall. On both elevations, the degraded areas that are cracking and flaking show signs of corrosion with rust coming through the cracks. On elevation 803', several areas of mechanical damage on the inner concrete wall were repaired with new coatings. Overall, for these elevations, the liner plate and inner wall are in good condition.

828' Elevation Most of the containment floor is usually covered with plastic for FME and to facilitate refuel services activities during outages and as a result, it was not feasible to inspect. However, the area is usually inspected when online. No concerns or issues about coatings degradation on this elevation were identified. The dome was remotely observed from the 828' floor elevation by Page 20

ATTACHMENT 1 Evaluation of Proposed Change UESI level III coatings inspectors with the plant online in 2018, but no issues or concerns were identified.

Containment Steam Tunnel In the Containment Steam Tunnel, the liner plate, floors and piping exhibit a few areas of minor mechanical damage. Areas of mechanical damage and cracking are present on the liner plate.

The degraded areas on the liner plate that are cracking show signs of corrosion with rust coming through. Mechanical damage also has corrosion to substrate on the liner plate. Overall, the liner plate and inner wall are in good condition.

3.8.1.3 Conclusion The coating assessment identified areas of coating degradation requiring repair. Importantly, no current coating conditions were identified that impact structural integrity, plant operations, or the safe shutdown of the plant. Many of the degraded areas have been identified in previous reports and are repaired by the CPS Protective Coating Program to protect surfaces and equipment from contamination and corrosion. The objective of the CPS Protective Coating Program is to protect plant systems, structures and components from degradation by applying and maintaining protective coatings. The program assures that the station shall be clean, neat and easily maintained from the aspect of personnel safety, housekeeping, and radiological control. Timely repairing of degraded coatings will continue to maintain higher coatings margin for the safe operability of the suction strainers in the suppression pool. The routine coating assessment during outages will continue to be performed.

3.8.2 CISI Program Inspection Results 3.8.2.1 CISI on Containment Liner - C1R19 - 2019 737' Elevation CISI of the containment liner was performed on the 737' elevation during C1R19 (September 2019) for ASME Section XI IWE inspection requirements. The following observations were documented.

• At 114 degrees, 748' elevation, two 6" diameter peeled, cracked, and stained coatings were identified.

• At 153 degrees, 751' elevation, 4" x 4" peeled, cracked, and stained coatings were identified.

• At 157 degrees, 748' elevation, 1" x 2" cracked and stained coatings were identified.

• At 245 degrees, 738' elevation, two areas 2" x 3" each, missing and cracked paint was identified.

• At 255 degrees, 740' elevation, 5" x 12" area was missing paint.

• At 268 degrees, 735' elevation, 3" x 18'"cracked and stained coatings were identified.

• At 269 degrees, 737' elevation, 1" x 6" cracked and stained coatings were identified.

• At 269 degrees, 735' elevation, 3" x 3" peeled and stained coatings were identified.

• At 273 degrees, 743' elevation, 8" x 12" peeled, cracked, and stained coatings were identified.

• At 308 degrees, 746' elevation, 6" x 3" cracked and stained coatings were identified.

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ATTACHMENT 1 Evaluation of Proposed Change

• At 316 degrees, 746' elevation, 12" x 3" cracked and stained coatings were identified.

• At 338 degrees, 746' elevation, 8" x 8" cracked and stained coatings were identified.

• At 340 degrees, 748' elevation, four areas 4" x 4" each, missing and cracked paint was identified.

• At 360 degrees, 745' elevation, 6" x 6" cracked and stained coatings were identified.

755' Elevation CISI of the containment liner was performed on the 755' elevation during C1R19 (September 2019) for ASME Section XI IWE inspection requirements. The following observations were documented.

• At 8, 12, and 20 degrees, 775' elevation, 6" x 6" no coating and rust were identified.

• At 24 degrees, 776' elevation, 1" x 2" peeled, cracked, and stained coatings were identified.

• At 25 degrees, 756' elevation, 3' x 1' no coating and rust were identified.

• At 70 degrees, 774' elevation, 4" x 4' peeled, cracked, and stained coatings were identified.

• At 96 degrees, 757' elevation, 8" x 8" peeled, cracked, and stained coatings were identified.

• At 192 degrees, 767' elevation, two 1" diameter peeled, cracked, and stained coatings were identified.

• At 200 degrees, 769' elevation, 1" x 2" peeled and stained coatings were identified.

• At 266 degrees, 758' elevation, three small areas cracked and stained coatings were identified.

781' Elevation CISI of the containment liner was performed on the 781' elevation during C1R19 (September 2019) for ASME Section XI IWE inspection requirements. The following observations were documented.

• At 27 degrees, 780' elevation, 5" x 14" area missing paint was identified.

• At 29 degrees, 777' elevation, four areas 3" x 3" each, missing paint and discoloration were identified.

• At 31 degrees, 798' elevation, 10" x 10" area cracked paint and discoloration were identified.

• At 32 degrees, 786' elevation, 4" x 3' peeled and stained coatings were identified.

• At 46 degrees, 781' elevation, 4" x 4' peeled and stained coatings were identified.

• At 72 degrees, 781 '-784' elevation, three areas 8" x 8" of cracked and stained coatings were identified.

• At 90 degrees, 781' elevation, 2" x 2", 2" x 4", and 2" x 8" areas of cracked and stained coatings were identified.

• At 90 degrees, 790' elevation, 8" x 8" area cracked paint and discoloration were identified.

• At 93 degrees, 798' elevation, 4" x 16' cracked and stained coatings were identified.

• At 93 degrees, 786' - 794' elevation, small areas and a 6" x 6' peeled and stained coatings were identified.

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ATTACHMENT 1 Evaluation of Proposed Change

• At 95 degrees, 791' elevation, small areas and 6" x 12" area of cracked and stained coatings, along with a small area of peeled and stained coatings were identified.

• At 125 degrees, 785' elevation, 6" x 24" peeled and stained coatings were identified.

• At 130 degrees, 794' elevation, 8" x 8" cracked and stained coatings were identified.

• At 228 degrees, 788' elevation, 3" x 3' small area of peeled coatings were identified.

• At 265 degrees, 786' -790' elevation, three small areas of cracked and stained coatings were identified.

• Penetration 1MC-50 at 790' elevation, identified cracks in the coatings on the outer ring, at 2:00 and between 10:00 and 11:00. Minor discoloration coming from cracks in coatings.

800' Elevation CISI of the containment liner was performed on the 800' elevation during C1R19 (September 2019) for ASME Section XI IWE inspection requirements. The following observations were documented.

• At 28 degrees, 824' elevation, 8" x 8" area of cracked coatings were identified.

• At 32 degrees, 803' elevation, six areas 6" x 6" of cracked paint, missing paint, and discoloration were identified.

• At 32 degrees, 822' elevation, 4" x 6" area of cracked paint and discoloration were identified.

• At 35 degrees, 822' elevation, 4" x 12" area of cracked paint and discoloration were identified.

• At 42 degrees, 803' elevation, three areas 2" x 3" of missing and cracked paint were identified.

• At 85 degrees, 815' elevation, 8" x 8" area of peeled coatings were identified.

• At 92 degrees, 806' elevation, small areas had blistered and stained coatings identified.

• At 120 degrees, 809' elevation, small areas of cracked and stained coatings were identified.

• At 131 degrees, 818' elevation, 4" x 6" area missing paint was identified.

• At 135 degrees, 813' elevation, 3" x 3" area missing paint and discoloration were identified.

• At 190 degrees, 823' elevation, stained coatings were identified.

• At 193 degrees, 806' elevation, nine areas 2" x 2" each of missing and cracked paint were identified.

• At 235 degrees, 817' elevation, 6" x 6" areas of cracked and stained coatings were identified.

• At 235 degrees, 817' elevation, 6" x 8" areas of cracked and stained coatings were identified.

• At 257 degrees, 803'-805' elevation, small areas of cracked, stained, and damaged coatings were identified.

• At 268 degrees, 807' elevation, small areas of cracked and stained coatings were identified.

• At 269-276 degrees, 804'-811' elevation, approximately thirty various areas of cracked and stained coatings were identified.

• At 276 degrees, 812'-816' elevation, two areas of cracked and stained coatings were identified.

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ATTACHMENT 1 Evaluation of Proposed Change

• At 278 degrees, 817' elevation, 6" x 12" area of cracked and stained coatings were identified.

• At 290 degrees, 813' elevation, 12" x 15" area no coating and mild corrosion were identified.

• At 290-300 degrees, 811'-815' elevation, 2" x 2" and 6" x 6" areas of cracked and stained coatings were identified.

• At 300-310 degrees, 811'-815' elevation, three 2" x 2" areas and a 6" x 6" area of cracked and stained coatings were identified.

• At 310-320 degrees, 806'-815' elevation, 11 small areas of cracked and stained coatings were identified.

• At 320-330 degrees, 811'-827' elevation, approximately 20 small areas of cracked and stained coatings were identified.

828' Elevation CISI of the containment liner was performed on the 828' elevation during C1R19 (September 2019) for ASME Section XI IWE inspection requirements. The following observations were documented.

• At 5 degrees, 840' elevation, four 6" x 6" areas of discolored and cracked paint were identified.

• At 230 degrees, 848' elevation, three areas with peeling paint were identified.

• At 290 degrees, 832' elevation, one square foot area of discoloration and flaking were identified.

Containment Dome CISI of the containment dome from inside containment was performed during C1R19 (September 2019) for ASME Section XI IWE inspection requirements. The following observations were documented.

• At 30 degrees, 858' elevation, three areas of peeling paint were identified.

• At 40 degrees, 890' elevation, 8" x 12" area of cracked and stained paint were identified.

• At 290 degrees, 860' elevation, four 8" x 8" areas of missing and cracked paint with coating discoloration were identified.

• Various areas of discolored paint with cracking and missing paint were identified. The indications are located on upper portion of the dome above Containment Spray.

3.8.2.2 CISI on Concrete Containment - C1R19 - 2019 CISI of the concrete containment was performed during C1R19 (September 2019) for ASME Section XI IWL inspection requirements. The following observations were documented.

• OD Concrete - 707' AB, 270 - 90 deg. Concrete coatings were satisfactory except for two areas: (1) Az. 285° El. 707' has flaking and peeling paint. (2) Az. 293° El. 710' has peeling paint.

• OD Concrete - 712' FB.90-270 deg. Satisfactory. No recordable indication.

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ATTACHMENT 1 Evaluation of Proposed Change

• OD Concrete - Reactor Water Cleanup Mezz. 750' AB. 40 - 90 deg. Satisfactory. No recordable indication.

• OD Concrete - FB 737', 90 - 270 deg. Satisfactory. No recordable indication.

• OD Concrete - AB 737'. 270- 90 deg. Satisfactory. No recordable indication.

• OD Concrete - AB 762'. 270 - 90 deg. Concrete and coatings were satisfactory except for two areas: (1) Az. 290° El. 770' has flaking paint. (2) Az. 285° El. 785' has flaking paint.

• OD Concrete - FB 755', 90 -270 deg. Satisfactory. No recordable indication.

• OD Concrete - AB 781', and AB Steam Tunnel. Upper Aux Steam Tunnel concrete is satisfactory. Recordable indications are: (1) missing surface coating and coating damage between feedwater penetrations and around penetration 1MC-14, apparent moisture damage, and (2) flaking paint on Az. 80°, El. 780' and Az. 85°, El. 785'.

• OD Concrete - FB 781', 90 -270 deg. Satisfactory. No recordable indication.

• OD Concrete - Gas Control Boundary 800' and above AB and FB 0-360 deg.

Satisfactory. No recordable indication.

3.8.2.3 CISI on Penetrations from Inside and Outside Containment - C1R19 - 2019 CISI of containment penetrations, from inside and outside containment, was performed during C1R19 (September 2019) for ASME Section XI IWE inspection requirements. The following observations were documented.

From Inside Containment

• Penetration IMC-202, 769' elevation, 82 degrees: flaking, peeling, coating damaged

• Penetration IMC-42, 756' elevation, 0 degrees: flaking, light rust, and coating damaged

• Penetration IMC-45, 756' elevation, 0 degrees: flaking, light rust, and coating damaged

• Penetration IMC-50, 790' elevation, 95 degrees: flaking, coating damaged, and discoloration

• Penetration IMC-114, 780' elevation, 51 degrees: peeling, blistering, and missing paint or coating From Outside Containment All containment penetrations from outside were inspected utilizing VT-3 inspectors. Several indications of minor significance were identified, such as paint or coating damage, flaking, light rust, missing paint, peeling, corrosion, cracked paint, tear, and discoloration. There was no apparent loss of material identified.

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ATTACHMENT 1 Evaluation of Proposed Change 3.8.2.4 Containment Liner, IWE - Wetted Surfaces of Submerged Areas and BWR Vent System - C1R18 - 2018 In accordance with the 2nd 10-Year Inspection Interval the scheduled examinations for ASME Section XI, IWE, wetted surfaces of submerged areas and BWR vent system were completed during outage C1R18 (May 2018). This inspection was conducted for accessible areas in accordance with ER-AA-335-018, "Visual Examination of ASME IWE Class MC and Metallic Liners of IWL Class CC Components." All results were acceptable.

3.8.2.5 Conclusion The above observations identified through CISI inspections on the containment liner, concrete containment, and containment penetrations (i.e., from inside and outside containment) were evaluated. There is no adverse effect on the metallic liner with respect to the design intent and overall structural integrity of the liner system. The conditions identified above are considered acceptable.

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, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."

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 one-time extension of the frequency of the containment Type A test will not affect the design, fabrication, or construction of the containment structure, and the design will continue to account for the effects of natural phenomena. The containment Type A test will continue to be done in accordance with 10 CFR 50 Appendix J using 10 CFR 50 Appendix B quality standards.

The frequency of the containment Type A test is being changed in accordance with standards reviewed and approved as compliant with Appendix J. Therefore, there will be no instances where the applicable regulatory criteria are not met.

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ATTACHMENT 1 Evaluation of Proposed Change EGC has determined that the proposed change does not require any exemptions or relief from regulatory requirements, other than the TS, and does not affect conformance with any regulatory requirements/criteria.

4.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Facility Operating License No. NPF-62 for Clinton Power Station (CPS), Unit 1. The proposed change would allow for a one-time extension to the 15-year frequency of the CPS Unit 1 containment leakage rate test (i.e., integrated leakage rate test (ILRT) or Type A test). This test is required by Technical Specifications (TS) Section 5.5.13, "Primary Containment Leakage Rate Testing Program." The proposed one-time change would permit the current ILRT interval of 15 years to be extended by eight months.

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

EGC has evaluated the proposed change, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change involves changes to the CPS Unit 1 containment leakage rate testing program. The proposed change does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. The primary containment function is to provide an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment itself, and the testing requirements to periodically demonstrate the integrity of containment, exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve any accident precursors or initiators.

The proposed change modifies TS 5.5.13 to allow for a one-time extension to the containment Type A test interval. The potential consequences of extending the containment Type A test interval one-time by eight months have been evaluated by Page 27

ATTACHMENT 1 Evaluation of Proposed Change analyzing the resulting changes in risk. The increase in risk in terms of person-rem per year within 50 miles resulting from design basis accidents was estimated to be acceptably small and determined to be within the guidelines published in the NRC Final Safety Evaluation for NEI Topical Report (TR) 94-01, Revision 3-A. Additionally, the proposed change maintains defense-in-depth by preserving a reasonable balance among prevention of core damage, prevention of containment failure, and consequence mitigation. EGC has determined that the increase in conditional containment failure probability due to the proposed change would be very small.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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 change modifies TS 5.5.13 to allow for a one-time extension to the containment Type A test interval. Containment, and the testing requirements to periodically demonstrate the integrity of containment, exist to ensure the plant's ability to mitigate the consequences of an accident and 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) or a change to the manner in which the plant is operated or controlled.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No The proposed change modifies TS 5.5.13 to allow for a one-time extension to the containment Type A test interval. The proposed change does not alter the manner in which safety limits, limiting safety system setpoints, or limiting conditions for operation are determined. The specific requirements and conditions of the containment leakage rate testing program, as defined in the TS, ensure that the degree of primary containment structural integrity and leak-tightness that is considered in the plant's safety analysis is maintained. The overall containment leakage rate limit specified by the TS is maintained, and the Type A, Type B, and Type C containment leakage tests would be performed at the frequencies established in accordance with the NRC-accepted guidelines of NEI 94-01, Revision 3-A. Containment inspections performed in accordance with other plant programs serve to provide a high degree of assurance that containment would not degrade in a manner that is not detectable by an ILRT. A risk assessment concluded that extending the ILRT test interval one-time by eight months results in a very small change to the risk profile.

Page 28

ATTACHMENT 1 Evaluation of Proposed Change Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above evaluation, EGC concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, paragraph (c),

and accordingly, a finding of no significant hazards consideration is justified.

4.3 Conclusions 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGC 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, "Standards for Protection Against Radiation." 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, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. CPS USAR Figure 3.8-11, "Containment Building Penetrations"

2. IEEE 317, "Standard for Electrical Penetration Assemblies in Containment Structures for Nuclear Power Generating Stations," dated December 1976

3. Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,"

September 1995

4. NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," July 1995

5. NUREG-1493, "Performance-Based Containment Leak-Test Program," January 1995

6. EPRI TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals," August 1994 Page 29

ATTACHMENT 1 Evaluation of Proposed Change

7. NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," October 2008

8. NRC Regulatory Issue Summary 2008-27, "Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50," dated December 8, 2008

9. NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," July 2012

10. Letter from M. J. Maxin (NRC) to J. C. Butler (NEI), "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 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

11. Letter from S. Bahadur (NRC) to B. Bradley (NEI), "Final Safety Evaluation of Nuclear Energy Institute (NEI) Report 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J (TAC No. ME2164)," dated June 8, 2012

12. Letter from D. Pickett (NRC) to M. Lyon (CPS), Clinton Power Station Unit 1 - Issuance of Amendment 105 Regarding Implementation of 10 CFR 50, Appendix J - Option B, (TAC NO. MB95321), dated June 21, 1996

13. Letter from J. Rankin (NRC) to B. C. Hanson (Exelon Generation Company, LLC),

"Clinton Power Station, Unit No. 1 - Issuance of Amendment Regarding Permanent Extension of Type A and Type C Leak Rate Test Frequencies (CAC No. MF7290),"

dated September 26, 2017

14. Electric Power Research Institute, EPRI TR-1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325," dated October 2008

15. NRC Regulatory Guide 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

16. Letter from C. H. Cruse (Constellation Nuclear, Calvert Cliffs Nuclear Power Plant) to U.S. NRC, "Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension," dated March 27, 2002

17. Letter from J. J. Kowalski (Exelon Generation Company, LLC) to U.S. NRC, "Post Outage 90-Day Inservice Inspection (ISI) Summary Report," dated January 15, 2020 Page 30

ATTACHMENT 2 Markup of Technical Specifications Page Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 REVISED TECHNICAL SPECIFICATIONS PAGE 5.0-16a

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.13 Primary Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the primary containment as required by 10 CFR 50.54 (o) and 10 CFR 50, Appendix J, Option B, as modified by approved

, (2) the next Type A exemptions. This program shall be in accordance with the test performed after guidelines contained in NEI 94-01, "Industry Guideline for the February 2008 Implementing Performance-Based Option of 10 CFR 50, Appendix J,"

Type A test shall be Revision 3-A, dated July 2012, and the conditions and limitations performed no later specified in NEI 94-01, Revision 2-A, dated October 2008, as than October 31, modified by the following exception: (1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Type A 2023, and (3) if the tests.

Type A test has not exceptions been performed by The peak calculated containment internal pressure for the design October 31, 2023, basis loss of coolant accident, Pa, is 9.0 psig.

and the unit is in Mode 4 or 5, the The maximum allowable primary containment leakage rate La, at Pa, Type A test shall be shall be 0.65% of primary containment air weight per day.

performed prior to Leakage Rate acceptance criteria are:

entering Mode 2

a. Primary containment leakage rate acceptance criterion is d 1.0 La. During the first unit startup following testing in accordance with this program, the leak rate acceptance criteria are d 0.60 La for the Type B and Type C tests and d 0.75 La for Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is d 5 scfh when tested at t Pa,

2) For each door, leakage rate is d 5 scfh when the gap between door seals is pressurized t Pa.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Primary Containment Leakage Rate Testing Program.

(continued)

CLINTON 5.0-16a Amendment No. 214

ATTACHMENT 3 Risk Assessment for CPS Regarding the ILRT (Type A) One-Time Extension Request

Risk Management Team Clinton Power Station (CPS)

Risk Assessment for CPS Regarding the ILRT (Type A) One-Time Extension Request CL-LAR-12 Revision 2

Risk Management Team RM DOCUMENTATION NO. CL-LAR-12 REV: 2 PAGE NO. 1 STATION: Clinton Power Station (CPS)

UNIT(s) AFFECTED: Unit 1 TITLE: Risk Assessment for CPS Regarding the ILRT (Type A) One-Time Extension Request

SUMMARY

CPS is pursuing a License Amendment Request (LAR) for a one-time extension of the Integrated Leak Rate Test (ILRT) interval from 15 years to approximately 15.7 years.

The purpose of this document is to provide an assessment of the risk associated with implementing a one-time extension of the CPS Unit 1 containment ILRT interval to 15.7 years.

This is a Category I Risk Management Document in accordance with ER-AA-600-1012, which requires independent review and approval.

[ ] Review required after periodic update

[ X ] Internal RM Documentation [ ] External RM Documentation Electronic Calculation Data Files:

Microsoft Excel CL_ILRT-8month.xlsx, 2/10/2021, 10:47 AM, 590 KB Method of Review: [ X ] Detailed [ ] Alternate [ ] Review of External Document This RM documentation supersedes: N/A in its entirety.

Prepared by: Justin Sattler / see email approval / 2/19/21 Print Sign Date Reviewed by: Matt Johnson / see email approval / 2/19/21 Print Sign Date Reviewed by: Charles Standridge / see email approval / 2/19/21 App. A Only Print Sign Date Reviewed by: Victoria Warren / see email approval / 2/19/21 Print Sign Date Approved by: Eugene M. Kelly / see email approval / 2/19/21 Print Sign Date Email approvals are appended to the end of this document

Risk Impact Assessment of Extending the CPS ILRT Interval Revisions:

REV. DESCRIPTION PREPARER/DATE REVIEWER/DATE APPROVER/DATE 0 Initial issue see email approval see email approval see email approval 1 Incorporated editorial changes see email approval see email approval see email approval 2 Incorporated editorial change see email approval see email approval see email approval i CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE OF CONTENTS Section Page 1.0 PURPOSE OF ANALYSIS ................................................................................ 1-1 1.1 PURPOSE .............................................................................................. 1-1

1.2 BACKGROUND

...................................................................................... 1-1 1.3 ACCEPTANCE CRITERIA...................................................................... 1-3 2.0 METHODOLOGY .............................................................................................. 2-1 3.0 GROUND RULES ............................................................................................. 3-1 4.0 INPUTS ............................................................................................................. 4-1 4.1 GENERAL RESOURCES AVAILABLE ................................................... 4-1 4.2 PLANT SPECIFIC INPUTS..................................................................... 4-5 4.3 CPS POPULATION DOSE DERIVATION ............................................ 4-11 4.4 IMPACT OF EXTENSION ON DETECTION OF COMPONENT FAILURES THAT LEAD TO LEAKAGE (SMALL AND LARGE) ........... 4-25 4.5 IMPACT OF EXTENSION ON DETECTION OF STEEL CORROSION THAT LEADS TO LEAKAGE ................................................................ 4-28 4.6 IMPACT OF DWBT INTERVAL EXTENSION OF RELEASE CATEGORIES ...................................................................................... 4-34 5.0 RESULTS .......................................................................................................... 5-1 5.1 STEP 1 - QUANTIFY THE BASE-LINE RISK IN TERMS OF FREQUENCY PER REACTOR YEAR .................................................... 5-2 5.2 STEP 2 - DEVELOP PLANT-SPECIFIC PERSON-REM DOSE (POPULATION DOSE) PER REACTOR YEAR ...................................... 5-7 5.3 STEP 3 - EVALUATE RISK IMPACT OF EXTENDING TYPE A TEST INTERVAL FROM 15-TO-15.7 YEARS ................................................ 5-10 5.4 STEP 4 - DETERMINE THE CHANGE IN RISK IN TERMS OF LARGE EARLY RELEASE FREQUENCY ......................................................... 5-13 5.5 STEP 5 - DETERMINE THE IMPACT ON THE CONDITIONAL CONTAINMENT FAILURE PROBABILITY ........................................... 5-13 5.6

SUMMARY

OF INTERNAL EVENTS RESULTS .................................. 5-14 5.7 STEP 6 - CONTRIBUTIONS FROM EXTERNAL HAZARDS .............. 5-16 5.8 16-YEAR ILRT INTERVAL SENSITIVITY ............................................. 5-19

6.0 CONCLUSION

S ................................................................................................ 6-1

7.0 REFERENCES

.................................................................................................. 7-1 APPENDIX A PRA TECHNICAL ADEQUACY ii CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 1.0 PURPOSE OF ANALYSIS 1.1 PURPOSE The purpose of this analysis is to provide an assessment of the risk associated with implementing a one-time extension of the Clinton Power Station (CPS) containment Type A integrated leak rate test (ILRT) interval by approximately 8 months, i.e., from 15 years to 15.7 years (a value of 15.7 years is used in this analysis). The risk assessment follows the guidelines from NEI 94-01 [1], the methodology outlined in EPRI TR-104285 [2], as updated by the EPRI Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (EPRI TR-1018243) [3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a request for a change to a plants licensing basis as outlined in Regulatory Guide (RG) 1.174 [4], and the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage going undetected during the extended test interval [5]. The format of this document is consistent with the intent of the Risk Impact Assessment Template for evaluating extended integrated leak rate testing intervals provided in the EPRI TR-1018243. Consistent with other previous ILRT extension requests for BWR Mark III containments [19, 36], the risk assessment also includes the impact of extending the Drywell Bypass Test (DWBT) interval from 15 years to 15.7 years on the ILRT risk metrics. The DWBT is in the scope of the surveillance frequency control program (SFCP);

however, the DWBT has been historically associated with the ILRT frequency because the plant line-ups are similar, and the same equipment is used to perform both tests.

Although not formally requested in this submittal, change to the DWBT could impact the ILRT dose metrics because bypass leakage above nominal through the drywell to the outer containment would bypass potential fission product scrubbing in the suppression pool that are released from the reactor into the drywell.

1.2 BACKGROUND

Option B to Appendix J Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, September 1995 [6], provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consists of qualitative and quantitative 1-1 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 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 EPRI Research Project TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals [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 for a BWR plant that increasing the containment leak rate from the nominal 0.5% per day to 5% per day leads to a barely perceptible increase in total population exposure, and increasing the leak rate to 50% per day increases the total population exposure by less than 1%.

Because ILRTs represent substantial resource expenditures, it is desirable to show that extending the ILRT interval will not lead to a substantial increase in risk from containment isolation failures to support a reduction in the test frequency for CPS.

Earlier ILRT frequency extension submittals have used the EPRI TR-104285 [2]

methodology to perform the risk assessment. In October 2008, EPRI 1018243 [3] was issued to develop a generic methodology for the risk impact assessment for ILRT interval extensions to 15 years using current performance data and risk informed guidance, primarily RG 1.174 [4]. This more recent EPRI document considers the change in population dose, large early release frequency (LERF), and containment conditional failure probability (CCFP), whereas EPRI TR-104285 considered only the change in risk based on the change in population dose. This ILRT interval extension risk assessment for CPS employs the EPRI 1018243 methodology, with the affected System, Structure, or Component (SSC) being the primary containment boundary. Similar methodology is used for this analysis as was used in the 2015/2016 CPS ILRT / DWBT permanent 15-year ILRT interval extension [36] that was approved by the NRC [37].

1-2 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 1.3 ACCEPTANCE CRITERIA The acceptance guidelines in RG 1.174 [4] are used to assess the acceptability of this one-time 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 1.0E-06 per reactor year and increases in large early release frequency (LERF) less than 1.0E-07 per reactor year. Since the Type A test does not impact CDF for CPS (CPS does not credit containment accident pressure to maintain net positive suction head (NPSH) for pumps that take suction from the suppression pool; MAAP analysis performed for the permanent ILRT extension analysis demonstrates there is no impact to containment overpressure failure probability [36]), the relevant criterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 1.0E-06 per reactor year, provided that the total LERF from all contributors (including external events) can be reasonably shown to be less than 1.0E-05 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) is also calculated to help ensure that the defense-in-depth philosophy is maintained.

With regard to population dose, examinations of NUREG-1493 and Safety Evaluation Reports (SERs) for one-time interval extension (summarized in Appendix G of [3])

indicate a range of incremental increases in population dose(1) 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 0.002 to 0.46% of the total accident dose. The total doses for the spectrum of all accidents (Figure 7-2 of NUREG-1493 [6]) result in health effects that are at least two orders of magnitude less than the NRC Safety Goal Risk. Given these perspectives, the NRC SER on this issue [7] defines a small increase in population dose as an increase of 1.0 person-rem per year, or 1 % of the total population dose, (1) The one-time extensions assumed a large leak (EPRI class 3b) magnitude of 35La, whereas this analysis uses 100La.

1-3 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval whichever is less restrictive for the risk impact assessment of the extended ILRT intervals.

This definition has been adopted for the CPS analysis.

The acceptance criteria are summarized below.

1) The estimated risk increase associated with temporarily extending the ILRT surveillance interval to 15.7 years must be demonstrated to be small.

(Note that Regulatory Guide 1.174 [4] defines very small changes in risk as increases in CDF less than 1.0E-6 per reactor year and increases in LERF less than 1.0E-7 per reactor year. Since the type A ILRT does not impact CDF for CPS, the relevant risk metric is the change in LERF.

Regulatory Guide 1.174 also defines small risk increase as a change in LERF of less than 1.0E-6 reactor year.) Therefore, a small change in risk for this application is defined as a LERF increase of less than 1.0E-6.

2) The increase in population dose must be small. Per the NRC SER, a small increase in population dose is also defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1% of the total population dose, whichever is less restrictive.

3) In addition, the SER notes that a small increase in Conditional Containment Failure Probability (CCFP) should be defined as a value marginally greater than that accepted in previous one-time 15-year ILRT extension requests (typically about 1% or less, with the largest increase being 1.2%). This requires the increase in CCFP be less than or equal to 1.5% (i.e., marginally greater than 1.2%).

1-4 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 2.0 METHODOLOGY A simplified bounding analysis approach consistent with the EPRI methodology [3] is used for evaluating the one-time change in risk associated with increasing the test interval to 15.7 years. The analysis uses results from the core damage and large early release scenarios from the current CPS PRA analyses of record [24, 25] and the subsequent containment responses to establish the various fission product release categories including the release size.

The six general steps of this assessment are as follows:

1. Quantify the baseline risk in terms of the frequency of events (per reactor year) for each of the eight containment release scenario types identified in the EPRI report [3].

2. Develop plant-specific population dose rates (person-rem per reactor year) for each of the eight containment release scenario types from plant specific consequence analyses.

3. Evaluate the risk impact (i.e., the change in containment release scenario type frequency and population dose) of extending the ILRT interval from 15 years to 15.7 years.

4. Determine the change in risk in terms of LERF in accordance with RG 1.174 and compare this change with the acceptance guidelines of RG 1.174 [4].

5. Determine the impact on the CCFP.

6. Evaluate the impact of external events.

Furthermore, x Consistent with the previous industry containment leak risk assessments, the CPS assessment uses population dose as one of the risk measures.

The other risk measures used in the CPS assessment are the conditional CCFP for defense-in-depth considerations and change in LERF to demonstrate that the acceptance guidelines from RG 1.174 are met.

This evaluation for CPS uses ground rules and methods to calculate changes in the above risk metrics that are consistent with those outlined in the current EPRI methodology [3].

2-1 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 3.0 GROUND RULES The following ground rules are used in the analysis:

x The technical adequacy of the CPS Level 1 and Level 2 full power internal events (FPIE) PRA models is consistent with the requirements of Regulatory Guide 1.200 as relevant to this ILRT interval extension. See Appendix A for additional information.

x The CPS Level 1 and Level 2 FPIE PRA models provide representative core damage frequency and release category frequency distributions to be used in this analysis.

x It is appropriate to use the CPS FPIE PRA model as a gauge to effectively describe the risk change attributable to the ILRT extension. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose) would not substantially differ if other hazard groups were included in the calculations. An analysis is performed in Section 5.7 to show the effect of including external event models for the ILRT extension. The Seismic CDF and LERF estimates [38] and Fire PRA

[8] are used for this analysis.

x Dose results for the containment failures modeled in the FPIE PRA can be characterized by information provided in NUREG/CR-4551 [17]. They are estimated by scaling the NUREG/CR-4551 population dose results by power level, population, and Tech Spec leak rate differences for Clinton Power Station compared to the NUREG/CR-4551 Mark III reference plant, Grand Gulf.

x The use of the estimated 2030 population data from SECPOP version 4.2

[31], the Illinois Department of Public Health (IDPH) [30], and 2019 US Census Bureau statistics [10] are appropriate for this analysis.

x The representative containment leakage for Class 1 sequences is 1L a.

Class 3 accounts for increased leakage due to Type A inspection failures.

x The representative containment leakage for Class 3a is 10La and for Class 3b sequences is 100La, based on the recommendations in the latest EPRI report [3] and as recommended in the NRC SER on this topic [7]. It should be noted that this is more conservative than the earlier previous industry ILRT extension requests, which used 35La for the Class 3b sequences.

x Based on the EPRI methodology and the NRC SER, the Class 3b sequences are categorized as LERF and the increase in Class 3b frequencies is used as a surrogate for the LERF metric.

3-1 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval x 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 on the conclusions from this analysis will result from this separate categorization.

x The reduction in ILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal.

3-2 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 4.0 INPUTS This section summarizes the following:

x Section 4.1 General Resources Available as input x Section 4.2 Plant Specific Resources Required x Section 4.3 CPS Population Dose Derivation x Section 4.4 Details on the EPRI Methodology that is followed x Section 4.5 Details of the Calvert Cliffs corrosion analysis method that is also used as a sensitivity for this assessment x Section 4.6 Details of the analysis performed on the available Mark III DWBT data to estimate the likelihood and magnitude of DWBT leakage rates that may occur due to extending the DWBT interval in addition to the ILRT interval (note: extending the DWBT interval is not part of the proposed LAR scope).

4.1 GENERAL RESOURCES AVAILABLE Various industry studies on containment leakage risk assessment are briefly summarized here:

x NUREG/CR-3539 [13]

This study, which also used information from WASH-1400 [22], 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.

x NUREG/CR-4220 [14]

This study provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident.

x NUREG-1273 [15]

This is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the same database.

x NUREG/CR-4330 [16]

This study provides an assessment of the impact of different containment leakage rates on plant risk.

x EPRI TR-105189 [12]

This study provides an assessment of the impact on shutdown risk from ILRT test interval extension.

4-1 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval x NUREG-1493 [6]

This 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.

x EPRI TR-104285 [2]

This study is an EPRI study of the impact of extending ILRT and local leak rate test (LLRT) intervals on at-power public risk.

x NUREG-1150 [23] and NUREG/CR 4551 [29]

These two reports provide 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 CNP.

x NEI Interim Guidance [28, 35]

This 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.

x Calvert Cliffs corrosion analysis [5]

This study addresses the impact of age-related degradation of the containment liners on ILRT evaluations.

x EPRI 1018243 [3]

This 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.

x NRC Final Safety Evaluation Report [7]

This report documents the NRC staffs evaluation and acceptance of NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2.

x Prior Mark III ILRT Extension Risk Assessments [19, 20, 21]

Reference is made to other extension requests for Mark III containments that considered extensions to the ILRT interval. While the DWBT interval is not being extended as part of this analysis, accounting for the status of the drywell in the ILRT extension PRA is a surrogate for the impact to the population dose associated with an intact containment but with a suppression pool bypass via the drywell leakage.

The DWBT has been historically associated with the ILRT frequency because the plant line-ups are similar and the same equipment is used to perform both tests. The DWBT is to verify that pre-existing drywell bypass leakage does not exceed the maximum requirements. The DWBT thus affects the likelihood of a suppression pool bypass in the Level 1 and 2 PRA analyses. The methodology for extending the DWBT has previously 4-2 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval been accepted by the NRC for analysis of Clinton, Grand Gulf, and River Bend [19, 20, 21].

Even though the methodologies used for the ILRT extension do not directly address the DWBT, it is judged that the ILRT methodology can be used to address the impact of extending both the ILRT and DWBT with a few additional considerations and assumptions. The primary difference in the methodology used to evaluate the extension of the DWBT is in the determination of the conditional probability of an existing drywell leak. In the base case DWBT analysis, the same release categories, consequence calculations, and acceptance criteria are used as in the ILRT analysis. The risk analysis will be performed assuming that both the ILRT and the DWBT are on the same frequencies. The impact of drywell leakage is to allow drywell atmosphere, including fission products, to be passed at some rate directly to the containment, without benefit of quenching and fission product retention in the suppression pool.

It is assumed in this methodology that the special leakage categories established by EPRI for use in ILRT risk assessments can also be applied to the drywell for the DWBT risk assessment. The Mark III drywell, which includes the RPV, is completely enclosed by the outer containment. As such, the drywell leakage does not leak directly to the environment but is further mitigated by the outer containment leakage barrier. Because of this dual containment, there are several possible leakage path combinations that must be considered. The drywell can be intact (base leakage assumed), it can have a small pre-existing failure (10 times base leakage using the EPRI ILRT assumption), or it can have a large pre-existing failure (100 times base leakage using the EPRI ILRT assumption).

As discussed in Section 4.6.1, this leads to nine combinations of drywell and containment leakage sizes (refer to Figure 4.1-1). Each combination has an impact on radionuclide releases that corresponds to one of the containment failure categories.

4-3 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval A A , La Normal DWLb B , 10La B

10DWLb RPV C 100DWLb C , 100La Drywell Containment Boundary Boundary FIGURE 4.1-1 CPS DRYWELL AND CONTAINMENT LEAKAGE CATEGORIES The Mark III and CPS plant-specific data used for the DWBT portion of this risk assessment is provided in Section 4.6.

4-4 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 4.2 PLANT SPECIFIC INPUTS The CPS specific information used to perform this ILRT interval extension risk assessment includes the following:

x FPIE PRA model Level 1 and LERF quantification results [24, 25]

x Population within a 50-mile radius (see Section 4.3) x Reactor Power Level [18]

x Allowable Containment Leakage [18]

CPS FPIE Core Damage Frequency The current CPS FPIE PRA analysis of record is an event tree/linked fault tree model characteristic of the as-built, as-operated plant. Based on the subsumed merged sequence cutset file results reported in the CPS PRA Summary Notebook for the 2017B PRA Update (CL117B) completed in 2020 [24], the mean value of the internal events core damage frequency (CDF) is 3.33E-06/yr (truncation limit 5E-13/yr). Core Damage Frequency by Class is provided in Table 4.2-1.

Release Category Definitions Table 4.2-2 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology [3]. 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 5.0 of this report.

4-5 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.2-1 2017 CLINTON LEVEL 1 CDF RESULTS(3)

CLASS DESCRIPTION CDF (/YR)  % OF CDF(1)

IA/IE Loss of Makeup at High RPV Pressure (Transient Initiators) 6.34E-07 19.05%

IBE Early Station Blackout (less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) 5.17E-07 15.52%

IBL Late Station Blackout (greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) 1.03E-06 30.95%

IC ATWS-Induced LOCA (Containment Intact) 7.76E-09 0.23%

ID Loss of Makeup at Low RPV Pressure (Transient Initiators) 1.37E-07 4.11%

IIA Loss of Containment Heat Removal (RPV initially Intact) 6.81E-07 20.46%

IIL Loss of Containment Heat Removal (RPV breached but no 1.50E-09 0.05%

initial core damage)

IIT Loss of Containment Heat Removal (RPV initially Intact; 1.32E-08 0.40%

core damage induced post high containment pressure)

IIV Class IIA and IIL except that the vent operates as designed 9.50E-08 2.85%

IIIA Excessive LOCA 9.66E-10 0.03%

IIIB Loss of Makeup at High RPV Pressure (LOCA Initiators) 2.33E-09 0.07%

IIIC Loss of Makeup at Low RPV Pressure (LOCA Initiators) 1.27E-08 0.38%

IIID Loss of Vapor Suppression (LOCA Initiators) 4.46E-09 0.13%

IV(2) Loss of Adequate Reactivity Control (ATWS) 1.53E-07 4.61%

V Containment Bypass 3.88E-08 1.17%

Total - 3.33E-06 100.0%

Notes to Table 4.2-1:

(1) Level 1 model results used as input to Level 2 update (based on 5E-13/yr truncation frequency).

(2)

Classes IVA and IVL included in Class IV.

(3) From Table 3.3-2 of The CPS PRA Summary Notebook [24]. The update is referred to as the 2017 update but reflects information incorporated in 2020.

4-6 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.2-2 EPRI [2] /NEI CONTAINMENT FAILURE CLASSIFICATIONS CLASS DESCRIPTION 1 Containment remains intact including accident sequences that do not lead to containment failure in the long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values L a, under Appendix J for that plant 2 Containment isolation failures (as reported in the IPEs) include those accidents in which there is a failure to isolate the containment.

3 Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal (i.e., provide a leak-tight containment) is not dependent on the sequence in progress.

4 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 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.

5 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 4 isolation failures, but is applicable to sequences involving Type C tests and their potential failures.

6 Containment isolation failures include those leak paths covered in the plant test and maintenance requirements or verified per in service inspection and testing (ISI/IST) program.

7 Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J testing requirements do not impact these accidents.

8 Accidents in which the containment is bypassed (either as an initial condition or induced by phenomena) are included in Class 8. Changes in Appendix J testing requirements do not impact these accidents.

CPS Internal Events Release Category Frequencies The CPS PRA Combined Level 1/Level 2 Model [25] is used to develop the initial set of internal events release categories for use in this analysis. Table 4.2-3 summarizes the pertinent CPS results in terms of release category. The OK release (i.e., intact containment limited to normal leakage) is 9.61E-7/yr. The total LERF, which corresponds to the High/Early (H/E) release category in Table 4.2-3, was calculated to be 1.65E-7/yr.

While the total release frequency shown in Table 4.2-3 is 2.43E-6, the total release frequency used in this ILRT extension analysis is 2.37E-6/yr, which is calculated by subtracting the OK release from total CDF (3.33E-6/yr - 9.61E-7/yr = 2.37E-6/yr). While 4-7 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval the total Level 2 frequency includes some double counting of release categories that leads to a slight overcounting, the OK release, by definition, is only one release category so its frequency is preserved.

Based on the Level 1 and LERF PRA model results described above, Table 4.2-4 lists the relevant EPRI release category frequencies pertinent for the ILRT extension risk assessment, including the delineation of LERF and non-LERF frequencies for Class 7.

The non-LERF frequencies for Class 7 are determined by reviewing 128 top L2 release sequences, which constitute more than 99% of all L2 release frequency. The sequences that did not end in a LERF endstate or did not include containment isolation failures (Class

2) were identified as Class 7 non-LERF contributors.

4-8 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.2-3

SUMMARY

OF CONTAINMENT EVALUATION(1)

INPUT OUTPUT LEVEL 1 PSA CET EVALUATION RELEASE CORE DAMAGE CHARACTERIZE FREQUENCY FREQUENCY RELEASE RELEASE BIN(3) (PER YEAR) 3.33E-6/year Little or No Release (Intact) OK 9.61E-07 LL & Late 5.09E-08 LL & I H(2)

Low Public LL & E 2.83E-09 Risk Impact L & Late 3.08E-07 L&I H(2)

L&E 4.10E-07 Moderate M & Late 5.51E-07 Release M&I H(2)

M&E 1.94E-07 H & Late 7.46E-07 High Release H&I H(2)

H&E 1.65E-07 Notes to Table 4.2-3:

(1) From Table 3.4-6B of The CPS PRA Summary Notebook [24].

(2) = negligibly small number (3) The Release Bin nomenclature is the following:

First Designator (Radionuclide release type)

1) High (H) - A radionuclide release of sufficient magnitude to have the potential to cause prompt fatalities.

2) Medium or Moderate (M) - A radionuclide release of sufficient magnitude to cause near-term health effects.

3) Low (L) - A radionuclide release with the potential for latent health effects.

4) Low-Low (LL) - A radionuclide release with undetectable or minor health effects.

5) Negligible (OK) - A radionuclide release that is less than or equal to the containment design base leakage.

Second Designator (Timing)

1) Early (E) Less than time when evacuation is effective (i.e., 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)

2) Intermediate (I) Greater than or equal to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> but less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

3) Late (L) Greater than or equal to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4-9 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.2-4 RELEVANT LEVEL 2 RELEASE CATEGORY FREQUENCIES FOR CPS EPRI RELEASE CATEGORY FREQUENCY/YR SOURCE 1: No Containment Failure 9.61E-07 Table 4.2-2 Level 2 contribution of sequences (top 99.5%)

that include failure of the 2: Containment Isolation Failure(1) 4.39E-07 containment isolation function (Event Tree Node IS=F)

Table 4.2-2 7: Phenomena-induced containment failures LERF (H/E) -

(LERF) 1.27E-07 Containment Bypass (Class V)

Table 4.2-2 and sequence evaluation 3.33E-06 CDF -

7: Phenomena-induced containment failures - 9.61E-07 (non-LERF) OK (EPRI Class 1) -

- 1.65E-07 LERF (EPRI

- 4.39E-07 Class 7 LERF & 8) -

= 1.76E-06 Cont. Isolation (EPRI Class 2)

Class V 8: Containment Bypass 3.88E-08 (ISLOCA + BOC Sequences)

Total: 3.33E-06 Note to Table 4.2-4:

(1) Not all Containment Isolation Failure sequences are found to be H/E (LERF) in the CPS PRA based on MAAP calculations. Nevertheless, EPRI Class 2 is conservatively assigned to Bin 1, which is the largest person-rem assignment (see Table 4.3-3). This category does not have a significant impact on results and only affects the total dose and % change metrics.

4-10 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 4.3 CPS POPULATION DOSE DERIVATION Since CPS does not maintain a detailed Level 3 PRA model, the approach recommended in EPRI 1018243 [3] is used. For the cases where plant-specific PRA dose information is not available, a representative population dose can be calculated using other references, such as NUREG/CR-4551 [17]. This approach was taken for the 2003 CPS ILRT / DWBT one-time extension [19] that was approved by the NRC [11] and the 2015/2016 CPS ILRT /

DWBT permanent extension [36] that was approved by the NRC [37]. To develop a representative population dose, the NUREG/CR-4551 plant that most closely resembles the analysis plant is chosen and the following steps are performed.

x Relate the NUREG/CR-4551 accident progression bins (APBs), EPRI Accident Classes, and plant-specific plant damage states (PDSs) based on the definitions contained in NUREG/CR-4551, and plant-specific PDSs.

x Adjust the resulting EPRI Accident Class 1, 2, 7, and 8 population doses to account for substantial differences in reactor power level, population density, allowable containment leak rate (La), and other plant-specific factors that may affect population dose as follows:

 Population density adjustment = (population within 50 miles of the CPS

÷ population within 50 miles of the NUREG/CR-4551 reference plant)

 Power level adjustment = (rated power level of CPS (MWt) ÷ rated power level of reference plant)

 La adjustment = La of CPS (%wt/day) ÷ La of reference plant Note that the population density and power level adjustments are applicable to all EPRI accident classes; however, the La adjustment should be made only to intact containment end states.

Reference Plant Population Dose Information Consistent with the EPRI guidance [3], the ex-plant consequence analysis for Grand Gulf is used as the reference plant for CPS since Grand Gulf is also a BWR Mark III containment. Table 4.3-1 reproduces the APB descriptions for Grand Gulf provided in Section 2.4.2 of NUREG/CR-4551 [17], and Table 4.3-2 provides a calculation to determine the relevant population dose associated with each APB. Note that Table 4.3-2 is consistent with the calculations previously performed for the Clinton ILRT interval extension submittals [19, 36].

4-11 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.3-1 COLLAPSED ACCIDENT PROGRESSION BIN DESCRIPTIONS FOR GRAND GULF [17]

COLLAPSED DESCRIPTION APB NUMBER 1 CD, vessel breach, Early CF, Early SP Bypass, CS Not Available Vessel breach occurs and both the containment and the drywell have failed either before or at the time of vessel breach. The containment sprays do not operate before or at the time of vessel breach.

2 CD, vessel breach, Early CF, Early SP Bypass, CS Available Vessel breach occurs and both the containment and the drywell fail either before or at the time of vessel breach. In this bin, however, the containment sprays operate before or at the time of vessel breach.

3 CD, vessel breach, Early CF, Late SP Bypass Vessel breach occurs and the containment fails either before or at the time of vessel breach. The drywell does not fail until the late time period and, thus, both the in-vessel releases and the releases associated with vessel breach are scrubbed by the suppression pool. Therefore, the availability of containment sprays during the time period that the suppression pool is not bypassed is not very important and, thus, the CS characteristic has been dropped.

4 CD, vessel breach, Early CF, No SP Bypass Vessel breach occurs and the containment fails either before or at the time of vessel breach. The drywell does not fail and, therefore, all of the radionuclide releases pass through the suppression pool. Because the pool has not been bypassed, the availability of the sprays is not very important and, thus, the CS characteristic has been dropped.

5 CD, vessel breach, Late CF Vessel breach occurs, however, the containment does not fail until the late time period. If the containment did not fail early, it is unlikely that the drywell will fail early. Thus, the suppression pool bypass characteristic and the containment spray characteristic have been dropped.

6 CD, vessel breach, Vent This summary bin represents the case in which vessel breach occurs and the containment was vented during any of the time periods in the accident.

7 CD, VB, No CF Vessel breach occurs but there is no containment failure and any releases associated with normal containment leakage are minor. Thus, the suppression pool bypass characteristic and the containment spray characteristic have been dropped. The risk associated with this bin will be negligible.

8 CD, No vessel breach Vessel breach is averted. Thus, there are no releases associated with vessel breach and there are no CCI releases. It must be remembered, however, that the containment can fail even if vessel breach is averted. Thus, the potential exists for some of the in-vessel releases to be released to the environment. It follows that there will be some risk associated with this bin.

Legend for Table 4.3-1:

CCI = Core Concrete Interaction CD = Core Damage CF = Containment Failure CS = Containment Sprays SP = Suppression Pool VB = Vessel Breach 4-12 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.3-2 GRAND GULF NUREG/CR-4551 [17] 50-MILE RADIUS POPULATION DOSE APB FRACTIONAL APB 50-MILE APB CONTRIBUTIO RADIUS DOSE APB 50-MILE FREQUENCY N TO 50-MILE RISK RADIUS DOSE APB (PER RADIUS TOTAL (PERSON- (PERSON-

1. (1) APB DEFINITION YEAR)(2) DOSE RISK(3) REM/YEAR)(4) REM)(5)

CD, VB, Early CF, 6.46E-7 .268 0.139 2.15E+5 1 Early SP Bypass, CS Not Available CD, VB, Early CF, 2.00E-7 .056 0.029 1.45E+5 2 Early SP Bypass, CS Available CD, VB, Early CF, 2.86E-8 .011 5.7E-3 1.99E+5 3

Late SP Bypass CD, VB, Early CF, 8.92E-7 .267 0.139 1.56E+5 4

No SP Bypass 5 CD, VB, Late CF 1.16E-6 .281 0.146 1.26E+5 6 CD, VB, Vent 1.55E-7 .039 0.0203 1.31E+5 7 CD, VB, No CF 2.05E-7 3E-4 1.56E-4 7.61E+2 8 CD, No VB 7.36E-7 .077 0.040 5.43E+4 Total 4.09E-6 1.0 0.52 --

Notes to Table 4.3-2:

(1)

This table is presented in the form of a calculation because NUREG/CR-4551 [17] does not document dose results as a function of accident progression bin (APB); as such, the dose results as a function of APB must be back calculated from documented APB frequencies and APB dose risk results in NUREG/CR-4551.

(2)

The total (i.e., internal accident sequences) CDF of 4.09E-6/yr and the CDF subtotals by APB are taken from Figure 2.5-7 of NUREG/CR-4551 Vol. 6 Rev.1 Part 1.

(3)

The individual APB contributions to total (i.e., internal accident sequences) 50-mile radius dose rate are taken from Table 5.1-3 of NUREG/CR-4551 Vol. 6 Rev.1 Part 1.

(4)

The APB 50-mile dose risk is calculated by multiplying the individual APB dose risk contributions (column 4) by the total mean 50-mile radius dose risk of 0.52 person-rem/yr (taken from Table 5.1-1 of NUREG/CR-4551 Vol. 6 Rev.1 Part 1).

(5)

The individual APB doses are calculated by dividing the individual APB dose risk by the APB frequencies.

Legend for Table 4.3-2:

CCI = Core Concrete Interaction CD = Core Damage CF = Containment Failure CS = Containment Sprays SP = Suppression Pool VB = Vessel Breach 4-13 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval The APBs described above are assigned to one of the EPRI release categories for the CPS assessment. These assignments and their basis are provided in Table 4.3-3. It is noted that no assignment of NUREG/CR-4551 APBs is made for EPRI Release Categories 3a and 3b because, per the EPRI methodology, these doses are calculated using factors of 10 and 100, respectively, of the population dose for EPRI Category 1.

Also, EPRI Categories 4, 5, and 6 are not affected by the ILRT frequency and are therefore (per the EPRI guidance) not included in the assignment process.

TABLE 4.3-3 ASSIGNED APB FOR EACH OF THE RELEVANT LEVEL 2 RELEASE CATEGORIES FOR CPS EPRI RELEASE CATEGORY ASSIGNED BASIS APB 1: No Containment Failure 7 The intact containment case with release limited to leakage is represented by APB 7 in the Grand Gulf assessment.

2: Containment Isolation 1 APB 1 is conservatively chosen since the drywell may Failure fail or the Suppression Pool bypassed, leading to an early release scenario. APB 1 results in the highest dose of all the Grand Gulf containment failure APBs (which is indicative of a containment failure with suppression pool or drywell bypass) 7: Phenomena-induced 1 APB 1 w/o containment sprays available is chosen for containment failures (LERF) CPS. APB 1 results in the highest dose of all the Grand Gulf containment failure APBs (which is indicative of LERF) 7: Phenomena-induced 1, 3, 4, 5, 6, For CPS, this release category has both early and late containment failures (non- and 8 releases. The sequences are binned accordingly, as LERF) presented in more detail in Table 4.3-7.

8: Containment Bypass 1 The containment bypass case is selected as APB 1 from the Grand Gulf assessment. APB 1 results in the highest dose of all the Grand Gulf containment failure APBs (which is indicative of containment bypass) 4-14 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval Adjustments to Ex-Plant Consequence Calculations The next step per the EPRI guidance is to adjust the resulting EPRI Accident Class 1, 2, 7, and 8 population doses from the reference plant to account for differences in reactor power level, population, and allowable containment leak rate (La).

The 50-mile radius population used in the Grand Gulf NUREG/CR-4551 consequence calculations is 3.25E+5 persons. This is based on 1980 Census data as documented in NUREG/CR-4551 Vol. 2, Rev. 1, Part 7 [33] Appendix A.3.

TABLE 4.3-4 NUREG/CR-4551 GRAND GULF POPULATION DISTANCE FROM PLANT (KM) (MILES) POPULATION 1.6 1.0 34 4.8 3.0 879 16.1 10.0 10,255 32.2 20.0 28,151 48.3 30.0 97,395 64.37 40.0 192,677 80.47 50.0 325,285 The 50-mile radius population dose for CPS is based on the 2030 population estimate projection using SECPOP 4.2 population data [31] for 2000 and 2010 and assuming the population growth rate from 2000 to 2010 continues for the next two decades. The SECPOP 4.2 code uses census data to calculate population counts for user defined sector segments. The EPRI methodology does not specify population projection to a future date. For this risk assessment, the CPS population was conservatively projected to the year 2030 to represent potential future average population density increases (although current trends indicate a population decrease).

4-15 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.3-5 SECPOP CODE POPULATION ESTIMATES SEC POP RADIUS 2000 2010 2020< 1>< 2 > 2030< 1>< 2 >

0-10 12,334 12,219 12,105 11,992 0-20 57,626 61, 143 64,875 68,834 0-30 351,649 374,458 398,746 424,610 0-40 538,318 574,660 613,455 654,870 0-50 768,541 813,071 860, 181 910,021 (1l The 2020 and 2030 estimates are made assuming the 5.79%

population increase experienced in the SO-mile radius region during the decade from July 2000 to July 2010 continues to occur each of the next two decades.

2

( l The Illinois Department of Public Health (IDPH) growth projections

[30] from 2010 to 2025 for the counties with areas within the 50 mile radius of CPS are shown in Table 4.3-58 below. The combined county growth rate for these counties is 3.7% for the 2010 to 2025 year period. The IDPH the population projection of 3.7% for a 15 year period demonstrates that the SECPOP 4.2 based growth rate of 5.79% per decade is conservative (leading to a higher CPS dose projection). Additionally, 2019 population data [1 O] was examined to determine the population change since 2010. Population decreased more than 2% from 2010 to 2019, as shown in Table 4.3-58.

Therefore, the dose calculation methodology is corroborated to be conservative. See map and table below.

ICHOX LIVINGSTON INDIANA Illinois Counties Within 50 Miles of CPS Shown in Map Above 4-16 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.3-5B POPULATION PROJECTION ESTIMATES AREA 2020 2025 2019 WITHIN 2010 POPULATION POPULATION POPULATION -

50 MILE POPULATION ESTIMATE - ESTIMATE - US CENSUS COUNTY RADIUS - IDPH [30] IDPH [30] IDPH [30] BUREAU [10]

Champaign 99% 201,370 217,735 225,626 209,689 Christian 40% 34,804 33,152 32,345 32,304 Coles 20% 53,945 56,851 58,405 50,621 De Witt 100% 16,583 15,832 15,495 15,638 Douglas 75% 19,976 19,767 19,709 19,465 Ford 75% 14,074 13,450 13,244 12,961 Iroquois <5% 29,657 27,687 26,816 27,114 Livingston 40% 38,882 39,390 39,596 35,648 Logan 100% 30,272 30,380 30,441 28,618 Macon 100% 110,757 105,401 103,126 104,009 Mason 25% 14,627 12,841 12,074 13,359 McLean 100% 169,838 188,341 197,855 171,517 Menard 40% 12,708 12,867 12,913 12,196 Moultrie 90% 14,846 14,715 14,706 14,501 Piatt 100% 16,722 16,205 16,000 16,344 Sangamon 35% 197,822 203,501 207,194 194,672 Shelby 25% 22,339 21,496 21,118 21,634 Tazewell 75% 135,439 136,051 136,436 131,803 Vermillion <5% 81,588 77,965 76,441 75,758 Woodford 50% 38,664 40,350 41,360 38,459 Total -- 1,255,013 1,283,977 1,300,900(1)(2) 1,226,310(3)

Notes to Table 4.3-5B:

(1) This total is used to determine average county growth over 15-year period. Since many counties have population outside the 50-mile radius, the population within 50 miles of CPS is significantly less than the totals shown above.

(2) IDPH Projected Population Increase from 2010-2025 =

(1,300,900 ÷ 1,255,013 - 1) ÷ 100% = 3.66%

(3) Actual population change from 2010-2019 = (1,226,310 ÷ 1,255,013 - 1) ÷ 100% = -2.29%

4-17 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval The ratio of the population surrounding CPS (Table 4.3-5), to that in the Grand Gulf analysis results in a factor increase of:

9.10E+5 persons / 3.25E+5 persons = 2.80 The Grand Gulf reactor power level used in the NUREG/CR-4551 consequence calculations is 3833 MWt [33]. The current CPS reactor power level is 3473 MWt [18].

Therefore, the ratio of the CPS reactor power to that used in the Grand Gulf analysis results in a multiplication factor of:

3473 MWt / 3833 MWt = 0.91 The containment leakage used in the NUREG/CR-4551 consequence calculations for Grand Gulf is 0.5 %wt/day [29]. The current CPS allowable leakage is 0.65 %wt/day [18].

Because the leakage rates are a function of the containment volume, these plant characteristics are also needed:

x Grand Gulf Containment Volume [29] = 1.40E+6 ft3 x CPS Containment Volume [18] = 1.51E+6 ft3 Therefore, the ratio of the CPS allowable leakage and containment volume to that used in the Grand Gulf analysis results in a multiplication factor of:

(0.65%

• 1.51E+6) / (0.5%

• 1.40E+6) = 1.40 As stated previously, this final adjustment factor is only applied to the intact containment case. Table 4.3-6 provides a summary of each of the adjustment factors used for each APB to estimate the population doses for CPS that can be used in this assessment.

4-18 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.3-6 CPS ADJUSTED 50-MILE RADIUS POPULATION DOSE GRAND GULF CPS POPULATION 50-MILE REACTOR CONTAINMENT DOSE ADJUSTED RADIUS DOSE POPULATION POWER LEAK RATE 50-MILE (PERSON- ADJUSTMENT ADJUSTMENT ADJUSTMENT RADIUS DOSE APB # REM)(1) FACTOR FACTOR FACTOR (PERSON-REM) 1 2.15E+05 2.8 0.91 N/A 5.48E+05 2 1.45E+05 2.8 0.91 N/A 3.69E+05 3 1.99E+05 2.8 0.91 N/A 5.07E+05 4 1.56E+05 2.8 0.91 N/A 3.97E+05 5 1.26E+05 2.8 0.91 N/A 3.21E+05 6 1.31E+05 2.8 0.91 N/A 3.34E+05 7 7.61E+02 2.8 0.91 1.40 2.71E+03 8 5.43E+04 2.8 0.91 N/A 1.38E+05 Note to Table 4.3-6:

(1) The NUREG/CR-4551 evaluation of Grand Gulf is used as input to the assessment of population dose for CPS.

Refer to Table 4.3-2.

Population Dose Risk Calculations The next step is to take the frequency information from Table 4.2-4 for each relevant EPRI release category class in Table 4.1-1 and associate a representative population dose from Table 4.3-6 for each release category based on the APB assignments made in Table 4.3-3. As discussed in more detail below, EPRI class 7 is further refined based on the CPS Level 2 PRA, as identified in Table 4.3-7, and allocated frequency and APB doses as identified in Table 4.3-8. Table 4.3-9 lists the population dose risk organized by EPRI release category for CPS, including the delineation of LERF and non-LERF frequencies for Class 7. Note that the population dose risk (Column 4 of Table 4.3-8 and Table 4.3-9) was found by multiplying the release category frequency (Column 2 of Table 4.3-8 and Table 4.3-9) by the associated population dose (Column 3 of Table 4.3-8 and Table 4.3-9). Also note that only the applicable EPRI release categories at this point are shown in the tables (i.e., the Class 3 frequencies are derived later and the Class 4, 5, and 6 frequencies are not used in the EPRI methodology for the ILRT extension risk assessment).

4-19 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval Application of Clinton PRA Model Results to NUREG/CR-4551 Dose Results A major factor related to the use of NUREG/CR-4551 in this evaluation is that the results of the current Clinton PRA Level 2 model are categorized by accident classes that differ from the NUREG/CR-4551 APB classification scheme. Therefore, an assignment process is required to apply the NUREG/CR-4551 dose results. This subsection provides a description of the process used.

The basic process used was to review 128 top sequences of the Clinton Level 2 model (which provide more than 99.4% of Level 2 release frequency) and to assign each sequence into one of the collapsed APBs from NUREG/CR-4551. The CPS Level 2 model (i.e., containment event tree structure) contains significantly more information about the accident sequences than in the collapsed APBs in NUREG/CR-4551, and this assignment process required simplification of CPS accident progression information and assumptions related to categorizations of certain items. The relevant assumptions used for these assignments are summarized in Table 4.3-7. Other containment event tree nodes are included in the Clinton Level 2 model, but these were not used in (or did not contribute to) to the APB assignment performed here for the ILRT assessment. Additionally, these bin assignments are all related to EPRI Class 7 and therefore influence the total base case population dose estimated for CPS but do not influence the change in dose calculated for the ILRT extension risk assessment.

Class 7 Sequences Dose Risk Adjustments EPRI Class 7 consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs. For this analysis, the associated radionuclide releases are based on the application of the Level 2 containment end states to the APBs from NUREG/CR-4551 as described in Section 4.2. The Class 7 Sequences are divided into two categories. Class 7 LERF is defined to consist of LERF sequences (excluding the Break Outside Containment (BOC) and Interfacing Systems LOCA (ISLOCA) Class V sequences) and Class 7 non-LERF is defined to consist of non-LERF sequences. The second category (non-LERF) is further divided into Class 7a, 7b, 7c, 7d, 7e, and 7f for assignments of NUREG/CR-4551 APB Bins 1, 3, 4, 5, 6, and 8. The failure 4-20 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval frequency and population dose for each specific APB is shown in Table 4.3-8. The total release frequency and total dose for the Non-LERF Accident Class 7a, 7b, 7c, 7d, 7e, and 7f are then used to determine a weighted average person-rem for use as the representative EPRI Class 7 non-LERF dose in the subsequent analysis. Note that the total frequency and dose associated from this EPRI class does not change based on the ILRT interval, since Class 7 involves containment failure.

4-21 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.3-7 CPS LEVEL 2 NODAL ASSUMPTIONS FOR APB ASSIGNMENTS CPS PRA CONTAINMENT ACCIDENT EVENT TREE CLASS NODE ASSUMPTION 1 and 3 RX - Core Melt A success at this node signifies that there is no vessel breach. The Arrested in Vessel sequences following this path are generally grouped in APB 8.

However, there are cases in which Engertic Containment Failure (CX) occur. In those cases, these scenairios are assumed to result in a high early release and are categorized as APB 1. Additionally Supp. Pool Failure below the water line (WW) is assumed to result in a high late release and is assigned APB 3.

Failure at this node means the core leaves the vessel. APB assignments are based on subsequent nodes.

CZ - No Energetic If there is energetic DW failure (CZ) and energetic containment DW Failure failure (CX), these are assumed to be high early releases and APB CX - No Energetic 1 (highest dose) is assigned.

Containment Failure DL - Drywell Isolation If the drywell is not isolated or Suppression Pool Scrubbing fails, it SP - Suppression is assumed that an un-scrubbed release to containment occurs as Pool scrubbing soon as the vessel is breached. If Containment Spray (CS) fails the sequence is categorized as APB 1 (the highest dose). If Containment Spray is successful, APB 2 is assigned.

If the drywell is isolated and Suppression Pool scrubbing (SP) is successful; Energetic Containment (CX) failure or Containment Isolation (IS) failure occur, early containment failure is assumed and APB 3 is assigned. If Energetic Containment (CX) failure or Containment Isolation (IS) failure do not occur, late containment failure is assumed and categorized as APB 5.

An exception to the above rules applies to IBL sequences. Class IBL is defined as Late Station Blackout events with core damage at greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Early injection is present and the drywell is likely not failed early allowing for Suppression Pool scrubbing.

Therefore, IBL sequences are categorized as APB 5.

SI - Late RPV RPV Injection before containment failure lowers the radioactive injection before release. If not preceded by an energetic containment (CX) failure Containment Failure the sequence is categorized as APB 4.

WW - Suppression If the Supp. Pool fails below the water line (WW); this is assumed Pool Failure Above to result in a high early release and is categorized as APB 1.

the Water Line VC - Containment Sequences with successful containment vent are typicaly assigned Vent to APB 6. However, if vessel breach does not occur because injection is available after core damage and remains available after containment venting, then APB 8 is assigned.

4-22 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.3-7 CPS LEVEL 2 NODAL ASSUMPTIONS FOR APB ASSIGNMENTS CPS PRA CONTAINMENT ACCIDENT EVENT TREE CLASS NODE ASSUMPTION 2 RX - Core Melt For accident class 2, RX is always assumed failed.

Arrested in Vessel WW - Suppression If Supp. Pool fails below the water line (WW fails) it is assumed to Pool Failure Above result in a high late release and is categorized as APB 5.

the Water Line Accident sequences IIE (late GE declaration) with WW failure are assumed early and are categorized as APB1, unless venting is successful (Class IIV) in which case APB 6 is assigned.

CZ - No Energetic For all Class II sequences with early GE declaration, containment DW Failure failure will occur in the late time frame. Therefore, these are DL - Drywell Isolation assigned to APB 5.

SP - Suppression For Class IIE sequences with late GE declaration, containment Pool scrubbing failure is assumed to occur in the early time frame. If the drywell is not isolated (CZ or DL fail) or Suppression Pool Scrubbing fails (SP fails), it is assumed that an un-scrubbed release to containment occurs as soon as the vessel is breached. These are assigned to ABP 1. If the drywell is isolated (CZ and DL success) and Suppression Pool scrubbing (SP) is successful; no Energetic Containment (CX) failure or Containment Isolation (IS) failure occur, then early containment failure is assumed and APB 4 is assigned.

VC - Containment Sequences with successful containment vent are typicaly assigned Vent to APB 6. However, if vessel breach does not occur because injection is available after core damage and remains available after containment venting, then APB 8 is assigned.

4 RX - Core Melt For accident class 4, RX is always assumed failed.

Arrested in Vessel CZ - No Energetic If there are no energetic failures of the Drywell, Suppression Pool DW Failure Scrubbing or Wetwell failure, the sequence is assigned APB 4. If SP - Suppression any do fail, APB 1 is assigned.

Pool Scrubbing WW - Suppression Pool Failure Above the Water Line 5 N/A No collapsed bin is available for containment bypass scenarios.

The closest match to a bypass scenario is assumed to be a vessel breach with early drywell and containment failure APB 1. Bin 1 is assigned as it represents the largest release.

4-23 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.3-8 ACCIDENT CLASS 7 FAILURE FREQUENCIES AND POPULATION DOSES (CPS LEVEL 2 MODEL)

POPULATION DOSE POPULATION RISK (50 MILES)

ACCIDENT CLASS RELEASE DOSE (50 MILES) (PERSON-REM / YR)

(APB NUMBER) FREQUENCY / YR PERSON-REM (1) (2) 7 LERF (APB 1) 1.27E-07 5.48E+05 6.94E-02 7 non-LERF 7a (APB 1) 2.89E-08 5.48E+05 1.59E-02 7b (APB 3) 6.54E-09 5.08E+05 3.32E-03 7c (APB 4) 5.72E-08 3.97E+05 2.28E-02 7d (APB 5) 1.36E-06 3.21E+05 4.36E-01 7e (APB 6) 2.90E-07 3.34E+05 9.68E-02 7f (APB 8) 1.62E-08 1.38E+05 2.25E-03 Class 7 non-LERF Total 1.71E-06 3.28E+05 (3) 5.77E-01 Notes to Table 4.3-8:

(1)

Population dose values obtained from Table 4.3-6 based on the Accident Progression Bin.

(2) Obtained by multiplying the Release Frequency value from the second column of this table by the Population dose value from the third column of this table.

(3) The weighted average population dose for Class 7 non-LERF is obtained by dividing the total population dose risk by the total release frequency of categories 7a, 7b, 7c 7d, 7e, and 7f.

TABLE 4.3-9 CPS POPULATION DOSE AND DOSE RISK ORGANIZED BY EPRI RELEASE CATEGORY POPULATION POPULATION DOSE DOSE RISK EPRI RELEASE CATEGORY FREQUENCY/YR (PERSON-REM) (PERSON-REM/YR) 1: No Containment Failure 9.61E-07 2.71E+03 2.61E-03 2: Containment Isolation Failure 4.39E-07 5.48E+05 2.41E-01 7: Phenomena-induced 1.27E-07 6.94E-02 5.48E+05 containment failures (LERF) 7: Phenomena-induced 1.76E-06 5.79E-01 3.28E+05 containment failures (non-LERF) 8: Containment Bypass 3.88E-08 5.48E+05 2.13E-02 Total: 3.33E-06 9.13E-01 4-24 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 4.4 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 breach and failure of some sealing surfaces, which can lead to leakage. The proposed ILRT interval extension may influence the conditional probability of detecting these types of failures. To ensure that this effect is properly accounted for, the EPRI Class 3 accident class as defined in Table 4.1-1 is divided into two sub-classes representing small and large leakage failures. These subclasses are defined as Class 3a and Class 3b, respectively.

The probability of the EPRI Class 3a failures may be determined, consistent with the EPRI guidance [3], as the mean failure estimated from the available data (i.e., 2 small failures that could only have been discovered by the ILRT in 217 tests leads to a 2/217=0.0092 mean value). For Class 3b, consistent with EPRI guidance [3], a non-informative prior distribution is assumed for no large failures in 217 tests (i.e., 0.5/(217+1) = 0.0023).

The EPRI methodology contains information concerning the potential that the calculated LERF values for several plants may fall above the very small change guidelines of the NRC Regulatory Guide 1.174. This information includes a discussion of conservatisms in the quantitative guidance for LERF. EPRI describes ways to demonstrate that, using plant-specific calculations, the LERF is smaller than that calculated by the simplified method. The following is from the EPRI guidance:

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 CPS (as detailed in Section

5) means that the Class 7 phenomena-induced containment failures LERF sequences, and Class 8 containment bypass sequences are subtracted from the CDF that is applied 4-25 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval to Class 3b, as these sequences always result in LERF. Also, Class 7 phenomena-induced containment failures non-LERF sequences that are Class IBL and Class II are treated as never resulting in LERF due to their late timing and are subtracted from the CDF that is applied to Class 3b. To be consistent, the same change is made to the Class 3a CDF, even though these events are not considered LERF. Class IBL (Late Station Blackout) sequences are events with core damage at greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />(1). Class II (Loss of Decay Heat Removal) are events with core damage caused by late containment failure (beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)(2). Class IIE sequences where the General Emergency (GE) is postulated to be declared late and there is a potential for LERF are conservatively retained.

Consistent with the EPRI methodology [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 yr / 2). This interval change would lead to a non-detection probability that is a factor of 5.23 (7.85/1.5) higher for the probability of a leak that is detectable only by ILRT testing, given a 15.7-year vs. a 3-yr interval.

(1) A MAAP 4.0.5 analysis CL06008

Title:

Accident Class IBL - SBO with RCIC calculates RPV breach in 8.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> [40]. The PRA assigns a high probability (i.e., 95%) of a General Emergency (GE) declaration in accordance with Emergency Action Level (EAL) MG1 at 1 hr based on assessment that restoration of power to both divisions vital buses will not occur within 4 hrs [32]. An early GE declaration scenario is a Class IBL and always results in a late release. The PRA assigns the scenario to Class IBE (i.e.,

5%) if the GE is declared late resulting in an early release. Class IBE scenarios are included as a contributor to EPRI accident class 3b. With 7+ hours between GE declaration and vessel breach, the IBL PRA scenario is always a late scenario and Class IBL is excluded from contributing to a 3b scenario.

(2) MAAP 4.0.5 Analysis CL110510

Title:

LOOP With Loss of Containment Heat Removal credits LPCS without containment heat removal [40]. Containment fails at 34.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> in this scenario. Class II Events only include events where a General Emergency (GE) is declared Early. A Late declaration is classified as Class IIE and is included as an EPRI Class 3b scenario. The PRA assigns a high probability (i.e., 95%) a GE would be declared Early at approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> before containment failure [32]. Assuming injection is lost at the time of containment failure, core damage would occur ~4.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after containment failure and ~8 hours after the GE declaration. Therefore, Class II events are always Late events and do not contribute to LERF. A late GE declaration (i.e., Class IIE) is assigned a low probability (i.e., 5%) in the PRA.

4-26 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval EPRI Methodology This analysis uses the approach outlined in the EPRI Methodology [3]. The steps of the methodology are as follows:

1. Quantify the baseline risk in terms of the frequency of events (per reactor year) for each of the eight containment release scenario types identified in the EPRI report [3].

2. Develop plant-specific population dose rates (person-rem per reactor year) for each of the eight containment release scenario types from plant specific consequence analyses.

3. Evaluate the risk impact (i.e., the change in containment release scenario type frequency and population dose) of extending the ILRT interval from fifteen years to 15.7 years.

4. Determine the change in risk in terms of Large Early Release Frequency (LERF) in accordance with RG 1.174 and compare this change with the acceptance guidelines of RG 1.174 [4].

5. Determine the impact on the CCFP.

6. Evaluate the impact of external events.

The first three steps of the methodology deal with calculating the change in dose. The change in dose is the historical principal basis upon which the Type A ILRT interval extension was previously granted and is a reasonable basis for evaluating additional extensions. The fourth step in the methodology calculates the change in LERF and compares it to the guidelines in Regulatory Guide 1.174 [4]. Because there is no change in CDF for CPS, the change in LERF forms the quantitative basis for a risk informed decision per current NRC practice, namely Regulatory Guide 1.174. The fifth step of the methodology calculates the change in containment failure probability, referred to as the CCFP. The NRC has identified a CCFP of less than 1.5% as the acceptance criteria for extending the Type A ILRT test intervals as the basis for showing that the proposed change is consistent with the defense in depth philosophy [7]. As such, this step suffices as the remaining basis for a risk informed decision per Regulatory Guide 1.174. Step 6 takes into consideration the additional risk due to external events and investigates the impact on results due to varying the assumptions associated with the liner corrosion rate and failure to visually identify pre-existing flaws.

4-27 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 4.5 IMPACT OF EXTENSION ON DETECTION OF STEEL CORROSION THAT LEADS TO LEAKAGE An estimate of the likelihood and risk implications of corrosion-induced leakage of the steel liners during the extended test interval is evaluated using the methodology from the Calvert Cliffs liner corrosion analysis [5], consistent with the EPRI methodology [3]. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner. The Clinton primary containment is a pressure-suppression BWR/Mark III containment type that also includes a steel-lined reinforced concrete structure.

The liner sections at Clinton are completely welded together and anchored into the concrete. There is no air space between the liner and the concrete structure. The corrosion/oxidation effects associated with water being in contact with the carbon steel liner and the concrete reinforcing bars are minimized due to the lack of available oxygen between the concrete and the liner. Furthermore, the liner is intended to be a membrane and constitutes a leak-proof boundary for the containment. The liner is nominally 0.25-inch to 0.50-inch thick depending on location and has been oversized to serve as form-work for concrete pouring during construction.

Because concrete was poured against the containment liner significant leakage from containment is not expected even if through-liner corrosion should occur.

The concrete side of the liner is not accessible and cannot be directly inspected by visual means. The large majority of the inside of the containment liner is fully exposed to the containment atmosphere and is accessible for inspection. There is a high likelihood that through-wall defects would be detected through the visual examinations performed.

Portions of the inside of the containment liner that are not accessible include the liner below the suppression pool surface and portions of the liner that are obstructed from view by equipment (e.g., piping, cable trays, ductwork) and structural elements (e.g.,

intermediate concrete floors) next to the containment wall. However, it is estimated that 80% of the inside of the containment liner that is exposed to air is accessible for inspection. The portion of the liner below the suppression pool surface is demonstrated 4-28 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval to be low leakage since it is capable of retaining suppression pool water. There are leak test channels at the containment liner seams in the suppression pool area to drain any water that leaks through the suppression pool liner. Therefore, leakage through the suppression pool liner is detectable.

The areas of the containment liner above the suppression pool surface that cannot be inspected are judged to be no more susceptible to degradation than those portions that are accessible. Corrosion identified in areas that are accessible for inspection could indicate that an investigation of similar areas that are not readily accessible may be required. Therefore, any widespread corrosion phenomena would be investigated, and corrective action would be taken. This does not completely preclude the possibility of undetected localized corrosion occurring in areas that are not accessible; however, industry experience has shown a fairly low incidence rate for through-liner corrosion.

Furthermore, localized breaches of the containment liner are not likely to lead to significant containment breaches since a leakage path through the reinforced concrete structure would also have to be present. A corrosion sensitivity study has been performed that estimates the impact on the ILRT risk assessment results based upon the above factors.

The following approach is used to determine the change in likelihood, due to extending the ILRT, of detecting corrosion of the containment steel structure. This likelihood is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:

x Differences between the containment basemat and the containment cylinder and dome x The historical flaw likelihood due to concealed corrosion x The impact of aging x The corrosion leakage dependency on containment pressure x The likelihood that visual inspections will be effective at detecting a flaw 4-29 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval Assumptions x Consistent with the Calvert Cliffs analysis, a half failure is assumed for the basemat concealed liner corrosion due to lack of identified failures (see Table 4.5-1, Step 1).

x The two corrosion events over a 5.5-year data period are used to estimate the flaw probability in the Calvert Cliffs analysis and are assumed to be applicable to the CPS containment analysis. These events, one at North Anna Unit 2 and one at Brunswick Unit 2, were initiated from the non-visible (backside) portion of the containment liner. It is noted that two additional events have occurred in recent years (based on a data search covering approximately 9 years documented in Reference [26]). In November 2006, the Turkey Point 4 containment building liner developed a hole when a sump pump support plate was moved. In May 2009, a hole approximately 3/8 by 1 in size was identified in the Beaver Valley 1 containment liner.

For risk evaluation purposes, these two more recent events occurring over a 9-year period are judged to be adequately represented by the two events in the 5.5-year period of the Calvert Cliffs analysis incorporated in the EPRI guidance (See Table 4.5-1, Step 1).

x Consistent with the Calvert Cliffs analysis, the steel 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 ages (See Table 4.5-1, Steps 2 and 3).

x In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere given that a flaw exists in the steel was estimated as 1.1% for the cylinder and dome region, 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, and the selected values are consistent with a pressure that corresponds to the ILRT target pressure of 37 psig. Consistent with the Calvert Cliffs analysis, probabilities of 1% for the cylinder and dome and 0.1% for the basemat are used in this analysis (See Table 4.5-1, Step 4).

x 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 walls (See Table 4.5-1, Step 4).

x In the Calvert Cliffs analysis, it is noted that approximately 85% of the interior wall surface is accessible for visual inspections. The amount at Clinton is approximately 80%, which is similar. Consistent with the Calvert analysis, a 5% visual inspection detection failure likelihood given the flaw is visible and a 5% likelihood of a non-detectable flaw are used. This results in a total undetected flaw probability of 10%, which is assumed in the base case analysis. (See Table 4.5-1, Step 5.) Additionally, it should be noted that to date, all liner corrosion events have been detected through visual inspection.

4-30 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval x 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. This is a particularly conservative assumption for Clinton because it is unlikely that any releases would not be scrubbed in the Mark III containment pool.

x Unlike the Calvert Cliffs design, the Clinton drywell has a steel liner.

However, due to the conservative treatment of the containment failures (see previous bullet), the impact of non-detection of corrosion will only be applied to the ILRT extension. The NEI/EPRI characterization of Category 3b as a LERF contributor is considered extremely conservative for a Mark III.

Inclusion of drywell liner non-detection failures due to steel corrosion would only increase the conservatism.

4-31 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.5-1 STEEL CORROSION BASE CASE CONTAINMENT CYLINDER STEP DESCRIPTION AND DOME CONTAINMENT BASEMAT 1 Historical Steel Flaw Likelihood Events: 2 Events: 0 (assume half a Failure Data: Containment failure) location specific (consistent with 2/(70

• 5.5) = 5.2E-3 0.5/(70

• 5.5) = 1.3E-3 Calvert Cliffs analysis).

2 Age Adjusted Steel Flaw Year Failure Rate Year Failure Rate Likelihood 1 2.1E-3 1 5.0E-4 During 15-year interval, assume 15.7 1.6E-2 15.7 3.9E-3 failure rate doubles every five years (14.9% increase per year).

The average for 5th to 10th year is 15.7 year average = 6.71E-3 15.7 year average = 1.68E-3 set to the historical failure rate (consistent with Calvert Cliffs analysis).

3 Flaw Likelihood at 3, 10, and 15 0.71% (1 to 3 years) 0.18% (1 to 3 years) years 9.65% (1 to 15 years) 2.42% (1 to 15 years)

Uses age adjusted flaw likelihood 11.23% (1 to 15.7 years) 2.81% (1 to 15 years)

(Step 2), assuming failure rate doubles every five years (consistent with Calvert Cliffs analysis - See Table 6 of Reference [5]).

4 Likelihood of Breach in 1% 0.1%

Containment Given Steel Flaw 5 Visual Inspection Detection 10% 100%

Failure Likelihood 5% failure to identify visual Cannot be visually Use assumptions consistent with flaws plus 5% likelihood that inspected.

Calvert Cliffs analysis. the flaw is not visible (not through-cylinder but could be detected by ILRT)

All events have been detected through visual inspection. 5% visible failure detection is a conservative assumption.

4-32 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.5-1 STEEL CORROSION BASE CASE CONTAINMENT CYLINDER STEP DESCRIPTION AND DOME CONTAINMENT BASEMAT 6 Likelihood of Non-Detected 0.00071% (at 3 years) 0.00018% (at 3 years)

Containment Leakage =0.71%

• 1%

• 10% =0.18%

• 0.1%

• 100%

(Steps 3

• 4

• 5) 0.00965% (at 15 years) 0.00242% (at 15 years)

=9.65%

• 1%

• 10% =2.42%

• 0.1%

• 100%

0.0112% (at 15.7 years) 0.00281% (at 15.7 years)

=11.2%

• 1%

• 10% =2.81%

• 0.1%

• 100%

The total likelihood of the corrosion-induced, non-detected containment leakage that is subsequently added to the EPRI Class 3b contribution is the sum of Step 6 for the containment cylinder and dome, and the containment basemat:

At 3 years: 0.00071% + 0.00018% = 0.00089%

At 15 years: 0.00965% + 0.00242% = 0.0121%

At 15.7 years: 0.00112% + 0.00281% = 0.0140%

4-33 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 4.6 IMPACT OF DWBT INTERVAL EXTENSION OF RELEASE CATEGORIES Clinton used a plant-specific MAAP 4.0 model to determine the effects of the increased drywell leakages that could lead to higher containment pressure. The MAAP results, along with other considerations, found that containment failure induced by containment pressurization aggravated by the drywell bypass leakage change is highly unlikely. The relatively small changes postulated due to the DWBT interval extension make no appreciable change in the containment pressurization compared to its ultimate capability.

The containment overpressure challenges due to the loss of containment heat removal capability are already accounted for in the Clinton PRA. As such, the perturbation on these sequences caused by slight changes in the drywell bypass area is considered a negligible contributor to CDF.

CPS assigned the equivalent EPRI category and LERF characterization as shown in Table 4.1-4. No credit for the availability of containment spray was taken in the CPS analyses. All EPRI class 1 and 3a were categorized as Non-LERF while 3b was categorized as LERF. The CPS MAAP runs did demonstrate that Class 3b, although treated as LERF, would result in releases significantly below the LERF threshold for CSI release. A similar approach to the previous CPS method to account for the DWBT will be used for this assessment.

4-34 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.6-1 CPS [19] DWBT AND ILRT LEAKAGE COMBINATION ACCIDENT CLASSES EPRI LEAKAGE DW BYPASS CONTAINMENT CLASSIFICATION COMBINATIONS LEAKAGE LEAKAGE ASSIGNMENT AA 1DWLb 1La 1 (Non-LERF)

AB 1DWLb 10La 3a (Non-LERF)

AC 1DWLb 35La(1) 3b (LERF)

BA1 10DWLb 1La 1 (Non-LERF)

BB1 10DWLb 10La 3a (Non-LERF)

BC1 10DWLb 35La 3b (LERF)

CA1 35DWLb(1) 1La 3a (Non-LERF)

CB1 35DWLb 10La 3b (LERF)

CC1 35DWLb 35La 3b (LERF)

Note to Table 4.6-1:

(1) Note that 35La was used in the prior assessments, but per the updated EPRI guidance as approved by the NRC, 100La is now used for EPRI Class 3b.

Consistent with the prior assessments, the probability for each combination in Table 4.6-1 is determined by multiplying the conditional probabilities for DWBT and ILRT category by each other. Section 4.6.1 provides an analysis of available Mark III DWBT data to estimate the likelihood of the different DW bypass leakage categories.

4-35 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 4.6.1 DWBT Data Analysis Table 4.6-2 summarizes the available DWBT results for the Mark III containment types previously reported in Attachment 1, Table 5 of Reference [27] updated with the latest CPS test results. In the prior CPS DWBT extension analysis [19], 300 SCFM was used as the reference leakage for the risk assessment. This will also be used in this assessment for the base drywell leakage rate, DWLb. Therefore, the analysis is performed using the leakage characteristic of the as found state of the drywell. This recognizes both the historical results of the DWBT and the fact that Clinton continuously monitors the DW leakage. CPS is committed to trending this monitored information and noting any adverse trends (which there have been none). Based on these results and the continuous on-line monitoring, it is considered appropriate to use the conservatively high leakage rate of 300 scfm (DWLb)(1) as the baseline leakage characteristic of a 3/10-year DWBT frequency. This is conservative but is not as large as the Technical Specification allowable. The rationale for using a conservative but more realistic value than the Technical Specification leakage for the drywell is that the last six DWBTs show that the drywell leakage is below 31 scfm (see Table 4.6-2), which is more than two orders of magnitude below the Technical Specification limit (3654 scfm @ 3psig) [18]. The conservative analysis characterization of the DWBT using 300 scfm bounds even the initial drywell leakage (January 1986 test leakage from Table 4.6-2), which had defective electrical penetrations. These defective electrical penetrations were subsequently repaired.

Table 4.6-2 shows industry bypass test results provided in a previous CPS RAI response

[27] and updated with the latest (Feb-08) CPS test data. Additional data for recent tests at other sites may be available; however, this information is adequate to show Clinton performance relative to the industry.

(1) A realistic estimate would be closer to 30 scfm. This conservatism affects the population dose estimates.

4-36 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.6-2 MARK III DRYWELL BYPASS TEST RESULTS LEAKAGE RATE ACTUAL LEAKAGE /

SITE TEST DATE (SCFM) 300 SCFM Jan-86 273 0.91 Nov-86 20.8 0.07 Apr-89 18.8 0.06 Clinton Mar-91 21.9 0.07 May-92 18 0.06 Nov-93 30.2 0.10 Feb-08 20.18 0.07 Nov-85 2315 7.72 Nov-86 1568 5.23 Dec-87 1500 5.00 Grand Gulf Apr-89 1631 5.44 Nov-90 1591 5.30 May-92 618 2.06 Nov-93 869 2.90 Aug-87 124 0.41 Jul-89 123 0.41 Dec-90 797 2.66 Perry May-92 253 0.84 Jun-94 2450 8.17 Jul-94 111 0.37 Dec-87 602 2.01 May-89 141 0.47 River Bend Nov-90 345 1.15 Aug-92 754 2.51 Jun-94 421 1.40 Figure 4.6-1 shows a scatter plot of the data in Table 4.6-2 compared to the reference assumed base leakage value, DWLb, of 300 SCFM. (Note that the assumed base drywell leakage value of 300 SCFM is less than the allowable drywell bypass leakage for CPS of 3654 SCFM at 3.0 psid) [18]. Two of the test data are above 6DWLb and all test data is below 10DWLb. The 300 SCFM base case drywell leakage therefore represents a conservative assumption but is used for consistency with the previously accepted ILRT/DWBT extension requests for CPS.

4-37 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval The Technical Specification allowable leakage for the drywell is not used because of the on-line monitoring that is established by the past DWBT. Use of the Technical Specification limit would mischaracterize the Clinton drywell integrity and would make the decision not risk-informed. Therefore, the DW leakage is characterized in the analysis to be 1 times, 10 times, or 100 times a conservative characterization of the drywell leakage, which is referred to in this analysis as DWLb.

This leads to the specification of the drywell leakage rates consistent with the EPRI ILRT methodology:

Minimal leakage case 300 SCFM @ 3 psid (DWLb) 10 DWLb case 3000 SCFM @ 3 psid 100 DWLb case 30,000 SCFM @ 3 psid These represent conservative characterizations of the as found drywell bypass leakage.

The 100 DWLb case leakage rate of 30,000 SCFM @ 3 psid is less than the maximum allowable rate of 36,540 SCFM @ 3 psid.(1)

By definition, the containment leakage rate for Category 1 (i.e., accidents with containment leakage at or below maximum allowable Technical Specification leakage) is 1.0La (or 1.0 DWLb for the drywell).

(1) USAR section 6.2.6.5.1 notes that a leakage rate of 3,654 scfm is 10% of the maximum allowable leakage rate. (3,654 scfm

• 10 = 36,540 scfm) 4-38 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 10.00 8 .00 6 .00 4 .00 2 .00 T

0 .00

,,.. ,,.. ,,.. ,.. T 0 5 10 15 20 25 30 FIGURE 4.6-1 MARK Ill DWBT RESULTS COMPARED TO 300 SCFM X Axis= Test# from Table 4.6-2 Y Axis= Test Leakage+ 300 SCFM Only 2 of the data points are measurably in the 6-10 range, and all are below the 10 DWLb leakage rate assumed for the intermediate category in this assessment.

Drywell Probability of Small and Large Drywell Failures The base case probabilities for containment (Wetwell) failure probability are applied to the drywell small and large failure events as shown in Table 4.6-3. This is consistent with the approach used in the previous ILRT LARs [19, 36]. Applying these failure probabilities is appropriate for the following reasons:

• In the older BWR containment designs (i.e., Mark I and II), the drywell enclosure is also part of the containment enclosure. Therefore, the data used in the NEl/EPRI approach is reflective of drywell failures. The body of plant experience used considered the older BWR containment designs.

Therefore, the NEl/EPRI data is reflective of typical BWR drywell failure mechanisms.

4-39 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval x The CPS containment and drywell designs are similar in many of their construction details. A comparison of the containment and drywell design features is provided in Table 4.1-2. As this comparison shows, the basic designs are much the same and therefore would be expected to have much the same leakage failure mechanisms.

x As noted in Section 4.6.1, Clinton has the ability to continuously monitor the DW leakage. As noted in Section 4.6.1, small airline leaks cause the drywell to pressurize at a rate of approximately 0.03 psi/hr. The operators vent the drywell approximately once per day. It is unlikely that the instrument air leaks will diminish during operation. Therefore, if the drywell pressurization rate went to zero, this would indicate a small drywell leak may have caused this drop in the pressurization rate.

x As shown in Figure 4.6-1, no industry event has exceeded the 10La threshold.

Based on the above discussion, the EPRI ILRT containment leakage probabilities documented in Section 4.4 are applied to the DW and WW as shown in Table 4.6-3.

TABLE 4.6-3 DW AND WW LEAKAGE PROBABILITIES DW DW WW WW LEAKAGE LEAKAGE PROBABILITY LEAKAGE LEAKAGE SIZE (Lb) SIZE (La) PROBABILITY (BASE) 1Lb 0.9885(1) 1La 0.9885(1) 10Lb 0.0092 10La 0.0092 100Lb 0.0023 100La 0.0023 (1)The probability of assumed normal drywell leakage (1L ) is [1-a (Prob. of 10Lb + Prob. of 100Lb)].

These values are therefore used for the base case assessment to represent the DW bypass leakage behavior. Increases to these values are assumed to occur for the different test intervals consistent with the ILRT methodology.

The combined DW Leakage probability and WW Leakage probability are used in the analysis. The base case combined probabilities are shown in the Table 4.6-4:

4-40 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 4.6-4 DW AND WW COMBINED LEAKAGE PROBABILITIES DW CTMT DW CTMT LEAK LEAK COMBINED CASE LEAKAGE LEAKAGE PROB PROB PROB EPRI CLASS AA' 1 1 0.99 0.99 0.98 1 AB' 1 10 0.99 0.0092 0.0091 3a AC' 1 100 0.99 0.0023 0.0023 3b BA'1 10 1 0.0092 0.99 0.0091 1 BB'1 10 10 0.0092 0.0092 8.5E-5 3a BC'1 10 100 0.0092 0.0023 2.1E-5 3b CA'1 100 1 0.0023 0.99 0.0023 3a CB'1 100 10 0.0023 0.0092 2.1E-5 3b CC'1 100 100 0.0023 0.0023 5.3E-6 3b Combined Probabilities (Sum) 0.9862 1 0.0115 3a 0.00232 3b Containment Overpressure In the case of accident sequences that are the result of the long-term loss of containment heat removal, containment pressurization and eventual failure are assumed to result in a loss of core coolant injection systems. The CPS PRA models long-term loss of containment heat removal and the resultant loss of core coolant injection systems.

As part of the 2003 LAR [19], an assessment of the possibility that Clinton overpressure containment failures may increase in frequency due to the extension of the DWBT interval was performed by examining those sequences with the highest potential to cause such containment pressure increases. The USAR was reviewed to identify that the limiting condition was a 2 primary system LOCA in the drywell. Using this information and the identified allowable leak areas, several confirmatory MAAP cases were performed to demonstrate the containment challenges for varying bypass flow areas.

4-41 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval The assessment documented the following:

x The containment pressurization due to a LOCA is insensitive to relatively large variations in the DW Bypass area and does not exceed 20 psia except for the worst case postulated condition of a 2 LOCA and maximum Technical Specification Bypass.

x The pressure suppression capability of the containment is robust.

x The large volume in the outer containment minimizes the effects of changes in the drywell bypass flow area.

x Any effects of the containment pressurization due to drywell bypass leakage can be effectively terminated by:

a) RPV depressurization which is directed by the EOPs on exceeding the pressure suppression pressure or b) Containment sprays which are directed by the EOPs upon exceeding relatively low containment pressures Both of these operating crew actions can be completed over many hours and therefore their success probability is high.

x Subsequent peaks of 30-40 psia in the containment pressure are due to hydrogen combustion events.

The conclusion from this investigation is that containment failure induced by containment pressurization aggravated by the drywell bypass leakage change is highly unlikely. The relatively small changes postulated due to the DWBT interval extension make no appreciable change in the containment pressurization compared to its ultimate capability. The containment overpressure challenges due to the loss of containment heat removal capability are already accounted for in the Clinton PSA. As such, the perturbation on these sequences caused by slight changes in the drywell bypass area are considered negligible contributors to CDF.

The pressurization issue was addressed in the Safety Evaluation Report (SER) that was part of the NRC letter [11] approving the one-time 15-year interval extension. The SER noted the following:

During a small-break loss-of-coolant accident, potential leak paths between the drywell and containment airspace could result in excessive containment pressure if the steam flow into the airspace would bypass the vapor suppression capabilities of the pool. The potential leakage paths between the drywell and the containment are: 1) piping and electrical penetrations; 2) the drywell equipment hatch; and 3) the drywell personnel air lock. The staff found that 1) the electrical penetrations are unlikely to leak significantly, and the design drywell bypass leakage rate is so large that, 4-42 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval even if the valves in many of the pipes were left open, the design limit would not be exceeded; and 2) both the equipment hatch and drywell air lock have double compression seals and are leak tested in accordance with TSs.

Based on the significant margin found in the 2003 LAR MAAP runs and the deterministic arguments noted above, there is no change in CDF due to the small increases in drywell bypass leakage associated.

4-43 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 5.0 RESULTS The application of the approach based on EPRI Guidance [3] leads to the following results. The results are displayed according to the eight accident classes defined in the EPRI report. Table 5.0-1 lists these accident classes.

TABLE 5.0-1 ACCIDENT CLASSES ACCIDENT CLASSES (CONTAINMENT RELEASE TYPE) DESCRIPTION 1 No Containment Failure 2 Large Isolation Failures (Failure to Close) 3a Small Isolation Failures 3b Large Isolation Failures 4 Small Isolation Failures (Failure to seal -Type B) 5 Small Isolation Failures (Failure to sealType C) 6 Other Isolation Failures (e.g., dependent failures) 7 Failures Induced by Phenomena (Early and Late) 8 Containment Bypass CDF All CET End states (including very low and no release)

The analysis examined CPS-specific accident sequences in which the containment remains intact or the containment is impaired. Specifically, the categorization of the severe accidents contributing to risk was considered in the following manner:

x Core damage sequences in which the containment remains intact initially and in the long term (EPRI Class 1 sequences).

x 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 bellows leakage, if applicable. (EPRI Class 3 sequences).

x Core damage sequences in which containment integrity is impaired due to containment isolation failures of pathways left opened following a plant post-maintenance test. (For example, a valve failing to close following a valve stroke test. (EPRI Class 6 sequences). Consistent with the EPRI Guidance, this class is not specifically examined since it will not significantly influence the results of this analysis.

5-1 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval x Accident sequences involving containment bypass (EPRI Class 8 sequences), large containment isolation failures (EPRI Class 2 sequences),

and small containment isolation failure-to-seal events (EPRI Class 4 and 5 sequences) are accounted for in this evaluation as part of the baseline risk profile. However, they are not affected by the ILRT frequency change.

x 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.

The steps taken to perform this risk assessment evaluation are as follows:

Step 1 Quantify the base-line risk in terms of frequency per reactor year for each of the eight accident classes presented in Table 5.0-1.

Step 2 Develop plant-specific person-rem dose (population dose) per reactor year for each of the eight accident classes.

Step 3 Evaluate the risk impact (i.e., the change in containment release scenario type frequency and population dose) of extending the ILRT interval from 15 years to 15.7 years .

Step 4 Determine the change in risk in terms of Large Early Release Frequency (LERF) in accordance with RG 1.174.

Step 5 Determine the impact on the CCFP.

Step 6 Evaluate the impact of external events.

It is noted that the calculations were generally performed using a spreadsheet such that the presented numerical results may differ very slightly as compared to values if calculated by hand.

5.1 STEP 1 - QUANTIFY THE BASE-LINE RISK IN TERMS OF FREQUENCY PER REACTOR YEAR The CPS PRA Level 2 Model [32] is used to develop the initial set of internal events release categories for use in this analysis. As described in Section 4.3, the release categories were assigned to the EPRI classes as shown in Table 4.3-3. This application combined with the CPS dose risk (person-rem/yr) as shown in Table 4.3-9 forms the basis for estimating the increase in population dose risk.

5-2 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval For the assessment of the impact on the risk profile due to the ILRT extension, the potential for pre-existing leaks is included in the model. These pre-existing leak events are represented by the Class 3 sequences in EPRI TR-1018243 [3]. Two failure modes were considered for the Class 3 sequences, namely Class 3a (small breach) and Class 3b (large breach).

The determination of the frequencies associated with each of the EPRI categories listed in Table 5.0-1 is presented next. Since the Class 1 frequency is determined based on remaining contribution not assigned to other classes, the discussion appears in reverse order starting with EPRI Class 8 and ending with EPRI Class 1. However, EPRI Class 2 is discussed prior to Class 3 since its value is used in the final determination of the Class 3 frequencies.

Class 8 Sequences This group represents sequences where containment bypass occurs. The failure frequency for Class 8 sequences is 3.88E-08/yr, as documented in Table 4.2-4.

Class 7 Sequences Dose Risk Adjustments Class 7 consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs. For this analysis, the associated radionuclide releases are based on the application of the Level 2 end states to the Accident Progression Bins from NUREG/CR-4551 as described in Section 4.2.

The Class 7 Sequences are divided into 2 categories, which consist of LERF sequences (excluding the BOC and ISLOCA Class V sequences) and non-LERF sequences. The second category (non-LERF) is further divided into Bins 1, 3, 4, 5, 6, and 8 from NUREG/CR-4551. These non-LERF sequences are grouped into Accident Classes 7a, 7b, 7c, 7d, 7e, and 7f as documented in Table 4.3-8. The failure frequency and population dose for each specific APB are shown in 4.3-8. As shown in Table 4.3-8, the population dose person-rem for Class 7 LERF sequences is based on the largest APB value in 5-3 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval NUREG/CR-4551, while population dose person-rem for Class 7 non-LERF sequences is based on a weighted average of six APB bins.

Class 6 Sequences These are sequences that involve core damage with a failure-to-seal containment leakage due to failure to isolate the containment. These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution.

Consistent with the EPRI guidance, this accident class is not explicitly considered since it has a negligible impact on the results.

Class 5 Sequences This group represents containment isolation failure-to-seal of Type C test components.

Because these 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.

Class 4 Sequences This group represents containment isolation failure-to-seal of Type B test components.

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 this analysis.

Class 2 Sequences This group consists of large containment isolation failures. The failure frequency for Class 2 sequences is 4.39E-07/yr, as documented in Table 4.2-4. Note that this frequency is not affected by the ILRT interval change.

Class 3 Sequences This group represents pre-existing leakage in the containment structure. The containment leakage for these sequences can be either small (in excess of design allowable but <10La) or large. In this analysis, a value of 10La was used for small pre-existing flaws and 100La 5-4 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval for large flaws, consistent with the EPRI methodology [3]. The probability of Class 3a and Class 3b leakages are combined leakage probabilities from Table 4.6-4 to account for the impact of drywell leakage on the ILRT interval extension risk metrics. As described in Section 4.4, additional consideration is made to not apply these failure probabilities to those cases that are already classified as LERF (i.e., Class 7 LERF and Class 8 LERF contributions), or would never lead to a LERF (EPRI Class 7 non-LERF contribution consisting of Class II and Class IBL sequences).

Class_3a = 0.0115 * [CDF - (EPRI Class 7 LERF + EPRI Class 8 + Class II + Class IBL)]

= 0.0115 * [3.33E (1.27E-07 + 3.88E-08 + 6.92E-07 + 7.01E-07)]

= 2.03E-08/yr Class_3b = 0.00232 * [CDF - (EPRI Class 7 LERF + EPRI Class 8 + Class II + Class IBL)]

= 0.00232 * [3.33E (1.27E-07 + 3.88E-08 + 6.92E-07 + 7.01E-07)]

= 4.11E-09/yr For this analysis, the associated containment leakage for Class 3a and Class 3b is 10La and 100La, respectively, which is consistent with the EPRI methodology [3].

Class 1 Sequences This group represents the frequency when the containment remains intact (modeled as Technical Specification Leakage). The frequency per year for these sequences is 9.37E-07/yr for CPS and is determined by subtracting all containment failure end states, including the EPRI/NEI Class 3a and 3b frequencies calculated above, from the total CDF.

Class 1 = CDF - (EPRI Classes)

= 3.33E (4.39E-07 (class 2) + 2.03E-08 (3a) + 4.11E-09 (3b) + 1.27E-07 (7 LERF) + 1.76E-06 (7-Non-LERF) + 3.88E-08 (Class 8))

= 9.37E-07/yr 5-5 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval For this analysis, the associated maximum containment leakage for this group is 1La, consistent with an intact containment evaluation. Note that the value for this Class reported in Table 5.1-1 is slightly lower than that reported in Tables 4.2-3 since the 3a and 3b frequencies are now subtracted from Class 1.

Summary of Accident Class Frequencies In summary, the accident sequence frequencies that can lead to release of radionuclides to the public have been derived in a manner consistent with the definition of accident classes defined in EPRI TR-1018243 [3] and are shown in Table 5.1-1.

TABLE 5.1-1 RADIONUCLIDE RELEASE FREQUENCIES AS A FUNCTION OF ACCIDENT CLASS (CPS BASE CASE)

ACCIDENT CLASSES (CONTAINMENT FREQUENCY RELEASE TYPE) DESCRIPTION (1/YR) 1 No Containment Failure 9.37E-07 2 Large Isolation Failures (Failure to Close) 4.39E-07 3a Small Isolation Failures 2.03E-08 3b Large Isolation Failures 4.11E-09 4 Small Isolation Failures (Failure to seal -Type B) N/A 5 Small Isolation Failures (Failure to sealType C) N/A 6 Other Isolation Failures (e.g., dependent failures) N/A 7 LERF Failures Induced by Phenomena (LERF) 1.27E-07 7 non-LERF Failures Induced by Phenomena (non-LERF) 1.76E-06 8 Containment Bypass 3.88E-08 CDF All CET End states (including intact case) 3.33E-06 5-6 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 5.2 STEP 2 - DEVELOP PLANT-SPECIFIC PERSON-REM DOSE (POPULATION DOSE) PER REACTOR YEAR Plant-specific release analyses were performed to estimate the weighted average person-rem doses to the population within a 50-mile radius from the plant. The releases are based on the assessment provided in Section 4.3 for CPS (see Table 4.3-9 of this analysis). The results of applying these releases to the EPRI containment failure classifications are summarized as follows:

Class 1 = 2.71E+03 person-rem (at 1.0La)

Class 2 = 5.48E+05 person-rem Class 3a = 2.71E+03 person-rem x 10La = 2.71E+04 person-rem Class 3b = 2.71E+03 person-rem x 100La = 2.71E+05 person-rem Class 4 = Not analyzed Class 5 = Not analyzed Class 6 = Not analyzed Class 7 LERF = 5.48E+05 person-rem Class 7 non-LERF = 3.28E+05 person-rem Class 8 = 5.48E+05 person-rem In summary, the population dose estimates derived for use in the risk evaluation per the EPRI methodology [3] for all EPRI classes are provided in Table 5.2-1, which includes the values previously presented in Table 4.3-9 as well as the Class 3a and 3b population doses calculated above.

5-7 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 5.2-1 CPS POPULATION DOSE FOR POPULATION WITHIN 50 MILES ACCIDENT CLASSES (CONTAINMENT PERSON-REM RELEASE TYPE) DESCRIPTION (0-50 MILES) 1 No Containment Failure (1La) 2.71E+03 2 Large Isolation Failures 5.48E+05 (Failure to Close) 3a Small Isolation Failures 2.71E+04 3b Large Isolation Failures 2.71E+05 4 Small Isolation Failures NA (Failure to seal -Type B) 5 Small Isolation Failures NA (Failure to sealType C) 6 Other Isolation Failures NA (e.g., dependent failures) 7 LERF Failures Induced by Phenomena (LERF) 5.48E+05 7 non-LERF Failures Induced by Phenomena (non-LERF) 3.28E+05 8 LERF Containment Bypass 5.48E+05 The above population dose, when multiplied by the frequency results presented in Table 5.1-1, yields the CPS baseline mean dose risk for each EPRI accident class. These results are presented in Table 5.2-2.

5-8 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 5.2-2 CPS ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR 3 IN 10 YEAR ILRT FREQUENCY EPRI METHODOLOGY PLUS EPRI METHODOLOGY CORROSION (2)

ACCIDENT CHANGE DUE TO CLASSES PERSON- PERSON- PERSON- CORROSION (CONTAINMENT REM FREQUENCY REM/YR FREQUENCY REM/YR (PERSON-RELEASE TYPE) DESCRIPTION (0-50 MILES) (1/YR) (0-50 MILES) (1/YR) (0-50 MILES) REM/YR) 1(1) No Containment Failure 2.71E+03 9.37E-07 2.54E-03 9.37E-07 2.54E-03 -4.27E-08 2 Large Isolation Failures 5.48E+05 4.39E-07 2.41E-01 4.39E-07 2.41E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 2.03E-08 5.51E-04 2.03E-08 5.51E-04 --

3b Large Isolation Failures 2.71E+05 4.11E-09 1.12E-03 4.13E-09 1.12E-03 4.27E-06 7 LERF Failures Induced by 5.48E+05 1.27E-07 6.94E-02 1.27E-07 6.94E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.28E+05 1.76E-06 5.79E-01 1.76E-06 5.79E-01 --

Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 3.88E-08 2.13E-02 3.88E-08 2.13E-02 --

CDF All CET end states 3.33E-06 0.914 3.33E-06 0.914 4.23E-06 Notes to Table: 5.2-2:

(1)

Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

(2)

Only release Classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every five years.

5-9 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval 5.3 STEP 3 - EVALUATE RISK IMPACT OF EXTENDING TYPE A TEST INTERVAL FROM 15-TO-15.7 YEARS The next step is to evaluate the risk impact of extending the test interval from its current 15-year value to 15.7 years. To do this, an evaluation must first be made of the risk associated with the 15-year interval.

Risk Impact Due to 15-year Test Interval As previously stated, ILRT 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). The risk contribution is changed based on the EPRI guidance as described in Section 4.4 by a factor of 5 compared to the base case values. The results of the calculation for a 15-year interval are presented in Table 5.3-1.

Risk Impact Due to 15.7-Year Test Interval The risk contribution for a 15.7-year interval is calculated in a manner similar to the 15-year interval. The difference is in the increase in probability of not detecting a leak in Classes 3a and 3b for the ILRT Type A tests. For this case, the value used in the analysis is a factor of 5.23 compared to the 3-year interval value, as described in Section 4.4, and a factor of 1.047 compared to the 15-year interval value calculated above. The results for this calculation are presented in Table 5.3-2.

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Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 5.3-1 CPS ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT FREQUENCY EPRI METHODOLOGY PLUS EPRI METHODOLOGY CORROSION (2) CHANGE DUE TO ACCIDENT CORROSION CLASSES PERSON- PERSON- PERSON- EXTENSION (CONTAINMENT REM FREQUENCY REM/YR FREQUENCY REM/YR (PERSON-RELEASE TYPE) DESCRIPTION (0-50 MILES) (1/YR) (0-50 MILES) (1/YR) (0-50 MILES) REM/YR) 1(1) No Containment Failure 2.71E+03 8.39E-07 2.28E-03 8.39E-07 2.28E-03 -5.68E-07 2 Large Isolation Failures 5.48E+05 4.39E-07 2.41E-01 4.39E-07 2.41E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 1.01E-07 2.75E-03 1.01E-07 2.75E-03 --

3b Large Isolation Failures 2.71E+05 2.06E-08 5.58E-03 2.08E-08 5.64E-03 5.68E-05 7 LERF Failures Induced by 5.48E+05 1.27E-07 6.94E-02 1.27E-07 6.94E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.28E+05 1.76E-06 5.79E-01 1.76E-06 5.79E-01 Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 3.88E-08 2.13E-02 3.88E-08 2.13E-02 --

CDF All CET end states 3.33E-06 0.921 3.33E-06 0.921 5.63E-05 Notes to Table 5.3-1:

(1)

Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

(2)

Only release Classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every five years.

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Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 5.3-2 CPS ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR 1 IN 15.7 YEAR ILRT FREQUENCY EPRI METHODOLOGY PLUS EPRI METHODOLOGY CORROSION (2) CHANGE DUE TO ACCIDENT PERSON- CORROSION CLASSES REM PERSON- PERSON- EXTENSION (CONTAINMENT (0-50 FREQUENCY REM/YR FREQUENCY REM/YR (PERSON-RELEASE TYPE) DESCRIPTION MILES) (1/YR) (0-50 MILES) (1/YR) (0-50 MILES) REM/YR) 1(1) No Containment Failure 2.71E+03 8.33E-07 2.26E-03 8.33E-07 2.26E-03 -6.61E-07 2 Large Isolation Failures 5.48E+05 4.39E-07 2.41E-01 4.39E-07 2.41E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 1.06E-07 2.88E-03 1.06E-07 2.88E-03 --

3b Large Isolation Failures 2.71E+05 2.15E-08 5.84E-03 2.18E-08 5.90E-03 6.61E-05 7 LERF Failures Induced by 5.48E+05 1.27E-07 6.94E-02 1.27E-07 6.94E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.28E+05 1.76E-06 5.79E-01 1.76E-06 5.79E-01 Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 3.88E-08 2.13E-02 3.88E-08 2.13E-02 --

CDF All CET end states 3.33E-06 0.921 3.33E-06 0.921 6.55E-05 Notes to Table 5.3-2:

(1)

Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

(2)

Only release Classes 1 and 3b are affected by the corrosion analysis. During the 15-year interval, the failure rate is assumed to double every five years.

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Risk Impact Assessment of Extending the CPS ILRT Interval 5.4 STEP 4 - DETERMINE THE CHANGE IN RISK IN TERMS OF LARGE EARLY RELEASE FREQUENCY Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of CDF below 1E-6/yr and increases in LERF below 1E-7/yr, and small changes in LERF as below 1E-6/yr. Because the ILRT interval extension does not impact CDF, the relevant metric is LERF.

For CPS, 100% of the frequency of Class 3b sequences can be used as a 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 15-year test interval from Table 5.3-1, the Class 3b frequency is 2.08E-08/yr; based on a 15.7-year test interval from Table 5.3-2, the Class 3b frequency is 2.18E-08/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 15 to 15.7 years (including corrosion effects) is 9.80E-10/yr. As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology), the estimated change in LERF is within Region III of Figure 4 of Reference

[4] (very small changes in LERF) when comparing the 15.7-year results to the 15-year results.

5.5 STEP 5 - DETERMINE THE IMPACT ON THE CONDITIONAL CONTAINMENT FAILURE PROBABILITY Another parameter that the NRC guidance in RG 1.174 states can provide input into the decision-making process is the change in the CCFP. The change in CCFP is indicative of the effect of the ILRT on all radionuclide releases, not just LERF. The CCFP can be calculated from the results of this analysis. One of the difficult aspects of this calculation is providing a definition of the failed containment. In this assessment, the CCFP is defined such that containment failure includes all radionuclide release end states other than the intact state. The conditional part of the definition is conditional given a severe accident (i.e., core damage).

5-13 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval The change in CCFP can be calculated by using the method specified in the EPRI methodology [3]. The NRC has previously accepted similar calculations [7] as the basis for showing that the proposed change is consistent with the defense-in-depth philosophy.

The following table shows the CCFP values that result from the assessment for the various testing intervals including corrosion effects in which the flaw rate is assumed to double every five years.

CCFP CCFP

'CCFP15.7-15

1 IN 15 YRS 1 IN 15.7 YRS 71.76% 71.79% 0.03%

CCFP = [1 - (Class 1 frequency + Class 3a frequency)/CDF] x 100%

The change in CCFP of approximately 0.03% as a result of extending the test interval to 15.7 years from 15 years is judged to be insignificant.

5.6

SUMMARY

OF INTERNAL EVENTS RESULTS Table 5.6-1 summarizes the internal events results of this ILRT extension risk assessment for CPS. The internal events risk results associated with a change in the test interval from 1-in-15 years to 1-in-15.7 years all are below the acceptance criteria defined in Section 1.3, namely:

1. Change in LERF = 9.80E-10/yr, which is less than 1.0E-7/yr for the very small risk increase as defined in RG 1.174.

2. Change in population dose rate is 3.78E-4 person-rem/yr (0.04%), which is less than 1.0 person-rem/year or 1% of the total population dose.

3. Change in CCFP is 0.03%, which is less than 1.5%.

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Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 5.6-1 CPS ILRT CASES:

BASE AND 15.7 YR EXTENSION (INCLUDING AGE ADJUSTED STEEL CORROSION LIKELIHOOD)

BASE CASE EXTEND TO 1 IN 15 YEARS 1 IN 15.7 YEARS EPRI DOSE PERSON- PERSON-CLASS PER-REM CDF (1/YR) REM/YR CDF (1/YR) REM/YR 1 2.71E+03 8.39E-07 2.28E-03 8.33E-07 2.26E-03 2 5.48E+05 4.39E-07 2.41E-01 4.39E-07 2.41E-01 3a 2.71E+04 1.01E-07 2.75E-03 1.06E-07 2.88E-03 3b 2.71E+05 2.08E-08 5.64E-03 2.18E-08 5.90E-03 7 LERF 5.48E+05 1.27E-07 6.94E-02 1.27E-07 6.94E-02 7 non-LERF 3.28E+05 1.76E-06 5.79E-01 1.76E-06 5.79E-01 8 5.48E+05 3.88E-08 2.13E-02 3.88E-08 2.13E-02 Total 3.33E-06 0.921 3.33E-06 0.921 Dose Rate --- 3.78E-04 3b LERF --- 9.80E-10 CCFP % 71.76% 71.79%

CCFP % --- 0.03%

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Risk Impact Assessment of Extending the CPS ILRT Interval 5.7 STEP 6 - CONTRIBUTIONS FROM EXTERNAL HAZARDS Since the risk acceptance guidelines in RG 1.174 are intended for comparison with a full-scope assessment of risk, including internal and external events, a bounding analysis of the potential impact from external events and other hazard groups is presented here because the external event models do not have a full Level 2 model. Appendix A provides a technical adequacy assessment of the Seismic CDF and LERF and the Fire PRA.

Internal Fire Risk [8]

The current CPS Fire PRA model (CL117BF0) is based upon the current FPIE PRA model (i.e., CL117B), both of which were initially updated as part of the 2017 periodic update.

The total Fire PRA CDF reported is 7.75E-05/yr; the Fire PRA LERF is 5.30E-06/yr [8].

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:







Seismic Risk [38]

A quantifiable seismic PRA model for Clinton has not yet been approved for general use in risk applications. However, a calculation for seismic CDF and LERF estimates for the CPS TSTF-505 (RICT) Program, including a review of available industry SSC Fragility Information, was performed [38]. This analysis yielded a seismic CDF of 6.4E-6/yr and a seismic LERF of 1.6E-6/yr. Subtracting seismic LERF from CDF, the Class 3b frequency can be calculated by the following formulas:







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Risk Impact Assessment of Extending the CPS ILRT Interval Other External Events In addition to internal fires and seismic events, the CPS IPEEE Submittal [34] analyzed a variety of other external hazards:

x High Winds/Tornadoes x External Flooding x Transportation and Nearby Facility Accidents x Other External Hazards The CPS IPEEE analysis of high winds, tornadoes, external floods, transportation accidents, nearby facility accidents, and other external hazards was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards.

Based upon this review, it was concluded that CPS meets the applicable Standard Review Plan requirements and therefore has an acceptably low risk with respect to these hazards. As such, these hazards were determined in the Clinton IPEEE to be negligible contributors to overall plant risk.

Accordingly, these other external event hazards are not included explicitly in this section and are reasonably assumed not to impact the results or conclusions of the ILRT interval extension risk assessment.

External Event Contributor Summary The external event contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The LERF increase is conservatively assumed to be the change in Class 3b frequency.

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Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 5.7-1 CPS 3B (LERF) AS A FUNCTION OF ILRT FREQUENCY FOR INTERNAL AND EXTERNAL EVENTS 3B FREQUENCY 3B FREQUENCY LERF (1-PER-15 YEAR ILRT) (1-PER-15.7 YEAR ILRT) INCREASE(1)

External Events 8.97E-07 9.38E-07 4.13E-08 Internal Events 2.08E-08 2.18E-08 9.80E-10 Combined 9.18E-07 9.60E-07 4.22E-08 Thus, the total increase in LERF (measured from the baseline 1-per-15 year ILRT interval to the proposed 1-per-15.7 year frequency) due to the combined internal and external events contribution is estimated as 4.22E-08/yr, which meets the guidance for very small change in risk, as it is less than 1.0E-7/yr [4].

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Risk Impact Assessment of Extending the CPS ILRT Interval 5.8 16-YEAR ILRT INTERVAL SENSITIVITY While it is expected the 15.7-year ILRT interval provides ample time for the reasons for the extension request to be mitigated, a sensitivity is performed with a 12-month one-time extension interval (16-year total interval). Table 5.8-1 summarizes the internal events results of this ILRT extension risk assessment for CPS. The internal events risk results associated with a change in the test interval from 1-in-15 years to 1-in-16 years all are less than the acceptance criteria defined in Section 1.3.

TABLE 5.8-1 CPS ILRT CASES:

BASE AND 16 YR EXTENSION (INCLUDING AGE ADJUSTED STEEL CORROSION LIKELIHOOD)

BASE CASE EXTEND TO 1 IN 15 YEARS 1 IN 16 YEARS EPRI DOSE PERSON- PERSON-CLASS PER-REM CDF (1/YR) REM/YR CDF (1/YR) REM/YR 1 2.71E+03 8.39E-07 2.28E-03 8.31E-07 2.25E-03 2 5.48E+05 4.39E-07 2.41E-01 4.39E-07 2.41E-01 3a 2.71E+04 1.01E-07 2.75E-03 1.08E-07 2.94E-03 3b 2.71E+05 2.08E-08 5.64E-03 2.22E-08 6.02E-03 7 LERF 5.48E+05 1.27E-07 6.94E-02 1.27E-07 6.94E-02 7 non-LERF 3.28E+05 1.76E-06 5.79E-01 1.76E-06 5.79E-01 8 5.48E+05 3.88E-08 2.13E-02 3.88E-08 2.13E-02 Total 3.33E-06 0.921 3.33E-06 0.921 Dose Rate --- 5.43E-04 3b LERF --- 1.41E-09 CCFP % 71.76% 71.80%

CCFP % --- 0.04%

Using the same methodology as in Section 5.7, the external event contributions to Class 3b frequencies are combined to obtain the total external event contribution to Class 3b frequencies with a sensitivity 16-year interval.

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Risk Impact Assessment of Extending the CPS ILRT Interval TABLE 5.8-2 CPS 3B (LERF) AS A FUNCTION OF ILRT FREQUENCY FOR INTERNAL AND EXTERNAL EVENTS 3B FREQUENCY 3B FREQUENCY LERF (1-PER-15 YEAR ILRT) (1-PER-16 YEAR ILRT) INCREASE(1)

External Events 8.97E-07 9.57E-07 5.97E-08 Internal Events 2.08E-08 2.22E-08 1.41E-09 Combined 9.18E-07 9.79E-07 6.11E-08 Thus, the total increase in LERF (measured from the baseline 1-per-15 year ILRT interval to the postulated 1-per-16 year sensitivity frequency) due to the combined internal and external events contribution is estimated as 6.11E-08/yr, which meets the guidance for very small change in risk, as it is less than 1.0E-7/yr [Reference 4].

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Risk Impact Assessment of Extending the CPS ILRT Interval

6.0 CONCLUSION

S Based on the results from Section 5, the following conclusions regarding the assessment of the plant risk are associated with temporarily extending the Type A ILRT test frequency to 15.7 years:

x Reg. Guide 1.174 [4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Reg. Guide 1.174 defines very small changes in risk as resulting in increases of CDF below 1E-6/yr and increases in LERF below 1E-7/yr. Small changes in risk are defined as increases in CDF below 1E-5/yr and increases in LERF below 1E-6/yr.

Since the ILRT extension has no impact on CDF for CPS, the relevant criterion is LERF. The increase in internal events LERF resulting from a change in the Type A ILRT interval with corrosion included is 9.80E-10/yr (see Table 6.1-1), which falls within the very small change region of the acceptance guidelines in Reg. Guide 1.174.

x The change in dose risk for changing the Type A ILRT interval from once-per-15 years to once-per-15.7-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.78E-04 person-rem/yr using the EPRI guidance with the corrosion included (see Table 5.6-1). NEI 94-01 [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.

x The increase in the conditional containment failure probability from the 15-year interval to the 15.7-year interval including corrosion effects using the EPRI guidance (see Table 5.6-1) is 0.03%, which is less than the acceptance criteria of 1.5% identified in the NRC SER on the issue [7] as discussed in Section 1.3.

x To determine the potential impact from external events, an additional assessment from the risk associated with external events was performed.

As shown in Table 5.7-1, the total increase in LERF due to internal events and external events is 4.22E-08/yr, which falls within the very small change region of the acceptance guidelines in Reg. Guide 1.174.

Therefore, increasing the ILRT interval on a one-time basis to a one-in-15.7-year frequency is not considered to be significant since it represents only a small change in the CPS risk profile.

6-1 CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval Previous Assessments The NRC in NUREG-1493 [6] has previously concluded the following:

x Reducing the frequency of Type A tests (ILRTs) from three per 10 years to one 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 and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

x 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 one in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test the integrity of the containment structure.

The findings for CPS confirm these general findings on a plant specific basis for the ILRT interval extension considering the severe accidents evaluated for CPS, the CPS containment failure modes, and the local population surrounding CPS.

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Risk Impact Assessment of Extending the CPS ILRT Interval

7.0 REFERENCES

1. Nuclear Energy Institute, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, NEI 94-01, Revision 3-A, July 2012.

2. Electric Power Research Institute, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, EPRI TR-104285, August 1994.

3. Electric Power Research Institute, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325. EPRI TR-1018243, October 2008.

4. U.S. Nuclear Regulatory Commission, 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. Letter from Mr. C. H. Cruse (Constellation Nuclear, Calvert Cliffs Nuclear Power Plant) to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Accession Number ML020920100, March 27, 2002.

6. U.S. Nuclear Regulatory Commission, Performance-Based Containment Leak-Test Program, NUREG-1493, September 1995.

7. U.S. Nuclear Regulatory Commission, 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), Accession Number ML081140105, June 25, 2008.

8. CPS Fire PRA Summary and Quantification Notebook, CL-PRA-021.11, Revision 2, February 2020.

9. ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RAS-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.

10. United States Census Bureau QuickFacts, accessed January 2021, https://www.census.gov/quickfacts/.

11. U. S Nuclear Regulatory Commission,

Subject:

CLINTON POWER STATION, UNIT 1 - ISSUANCE OF AMENDMENT (TAC NO. M137675), Issuance of License Amendment 160 and Safety Evaluation of One-time Extension from 10 to 15 Years, Accession Number ML033360470, January 8, 2004.

12. ERIN Engineering and Research, Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAMTM, EPRI TR-105189, Final Report, May 1995.

13. Oak Ridge National Laboratory, Impact of Containment Building Leakage on LWR Accident Risk, NUREG/CR-3539, ORNL/TM-8964, April 1984.

14. Pacific Northwest Laboratory, Reliability Analysis of Containment Isolation Systems, NUREG/CR-4220, PNL-5432, June 1985.

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15. U.S. Nuclear Regulatory Commission, Technical Findings and Regulatory Analysis for Generic Safety Issue II.E.4.3 (Containment Integrity Check), NUREG-1273, April 1988.

16. Pacific Northwest Laboratory, Review of Light Water Reactor Regulatory Requirements, NUREG/CR-4330, PNL-5809, Vol. 2, June 1986.

17. Sandia National Laboratories, Evaluation of Severe Accident Risks: Grand Gulf, Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 6, Revision 1, Part 1, December 1990.

18. Exelon, Clinton Updated Safety Analysis Report (USAR), Rev. 21, March 2020.

19. Letter from Keith R. Jury (AmerGen Energy Company, LLC, for Clinton Power Station) to U.S. Nuclear Regulatory Commission, Request for Amendment to Technical Specifications 3.6.5.1, Drywell" and 5.5.13, "Primary Containment Leakage Rate Testing Program", Accession Number ML030370524, January 29, 2003.

20. Letter from Jerry C. Roberts (Entergy Operations, for Grand Gulf Nuclear Station) to U.S. Nuclear Regulatory Commission, License Amendment Request One-time Extension of the Integrated Leak Rate Test and Drywell Bypass Test Interval, Accession Number ML031400345, May 12, 2003.

21. Letter from Rick J. King (River Bend Regulatory Assurance) to U.S. Nuclear Regulatory Commission, License Amendment Request One-time Extension of the Integrated Leak Rate Test and Drywell Bypass Test Interval, Accession Number ML040540445, February 16, 2004.

22. U.S. Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.

23. U.S. Nuclear Regulatory Commission, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, NUREG-1150, December 1990.

24. CPS PRA Summary Notebook (2017 PRA Update), CPS PSA-013, Revision 6, February 2020.

25. CPS PRA Quantification Notebook, CPS PSA-014, Revision 9, February 2020.

26. Letter from P. B. Cowan (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information -

License Amendment Request for Type A Test Extension, Accession Number ML100560433, February 25, 2010.

27. Letter from Keith R. Jury (AmerGen Energy Company, LLC, for Clinton Power Station) to U.S. Nuclear Regulatory Commission, Additional Information Supporting the Request for Amendment to Technical Specifications 3.6.5.1, "Drywell" and 5.5.13, "Primary Containment Leakage Rate Testing Program",

Accession Number ML032671333, September 15, 2003

28. 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.

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Risk Impact Assessment of Extending the CPS ILRT Interval

29. Sandia National Laboratories, Evaluation of Severe Accident Risks: Grand Gulf, Unit 1, Appendices NUREG/CR-4551, SAND86-1309, Volume 6, Revision 1, Part 2, December 1990.

30. Illinois Department of Public Health (IDPH) Population Projections - Illinois, Chicago and Illinois Counties by Age and Sex: July 1, 2010 to July 1, 2025 (2014 edition).

31. SECPOP 4.2, Sector Population, Land Fraction, and Economic Estimation Program, Sandia National Laboratories.

32. CPS PRA Detailed Level 2 Evaluation Notebook, CPS PSA-015, Revision 4, February 2020.

33. Sandia National Laboratories, Evaluation of Severe Accident Risks: Quantification of Major Input Parameters, MACCS Input, NUREG/CR-4551, SAND86-1309, Volume 2, Revision 1, Part 7, December 1990.

34. Clinton Power Station Individual Plant Examination For External Events, September 1995.

35. 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.

36. Letter from Patrick Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, License Amendment Request to Revise Technical Specification Section 5.5.13, "Primary Containment Leakage Rate Testing Program," for Permanent Extension of Type A and Type C Leak Rate Test Frequencies, January 25, 2016.

37. Letter from Jennivine Rankin (U.S. Nuclear Regulatory Commission) to Bryan Hanson (Exelon Generation Company, LLC), Clinton Power Station, Unit No. 1 -

Issuance of Amendment Regarding Permanent Extension of Type A and Type C Leak Rate Test Frequencies (CAC No. MF7290), September 26, 2017.

38. External Hazards Assessment for Clinton Power Station, CL-MISC-026, Revision 0, March 2020.

39. Report # 032362-RPT-12, Clinton PRA Finding and Suggestion Level Fact and Observation Independent Assessment, Rev. 0, January 2019.

40. CPS PRA Deterministic Calculations Notebook, CPS PSA-007, Revision 2, December 2011.

41. Report # 032434-RPT-01, Clinton Power Station Fire PRA Finding Level Fact and Observation Closure by Independent Assessment, Rev. 0, February 2020.

42. Regulatory Guide 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 3, December 2020.

43. NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines," Rev. 1, June 2010.

44. NEI 05-04, "Process for Performing PRA Peer Reviews Using the ASME PRA Standard," Rev. 2, November 2008.

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45. Clinton Power Station Fire PRA Peer Review Report Using ASME/ANS PRA Standard Requirements, Boiling Water Reactor Owners Group (BWROG), August 2018.

46. Letter from Patrick Simpson (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, Application to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," April 2020.

47. Clinton Power Station 2009 PRA Peer Review Report, Boiling Water Reactor Owners Group (BWROG), April 2010.

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Risk Impact Assessment of Extending the CPS ILRT Interval Appendix A - PRA Technical Adequacy APPENDIX A PRA TECHNICAL ADEQUACY CL-LAR-012 Rev. 2

Risk Impact Assessment of Extending the CPS ILRT Interval Appendix A - PRA Technical Adequacy A.1 PRA TECHNICAL ADEQUACY INTRODUCTION This appendix provides information on the technical adequacy of the Clinton Power Station (CPS) Probabilistic Risk Assessment (PRA) Full Power Internal Events (FPIE) model (including internal flooding) and the Fire PRA model.

Exelon employs a multi-faceted approach to establishing and maintaining the technical adequacy and fidelity of PRA models for all operating Exelon nuclear generation sites.

This approach includes both a proceduralized PRA maintenance and update process and the use of self-assessments and independent peer reviews.

Section A.2 describes requirements related to the scope of the CPS PRA models. Section A.3 addresses the technical adequacy of the FPIE PRA for this application. Section A.4 similarly addresses the technical adequacy of the Fire PRA for this application.

All PRA models described below have been peer reviewed, and the review and closure of F&Os from the peer review have been independently evaluated to confirm that the associated model changes did not constitute a model upgrade. Currently, there are two (2) open FPIE F&Os that are linked to Supporting Requirements (SRs) assessed as Not Met or Capability Category (CC) I. There are no open Fire F&Os (i.e., all F&Os linked to SRs assessed as Not Met or CC I have been resolved as confirmed by the independent review team). The resolved F&Os and the basis for resolution are documented in the F&O Closure Review reports [39, 41].

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Risk Impact Assessment of Extending the CPS ILRT Interval Appendix A - PRA Technical Adequacy A.2 REQUIREMENTS RELATED TO SCOPE OF CPS PRA MODELS The PRA models discussed in this appendix have been assessed against RG 1.200 [42].

Both the CPS Internal Events PRA model (including internal flooding) and the CPS Fire PRA model are at-power models. The models are capable of quantifying Core Damage Frequency (CDF) and Large Early Release Frequency (LERF). Note that this portion of the CPS PRA model does not incorporate the risk impacts of external events. The treatment of seismic risk and other external hazards for this application are discussed in Section A.5.

A.3 SCOPE AND TECHNICAL ADEQUACY OF CPS INTERNAL EVENTS AND INTERNAL FLOODING PRA MODEL The information provided in this section demonstrates that the CPS FPIE PRA model (including internal flood) meets the expectations for PRA scope and technical adequacy as presented in RG 1.200. The CPS PRA model for internal events received a formal industry peer review in October 2009 [47]. The CPS (including internal flooding) Peer Review was performed using the NEI 05-04 process [44], the ASME/ANS PRA Standard and Regulatory Guide 1.200. The Peer Review found that 78.5% of the SRs evaluated met Capability Category II or better. There were fifty-six (56) SRs that were assessed as "Not Met" and twelve (12) SRs that were assessed as meeting only Capability Category I. Of the 68 SRs which were assessed as not meeting Capability Category II or better, seven (7) were related to Internal Flooding SRs. Several of the F&Os associated with the open SRs were related to documentation issues.

The 2009 FPIE Peer Review F&Os were addressed during several periodic PRA updates and the resolutions to the F&Os were reviewed by independent review teams in two separate F&O Closures (in December 2018 and November 2019) that included FPIE &

Fire PRA F&Os [39, 41]. The independent review teams concluded that for the FPIE PRA, one F&O was dispositioned as "partially resolved" and one F&O was dispositioned as "open." All other F&Os representing a gap to meeting CC II for all SRs were dispositioned as "resolved."

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Risk Impact Assessment of Extending the CPS ILRT Interval Appendix A - PRA Technical Adequacy The FPIE PRA Peer Review identified FPIE F&Os associated with SRs assessed as less than CC II. Table A-1 summarizes those F&Os that remain "open" (including those that may be only "partially resolved") at the time of this report. The F&Os discussed in Table A-1 represent the gaps to meeting Capability Category II for the FPIE PRA model.

As documented in Table A-1, only two FPIE F&Os remain open. An assessment with respect to the impact on this application is also provided for each open F&O. The ILRT extension evaluation only requires an assessment of CC-I [1]. Based on the assessments provided in Table A-1, it is concluded the CPS Internal Events PRA (including internal flooding) is of adequate technical capability to support this ILRT extension application.

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Risk Impact Assessment of Extending the CPS ILRT Interval Appendix A - PRA Technical Adequacy TABLE A-1 CLINTON FPIE / INTERNAL FLOODING PRA PEER REVIEW OPEN FACTS AND OBSERVATIONS (POST F&O CLOSURE)

F&O POSSIBLE STATUS DISPOSITION FROM F&O CLOSURE REVIEW MAINTENANCE IMPACT TO ILRT ORIGINATING BASIS FOR ID SR(S) F&O DETAILS SIGNIFICANCE RESOLUTION VS. UPGRADE EXTENSION APPLICATION 1-32 LE-E1 CPS-PSA-004 Section 5.2 discusses An inadequate Solve with all Partially Clinton Assessment: Clinton This issue has minimal impact on the ILRT QU-C1 the use of screening values used for process to post-initiator Resolved Section 5.3 and Appendix K of the Human Reliability Analysis Assessment: extension application since all risk-significant HRA HEPs in order to identify cutsets with identify HEPs set to 1.0 Notebook (CL-PRA-004) summarizes HRA Dependency Analysis Maintenance: dependencies are captured through the current QU-C2 dependent HEPs. However, only twelve combinations of and identify all methodology and results. For CDF and LERF, the FPIE model was Methodology and methodology and results.

HR-H3 of the over 100 basic events modeling operator actions combinations of tools consistent quantified with all post-initiator HEPs set to 0.1 or higher at the A review of the CDF & LERF cutsets was performed LE-E4 post-initiator operator actions are listed can result in operator action- truncation levels of 5E-9/yr (CDF) and 5E-10/yr (LERF). These with previous to determine if any HRA dependent combinations QU-A5 in Table 5.2-1 as using screening significantly related HEPs. truncation levels were selected because they capture all risk- PRA updates. exist without escalated dependent joint HEPs (i.e.,

values to identify dependency. Of underestimating Perform significant post-initiator operator actions. they assume zero dependence and thus the HEPs HR-G7 these, six use a value of 1.0E-02 and CDF and LERF. dependency are unaltered). Separately, a review of the Using the HRA Calculator Dependency Module, all dependent Independent one uses a value of 1.0E-03. The analyses for all combinations concluded that a majority of the combinations were reviewed for proper dependency levels and Review Team remaining five use a value of 0.1. It combinations. unanalyzed dependent combinations are related to order. Once reviewed, a floor value of 1E-06 or 5E-07 may be Assessment:

appears that all other HEPs are time-phased actions (i.e., early vs. late) where no imposed on the dependent joint HEP depending on the timing of Since no new quantified with their nominal values. additional dependency need be assigned between the operator actions. The final FPIE model quantification uses the methods were Use of such low probability values is the actions because the time-phased calculations 0.1 or higher seed values for all post-initiator HEPs and the applied and likely to result in combinations of already reflect the impacts of those dependencies.

adjusted dependent joint HEP is recovered using a post-processing existing methods dependent HEPs being omitted by A few legitimate dependent combinations were recovery file. were not applied truncation values. Use of a sufficiently identified upon further review; however, increasing high value for HEPs is required by SR Independent Review Team Assessment: in a different the dependent joint HEPs for these groups does not QU-C1 and not using a sufficiently high A check of the CAFTA RR Database indicates that the post-initiator context, this substantially impact the overall risk results.

value would result in an inadequate HEPs were set to 0.1 (or greater) prior to dependency analysis. constitutes model maintenance. Further justification for the chosen truncation level assessment of dependent HEPs. The value of 0.1 can be acceptable depending upon what used in the HRA Dependency Analysis is required (This F&O originated from SR HR-G7) truncation level is used for the dependency analysis and whether in a future model update.

all multiple independent HFEs are recovered by combination HFEs Therefore, this open item is primarily a and Joint HEPs. The resolution of this Finding is correlated to documentation issue.

Finding 1-34.

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Risk Impact Assessment of Extending the CPS ILRT Interval Appendix A - PRA Technical Adequacy TABLE A-1 CLINTON FPIE / INTERNAL FLOODING PRA PEER REVIEW OPEN FACTS AND OBSERVATIONS (POST F&O CLOSURE)

F&O POSSIBLE STATUS DISPOSITION FROM F&O CLOSURE REVIEW MAINTENANCE IMPACT TO ILRT ORIGINATING BASIS FOR ID SR(S) F&O DETAILS SIGNIFICANCE RESOLUTION VS. UPGRADE EXTENSION APPLICATION 1-34 LE-E1 Solving the PRA models with some The solution Solve the PRA Open Clinton Assessment: Clinton See discussion for F&O 1-32.

QU-C1 HEPs at nominal can result in cutsets method used model with See discussion for F&O 1-32. Assessment:

with multiple operator actions being likely under operator action Maintenance:

QU-C2 Independent Review Team Assessment:

truncated out or with the combined predicts the risk failure Methodology and HR-H3 probability of all operator actions much values. This probability The CL-PRA-004 Rev. 6 document was reviewed. The final model tools consistent LE-E4 below the 1E-6 or 5E-7 floor that the under prediction values set to a cutsets were re-imported into the existing HRA DAF files (for FPIE with previous HRA notebook says is used. The peer could be high value. CDF only), using a copy of the HRAC database with all 1.0 HEPs QU-A5 PRA updates.

review team quantified the PRA model significant removed and the inhibit ADS also removed per the analyst notes HR-G7 with post-initiator HEPs set to 0.1 and based on the for that HFE. This process was used to determine if there are identified a significant number of total number of combinations of HFEs occurring in the final results with all HEPs Independent cutsets containing combinations of operator actions set to nominal values and no combination event applied. 318 new Review Team basic events representing operator included in the combinations were identified (in addition to the 216 that were Assessment:

action failure. These combinations were CPS model. originally identified and implemented), several of which had FV Maintenance -

reviewed and a large number of values above 5E-03 as calculated by the HRAC (which is not a true modeling error, combinations identified in this review risk metric but a good approximation). approach will not were not included in the CPS HRA For example, 1FWOPFLWCTRL-H-- and 1FWOPMANINIT-H-- change.

dependency evaluation. appear as a combination together and have a dependency level of (This F&O originated from SR HR-G7) HD, confirmed in the HRA Calculator via override notes, however when this pair of HFEs appears together it is not recovered with a combination event. This combination has a an FV value of 2.9E-01 as calculated by the HRAC (again, not a true risk metric but a good approximation). This suggests it is likely risk significant when dependencies are accounted for, and additional unanalyzed combinations may also be present when dependencies are accounted for.

The review teams concern is that potentially risk significant combinations of HFEs are not captured through the current approach, due to the chosen truncation level for the dependency identification (5E-9 / 5E-10 for CDF/LERF) in conjunction with the elevated HEP level chosen (0.1). This could under predict risk results as stated in the original F&O, and is supported by the observations noted above. It is noted that the example combination above did appear in the 1E-9 / 5E-11 identification cutsets that were included in the dependency files, but not used.

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Risk Impact Assessment of Extending the CPS ILRT Interval Appendix A - PRA Technical Adequacy A.4 SCOPE AND TECHNICAL ADEQUACY OF CPS FIRE PRA MODEL The CPS Fire PRA (FPRA) Peer Review [45] was performed in April 2018 using the NEI 07-12 Fire PRA peer review process [43], the ASME/ANS PRA Standard, ASME/ANS RA-Sa-2009 [9] and Regulatory Guide 1.200 [42]. The purpose of this review was to establish the technical adequacy of the FPRA for the spectrum of potential risk-informed plant licensing applications for which the FPRA may be used.

The 2018 CPS FPRA Peer Review was a full-scope review of the CPS at-power FPRA against all technical elements in Part 4 of the ASME/ANS PRA Standard, including the referenced internal events Supporting Requirements (SRs). The Peer Review found that 96.9% of the SRs evaluated met Capability Category (CC) II or better. There were five (5)

SRs that were assessed as "Not Met" and eight (8) SRs that were assessed as meeting only CC I. Many of the F&Os, leading to open SRs, were related to documentation issues.

The 2018 FPRA Peer Review F&Os were addressed in subsequent FPRA updates and the resolutions to the F&Os were reviewed by independent review teams in two separate F&O Closures (in December 2018 and November 2019) that included FPIE & FPRA F&Os [39, 41]. The independent review teams concluded that for the FPRA, all F&Os have been dispositioned as "resolved." Therefore, there are no open F&Os to discuss for this application.

Given the resolution of all F&Os related to SRs assessed with less than a CC II, it is concluded that the CPS FPRA is of adequate technical capability to support the ILRT extension application.

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Risk Impact Assessment of Extending the CPS ILRT Interval Appendix A - PRA Technical Adequacy A.5 SCOPE AND TECHNICAL ADEQUACY OF CPS SEISMIC CDF AND LERF ESTIMATES Estimates of Seismic CDF and Seismic LERF we derived for use in the CPS TSTF-505 program [38]. The Seismic CDF and Seismic LERF derived for use in the TSTF-505 program reflect best estimates using available information and potentially conservative assumptions; the methodology and technical adequacy are detailed in the CPS TSTF-505 LAR [46]. The technical adequacy of the Seismic CDF and Seismic LERF are qualitatively found to be adequate to support the conclusions found in Section 6.0 of this document.

A.6

SUMMARY

A PRA technical adequacy evaluation was performed consistent with the requirements of RG 1.200. This evaluation combined with the details of the results of this analysis demonstrate with reasonable assurance that the proposed one-time extension to the ILRT interval for CPS Unit 1 to 15.7 years satisfies the risk acceptance guidelines in RG 1.174 and that the associated PRA models are technically adequate.

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From: Kelly, Eugene M:(Exelon Nuclear) <eugene.kelly@exeloncorp.com>

Sent: Friday, February 19, 2021 9:34 AM To: Brinsfield, Wes <WBrinsfield@jensenhughes.com>

Cc: Nicely, Ken M.:(Exelon Nuclear) <ken.nicely@exeloncorp.com>

Subject:

RE: CL-LAR-012 Revision 2 Signatures for Approval By copy of this email, I approve the Rev.2 version Gene Kelly Senior Manager Exelon Nuclear Risk Management (267) 533-1467 Office (215) 767-4740 Mobile From: Sattler, Justin <JSattler@jensenhughes.com>

Sent: Friday, February 19, 2021 9:45 AM To: Standridge, Charles <CStandridge@jensenhughes.com>; Brinsfield, Wes

<WBrinsfield@jensenhughes.com>; Johnson, Matt <MJohnson@jensenhughes.com>;

Warren, Vicki <VWarren@jensenhughes.com>

Subject:

RE: CL-LAR-012 Revision 2 Signatures for Approval I, Justin Sattler, Jensen Hughes PRA Engineer, approve CL-LAR-012 R2 as the PRA Preparer.

Justin Sattler, PE Staff Engineer 1 1500 McConnor Parkway, Suite 500, Schaumburg, IL 60173 O: +1 630-491-8155 l C: +1 262-370-7728 jsattler@jensenhughes.com jensenhughes.com From: Warren, Vicki <VWarren@jensenhughes.com>

Sent: Friday, February 19, 2021 9:05 AM To: Standridge, Charles <CStandridge@jensenhughes.com>; Brinsfield, Wes

<WBrinsfield@jensenhughes.com>; Johnson, Matt <MJohnson@jensenhughes.com>;

Sattler, Justin <JSattler@jensenhughes.com>

Subject:

CL-LAR-012 Revision 2 Signatures for Approval I, Victoria Warren, Fleet risk Management Engineer, approve CL-LAR-012 R2 as the PRA Reviewer.

Victoria Warren VICTORIA WARREN CL-LAR-012 Rev. 2

Senior Engineer 1 158 West Gay Street, Suite 400 West Chester, PA 19380 O: +1 484 260 4730 l C: +1 484 716 2104 vwarren@jensenhughes.com jensenhughes.com From: Standridge, Charles <CStandridge@jensenhughes.com>

Sent: Friday, February 19, 2021 8:37 AM To: Brinsfield, Wes <WBrinsfield@jensenhughes.com>; Johnson, Matt

<MJohnson@jensenhughes.com>; Sattler, Justin <JSattler@jensenhughes.com>;

Warren, Vicki <VWarren@jensenhughes.com>

Subject:

RE: CL-LAR-012 Revision 2 Signatures for Approval I, Charles Standridge, Jensen Hughes PRA Engineer, approve CL-LAR-012 R2 as the PRA Reviewer.

CHARLES STANDRIDGE Manager, Risk & Reliability C: +1 408-892-5876 cstandridge@jensenhughes.com jensenhughes.com How were serving our clients and partners during the COVID-19 pandemic:

Jensen Hughes continues to operate around the world, while following all local and federal guidance.

Please know that we are taking extensive and unprecedented precautions to keep our employees healthy, protected and productive to support you and our ongoing projects.

From: Johnson, Matt <MJohnson@jensenhughes.com>

Sent: Friday, February 19, 2021 8:35 AM To: Brinsfield, Wes <WBrinsfield@jensenhughes.com>; Sattler, Justin

<JSattler@jensenhughes.com>; Warren, Vicki <VWarren@jensenhughes.com>;

Standridge, Charles <CStandridge@jensenhughes.com>

Subject:

CL-LAR-012 Revision 2 Signatures for Approval I, Matt Johnson, Jensen Hughes Lead PRA Engineer, approve CL-LAR-012 R2 as the PRA Reviewer.

MATT JOHNSON, PE Lead Engineer C: +1 949-315-0407 mjohnson@jensenhughes.com jensenhughes.com CL-LAR-012 Rev. 2