RS-16-015, 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

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

ML16025A182
Person / Time
Site:	Clinton
Issue date:	01/25/2016
From:	Simpson P Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-16-015
Download: ML16025A182 (191)

Text

300 W'nl1Plcf ,c.1<1 l/v.11 1~1!1" IL 60'>" 'i Exelon Generation ~0 7 tOOO 011 ce RS-16-015 10 CFR 50.90 January 25, 2016 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

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 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 is a request to revise TS 5.5.13, "Primary Containment Leakage Rate Testing Program" to allow for the permanent extension of the Type A Integrated Leak Rate Testing (ILRT) and Type C Leak Rate Testing frequencies.

Specifically, the proposed change will revise CPS TS 5.5.13, by replacing the references to Regulatory Guide (AG) 1.163, "Performance-Based Containment Leak-Test Program," and 10 CFR 50, Appendix J, Option B with a reference to NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, as the documents used by CPS to implement the performance-based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J. This license amendment request (LAA) also proposes an administrative change to TS 5.5.13 to delete the information regarding the performance of the next CPS Type A test to be performed no later than November 2008 as this Type A test has already occurred.

Attachment 1 contains the evaluation of the proposed changes. Attachment 2 provides the marked up TS pages. The marked up TS Bases page is provided in Attachment 3 for information only.

The proposed amendment is risk-informed and follows the guidance in Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis," Revision 2. EGC has performed a CPS-specific evaluation to assess the risk impact of the proposed amendment. A copy of the risk assessment is provided in Attachment 4.

January 25, 2016 U. S. Nuclear Regulatory Commission Page2 The proposed change has been reviewed by the CPS Plant Operations Review Committee, and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed amendment by January 31, 2017 in order to support the extension of the CPS Unit 1 ILRT, which is required to be performed during the outage in the Spring of 2017. Once approved, this amendment shall be implemented within 30 days.

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.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Timothy Byam at (630) 657-2818.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 251h day of January 2016.

t Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC Attachments:

1) Evaluation of Proposed Change

2) Markup of Technical Specifications Page

3) Markup of Technical Specifications Bases Page

4) Risk Assessment for CPS Regarding the ILRT (Type A) Permanent Extension Request cc: NRC Regional Administrator, Region Ill 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 - Revise Technical Specification Section 5.5.13 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies 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 ECCS Net Positive Suction Head (NPSH) Analysis 3.4 Justification for the Technical Specification Change 3.5 Plant Specific Confirmatory Analysis 3.6 Non-Risk Based Assessment 3.7 Operating Experience 3.8 NRC SE Limitations and Conditions 3.9 Conclusion

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

Page 1 of 52

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 NPF-62 for Clinton Power Station (CPS) Unit 1. The proposed change is a request to revise TS 5.5.13, "Primary Containment Leakage Rate Testing Program" to allow the following:

Increase in the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with Nuclear Energy Institute (NEI) Technical Report NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A and the conditions and limitations specified in NEI 94-01, Revision 2-A.

Adopt an extension of the containment isolation valve (CIV) leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,"

Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.

Adopt the use of ANSI/ANS 56.8-2002, "Containment System Leakage Testing Requirements."

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

Specifically, the proposed change contained herein would revise CPS TS 5.5.13, by replacing the references to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," (Reference 1) and 10 CFR 50, Appendix J, Option B with a reference to NEI 94-01, Revision 3-A (Reference 2), and the conditions and limitations specified in NEI 94-01, Revision 2-A (Reference 8), as the documents used by CPS to implement the performance-based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J. This license amendment request (LAR) also proposes an administrative change to TS 5.5.13 to delete the information regarding the performance of the next CPS Type A test to be performed no later than November 2008 as this Type A test has already occurred.

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 Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exceptions:

(1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Page 2 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Type A tests, and (2) NEI 94-01 1995, Section 9.2.3: The first Type A test performed after November 23, 1993 shall be performed no later than November 23, 2008.

The proposed changes to CPS TS 5.5.13 will replace the reference to RG 1.163 with a reference to NEI Topical Report NEI 94-01 Revisions 2-A and 3-A.

The proposed change will revise TS 5.5.13 to state, 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 modified by the following exception: (1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Type A tests."

Markup of TS 5.5.13 is provided in Attachment 2.

Markup of TS Bases 3.6.1.1 is provided in Attachment 3 for information only. contains the plant specific risk assessment conducted to support this proposed change. This risk assessment followed the guidelines of NRC RG 1.174, Revision 2 (Reference

3) and NRC RG 1.200, Revision 2 (Reference 4). The risk assessment concluded that increasing the ILRT on a permanent basis to one-in-fifteen year frequency is considered to represent a small change in the CPS risk profile.

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 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 Page 3 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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 37) 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 38).

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.

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.

Page 4 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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 Technical Specifications.

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:

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 inch (max.).

Page 5 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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 drywell is not normally entered during operation, but access is possible during a hot shutdown with the reactor subcritical.

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.

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.

Page 6 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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 200oF 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.

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.

Page 7 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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 ECCS Net Positive Suction Head (NPSH) Analysis NPSH available to the ECCS pumps has been determined in accordance with RG 1.1, "Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps." Pressure drop across the suction strainer is based on results from testing and conservative analysis. The vapor pressure for suppression pool water used in NPSH calculations for events where significant debris generation is expected is based on a suppression pool bulk water temperature of 185oF, which is the maximum design temperature of the containment. Analyses show maximum suppression pool temperatures to be less than the containment design temperature of 185oF. For events in which no significant debris generation is expected, NPSH will continue to be evaluated for 212oF suppression pool water temperature.

Containment pressure is assumed to be atmospheric in accordance with RG 1.1 requirements.

3.4 Justification for the Technical Specification Change 3.4.1 Chronology of Testing Requirements of 10 CFR 50, Appendix J The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the TS. Title 10 CFR 50, 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.

Page 8 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50, Appendix J refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.

Also in 1995, RG 1.163 (Reference 1) was issued. The RG endorsed NEI 94-01, Revision 0, (Reference 5) with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) test from three tests in 10 years to one test in 10 years. This relaxation was based on an NRC risk assessment contained in NUREG-1493, (Reference 6) and Electric Power Research Institute (EPRI) TR-104285 (Reference 7) both of which showed that the risk increase associated with extending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval, provisions for extending the test interval an additional 15 months 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 8), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC Safety Evaluation (SE) 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 1). 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.

In 2012, NEI 94-01, Revision 3-A (Reference 2), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J and includes provisions for extending Type A ILRT intervals to up to 15 years. NEI 94-01 has been endorsed by RG 1.163 and NRC SEs of June 25, 2008 (Reference 9) and June 8, 2012 (Reference 10) as an acceptable methodology for complying with the provisions of Option B in 10 CFR 50, Appendix J. The regulatory positions stated in RG 1.163 as modified by References 9 and 10 are incorporated in 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. Extensions of Type B and Type C test intervals are allowed based upon completion of two consecutive periodic as-found tests where the results of each test are within a licensees allowable administrative limits. Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment airlocks) and up to a maximum of 75 months for Type C tests. If a licensee considers extended test intervals of greater than 60 months for Type B or Type C tested components, the review should include the additional considerations of as-found tests, schedule and review as described in NEI 94-01, Revision 3-A, Section 11.3.2.

The NRC has provided 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:

Page 9 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Section 9.2.3, NEI TR 94-01, Revision 2, states, "Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 15 years based on acceptable performance history." However, Section 9.1 states that the "required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes." The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

NEI 94-01, Revision 3-A, Section 10.1 concerning the use of test interval extensions in the deferral of Type B and Type C LLRTs past intervals of up to 120 months for the recommended surveillance frequency for Type B testing and up to 75 months for Type C testing, states:

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

Notes: For routine scheduling of tests at intervals over 60 months, refer to the additional requirements of Section 11.3.2.

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

The NRC 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.

3.4.2 Current CPS ILRT Requirements 10 CFR 50, Appendix J was revised, effective October 26, 1995, to allow licenses to choose containment leakage testing under either Option A, "Prescriptive Requirements," or Option B, "Performance-Based Requirements." On June 21, 1996 the NRC approved Amendment 105 for CPS (Reference 13) 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:

Page 10 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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.

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.

Current TS 5.5.13 requires that a program be established to comply with the containment leakage rate testing requirements of 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The program is required to be in accordance with the guidelines contained in RG 1.163. RG 1.163 endorses, with certain exceptions, NEI 94-01, Revision 0, as an acceptable method for complying with the provisions of Appendix J, Option B.

RG 1.163, Section C.1 states that licensees intending to comply with 10 CFR 50, Appendix J, Option B, should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01 (Reference 5) rather than using test intervals specified in ANSI/ANS 56.8-1994. NEI 94-01, Section 11.0 refers to Section 9, which states that Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per ten years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage was less than 1.0La (where La is the maximum allowable leakage rate at design pressure). Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine performance shall be at least 24 months.

Adoption of the Option B performance based containment leakage rate testing program altered the frequency of measuring primary containment leakage in Types A, B, and C tests but did not alter the basic method by which Appendix J leakage testing is performed. The test frequency is based on an evaluation of the "as found" leakage history to determine a frequency for leakage testing which provides assurance that leakage limits will not be exceeded. The allowed frequency for Type A testing as documented in NEI 94-01 is based, in part, upon a generic evaluation documented in NUREG-1493. The evaluation documented in NUREG-1493 included a study of the dependence or reactor accident risks on containment leak tightness for differing types of containment types, including a post tensioned, shallow domed concrete containment similar to CPSs containment structures. NUREG-1493 concluded in Section 10.1.2 that reducing the frequency of Type A tests from the original three (3) tests per 10 years to one (1) test per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Types B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements. Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A Page 11 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE testing, NUREG-1493 concluded that increasing the interval between ILRTs is possible with minimal impact on public risk.

3.4.3 CPS 10 CFR 50, Appendix J, Option B Licensing History June 21, 1996 The NRC issued Amendment 105 on June 21, 1996 (Reference 13). The amendment revised the Operating License and TS to implement 10 CFR 50, Appendix J - Option B, by referring to Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program." Specifically, changes were made to paragraph 2.D of the Operating License; TS Section 1.1, "Definitions;"

TS 3.6.1.1, "Primary Containment;" TS 3.6.1.2, "Primary Containment Air Locks;" TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs);" and TS Section 5.5, "Programs and Manuals."

September 4, 1996 The NRC issued Amendment 106 on September 4, 1996 (Reference 14). The amendment revised Technical Specifications for the drywell to permit bypass testing on a 10-year frequency with increased testing if performance degrades, changed the drywell air lock testing and surveillance requirements, deleted action notes for the drywell air lock and drywell isolation valves when the bypass leakage limit is not met, and deleted the specific leakage limits for the drywell air lock seal.

March 8, 1999 The NRC issued Amendment 121 on March 8, 1999 (Reference 18). The amendment allowed the deferral of the next scheduled local leak rate test for penetration 1MC-042 until the seventh refueling outage.

March 26, 2002 The NRC issued Amendment 145 on March 26, 2002 (Reference 19). The amendment replaced individual main steamline leakage limits with an aggregate leakage limit, revising technical specification surveillance requirement 3.6.1.3.9, which provides leakage rate limits applicable to the main steamline isolation valves.

January 8, 2004 The NRC issued Amendment 160 on January 8, 2004 (Reference 15). The amendment proposed a one-time Technical Specification change to extend the test interval for the next Appendix J Type A test and the next drywell bypass leakage rate test from 10 to 15 years.

September 19, 2005 The NRC issued Amendment 167 on September 19, 2005 (Reference 16). This amendment supported the application of an alternative source term (AST) methodology, in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, "Accident Source Term,"

with the exception that Technical Information Document (TID) 14844, "Calculation of Distance Page 12 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Factors for Power and Test Reactor Sites," used as the radiation dose basis for equipment qualification at CPS.

March 21, 2006 The NRC issued Amendment 173 on March 21, 2006 (Reference 17). This amendment revised TS SR 3.6.1.3.8 to exclude the containment purge valve leakage rates from the summation of secondary containment bypass leakage rates.

November 16, 2006 On November 16, 2006, AmerGen requested an amendment (Reference 21) to TS 3.6.5.1, "Drywell" and 5.5.13, "Primary Containment Leakage Rate Testing Program," to delay the performance of the next primary containment Type A ILRT from the current requirement of "no later than November 23, 2008" to "prior to startup from the C1R12 refueling outage."

April 30, 2007 On April 30, 2007, AmerGen withdrew the Request for Amendment (Reference 22) to TS 3.6.5.1, "Drywell" and 5.5.13, "Primary Containment Leakage Rate Testing Program," dated November 16, 2006.

3.4.4 Integrated Leakage Rate Testing History As noted previously, CPS TS 5.5.13 currently requires Type A, B, and C testing in accordance with RG 1.163, which endorses the methodology for complying with Option B. Since the adoption of Option B, the performance leakage rates are calculated in accordance with NEI 94-01, Section 9.1.1 for Type A testing. Table 3.4.4-1 lists the past Periodic Type A ILRT results for CPS.

Table 3.4.4-1, 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 13 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE (1) Subsequent to the ILRT, a hole was discovered through the containment liner plate. The hole was identified on NCR-40031 and CR 1-86-01-119. This hole was evidently in existence during the ILRT and was created by the removal of temporary Attachment No.

CL-J-12-4. After repair the hole was retested utilizing a leak chase channel and bubbler per procedure XTP-00-07. The retest showed zero leakage. ILRT was re-performed in November 1986.

The results of the last two Type A ILRTs for CPS were less than the maximum allowable containment leakage rate of 0.65 weight%/day. As a result, since both tests were successful, CPS has been placed on an extended ILRT frequency. The current ILRT interval frequency for CPS is 10 years.

3.4.5 Drywell Bypass Leakage Rate Test (DBLRT) History The leaktightness of the drywell is periodically verified by performance of the DBLRT. This test ensures that the measured drywell bypass leakage is bounded by the safety analysis assumptions. The drywell integrity is further verified by a number of additional tests, including drywell airlock door seal leakage tests, overall drywell airlock leakage tests, drywell isolation valve tests and periodic visual inspections of exposed accessible interior and exterior drywell surfaces. Additional confidence that significant degradation in the drywell integrity has not developed is provided by the periodic qualitative assessment of drywell performance. This assessment was credited in the NRC's acceptance of the current performance-based surveillance frequency of 120 months, approved with TS Amendment 106 for CPS (Reference 14).

The DBLRT Surveillance Frequency is controlled under the Surveillance Frequency Control Program (SFCP). The scheduling of TS SR 3.6.5.1.3 in accordance with the SFCP was approved as part of TS Amendment 192 (Reference 36). As defined in CPS TS 5.5.16, "Surveillance Frequency Control Program," changes to the DBLRT frequency listed in the SCFP shall be made in accordance with NEI 04-10, "Risk-Informed Method for control of Surveillance Frequencies," Revision 1. As such, any changes to the DBLRT frequency do not require NRC review and approval. However, the DBLRT has been historically associated with the ILRT frequency because the plant line-ups are similar, the same equipment is used to perform both tests, and EGC intends to extend the frequency associated with the DBLRT as well in accordance with the SFCP. Therefore, in support of this assumption the risk assessment presented in Attachment 4 of this submittal also includes an assessment for extending the DBLRT interval to once in 15 years.

3.5 Plant Specific Confirmatory Analysis 3.5.1 Methodology An evaluation has been performed to provide an assessment of the risk associated with implementing a permanent extension of the CPS containment Type A ILRT interval from ten years to fifteen years. The risk assessment follows the guidelines from NEI 94-01 (Reference 2), the methodology outlined in EPRI TR-104285 (Reference 7), as updated by the EPRI Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (EPRI TR-1018243)

(Reference 11), the NRC regulatory guidance on the use of Probabilistic Risk Assessment Page 14 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE (PRA) findings and risk insights in support of a request for a plants licensing basis as outlined in RG 1.174 (Reference 3), 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. The format of this document is consistent with the intent of the Risk Impact Assessment Template for evaluating extended ILRT intervals provided in the EPRI TR-1018243 (Reference 11).

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 percent per day to 5 percent per day leads to a barely perceptible increase in total population exposure, and increasing the leak rate to 50 percent per day increases the total population exposure by less than 1 percent. 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 methodology to perform the risk assessment. In October 2008, EPRI TR-1018243 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 NRC RG 1.174. 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/DWBT interval extension risk assessment for CPS employs the EPRI 1018243 methodology, with the affected System, Structure, or Component (SSC) being the primary containment boundary.

In the SE issued by NRC letter dated June 25, 2008 (Reference 9), the NRC concluded that the methodology in EPRI TR-1009325, Revision 2, was acceptable for referencing by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.0 of the SE. Table 3.5.1-1 addresses each of the four limitations and conditions for the use of EPRI TR-1009325, Revision 2.

Table 3.5.1-1, EPRI Report No. 1009325 Revision 2 Limitations and Conditions Limitation/Condition (From Section 4.2 of SE) CPS Response

1. The licensee submits documentation CPS PRA technical adequacy is addressed in indicating that the technical adequacy of Section 3.5.2 of this LAR and Attachment 4, their PRA is consistent with the "Risk Impact Assessment of Extending the requirements of RG 1.200 relevant to the CPS ILRT/DWBT Interval," Appendix A, "PRA ILRT extension. Technical Adequacy."

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Attachment 1 EVALUATION OF PROPOSED CHANGE Table 3.5.1-1, EPRI Report No. 1009325 Revision 2 Limitations and Conditions Limitation/Condition (From Section 4.2 of SE) CPS Response 2.a The licensee submits documentation Because the ILRT does not impact CDF, the indicating that the estimated risk increase relevant criterion is LERF. The increase in associated with permanently extending the internal events LERF resulting from a change ILRT surveillance interval to 15 years is in the Type A ILRT interval for the base case small, and consistent with the clarification with corrosion included is 9.81E-09/yr, which provided in Section 3.2.4.5 of this SE. falls within the "very small" change region of the acceptance guidelines in RG 1.174.

If the EPRI Expert Elicitation methodology Class 3a and Class 3b failure probabilities are used, the change is estimated as 1.15E-09/yr, which falls further within the very small change region of the acceptance guidelines in RG 1.174.

2.b Specifically, a small increase in The change in dose risk for changing the population dose should be defined as an Type A ILRT interval from three-per-ten years increase in population dose of less than or to once-per-fifteen-years, measured as an equal to either 1.0 person-rem per year or increase to the total integrated dose risk for 1% of the total population dose, whichever all accident sequences, is 3.80E-03 person-is less restrictive. rem/yr using the EPRI guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr.

The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology. The change in dose risk meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1%

person-rem/yr.

2.c In addition, a small increase in CCFP The increase in CCFP from the three in ten should be defined as a value marginally year interval to one in fifteen years including greater than that accepted in a previous corrosion effects using the EPRI guidance is one-time 15-year ILRT extension requests. 0.44%, which is below the acceptance criteria This would require that the increase in of 1.5%. The increase in CCFP drops to CCFP be less than or equal to 1.5 about 0.05% using the EPRI Expert Elicitation percentage point. methodology.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Table 3.5.1-1, EPRI Report No. 1009325 Revision 2 Limitations and Conditions Limitation/Condition (From Section 4.2 of SE) CPS Response

3. The methodology in EPRI Report No. The representative containment leakage for 1009325, Revision 2, is acceptable except Class 3b sequences used by CPS is 100 La, for the calculation of the increase in based on the recommendations in the latest expected population dose (per year of EPRI report (Reference 20) and as reactor operation). In order to make the recommended in the NRC SE on this topic methodology acceptable, the average leak (Reference 9). It should be noted that this is rate accident case (accident case 3b) more conservative than the earlier previous used by the licensees shall be 100 La industry ILRT extension requests, which instead of 35 La. utilized 35 La for the Class 3b sequences.

4. A licensee amendment request (LAR) is For CPS containment over-pressure is not required in instances where containment relied upon for ECCS performance.

over-pressure is relied upon for ECCS Reference Section 3.3 of this Attachment for performance. details.

3.5.2 Technical Adequacy of the PRA The PRA Technical Adequacy Evaluation is presented in Appendix A, "PRA Technical Adequacy" of Attachment 4 of this submittal. The following is a summary of that evaluation.

3.5.2.1 Demonstrate the Technical Adequacy of the PRA The guidance provided in RG 1.200, Section 4.2 "License Submittal Documentation," indicates that the following items be addressed in documentation submitted to the NRC to demonstrate the technical adequacy of the PRA:

Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the application.

Document peer review findings and observations that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.

Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the RG. Provide justification to show that where specific requirements in the standard are not met, it will not unduly impact the results.

Identify key assumptions and approximations relevant to the results used in the decision-making process.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 3.5.2.2 Technical Adequacy of the PRA Model The risk assessment performed for the ILRT extension request is based on the current Level 1 and Level 2 PRA model. Note that for this application, the accepted methodology involves a bounding approach to estimate the change in the PRA risk metric of LERF from extending the ILRT interval. Rather than exercising the PRA model itself, it involves the establishment of separate evaluations that are linearly related to the plant Core Damage Frequency (CDF) contribution. Consequently, a reasonable representation of the plant CDF that does not result in a LERF does not require that Capability Category II be met in every aspect of the modeling if the Category I treatment is conservative or otherwise does not significantly impact the results.

3.5.2.3 PRA Model Evolution and Peer Review Summary The 2014A version of the CPS PRA model is the most recent evaluation of the Unit 1 risk profile at CPS for internal event challenges. The CPS PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events.

The PRA model quantification process used for the CPS PRA is based on the event tree/fault tree methodology, which is a well-known methodology in the industry.

EGC employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating EGC nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the CPS PRA.

PRA Maintenance and Update The EGC risk management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plant. This process is defined in the Exelon Risk Management program, which consists of a governing procedure and subordinate implementation procedures. The PRA model update procedure delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating EGC nuclear generation sites. The overall Exelon Risk Management program defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.), and for controlling the model and associated computer files.

Plant Changes Not Yet Incorporated into the PRA Model A PRA updating requirements evaluation (URE) documented in the CPS PRA model update tracking database is created for all issues that are identified that could impact the PRA model.

The URE database includes the identification of those plant changes that could impact the PRA model.

A review of the open UREs indicates that there are no plant changes that have not yet been incorporated into the PRA model that would affect this application. UREs are evaluated for potential impact to applications and to the PRA base model results and are classified as High, Medium or Low priority. High priority items could significantly impact applications. Medium priority items are items that are assessed as potentially important to applications and Low Page 18 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE priority items area items that are assessed as not important to applications and likely to have minimal or no numeric impact. There are no open High priority UREs and seventeen UREs identified as Medium priority. The remaining open UREs are low priority, having little or no impact to the PRA results. Medium priority UREs were found not to impact this application.

Low priority UREs were also reviewed and none were found that would impact this application.

Consistency with Applicable PRA Standards The CPS Full Power Internal Events (FPIE) PRA model has undergone several reviews, including a BWROG Peer Review in 2000 (References 12, 23 and 39), UREs were created, and extensive changes to the PRA model were made in updates through 2006. As a result of the extensive changes made to the model, a full Peer review was again performed in 2009 (Reference 31). The results of the 2009 Peer review and the actions taken to address "gaps" identified, best represent the consistency of the model to current PRA standards.

The 2009 Peer Review was conducted using the 2009 ASME/ANS PRA Standard (Reference

30) and the NRCs comments and clarifications contained in RG 1.200, Revision 2. The Peer Review was conducted using the CPS 2006C FPIE PRA model (Reference 35). The general objective for the CPS PRA is to meet Capability Category II. The findings and observations (F&Os) that were identified in the Peer Review were between the 2006C CPS PRA and the requirements for Capability Category II. These F&Os (i.e., both findings and suggestions) were entered into the CPS URE database for tracking purposes. These UREs were used for scoping of the CPS 2011 PRA update.

All "Findings" from the 2009 Peer Review were addressed as part of the 2011 PRA update. The 2009 CPS Peer Review observations were incorporated into the CPS URE database for tracking. All but one of the "Suggestion" observations have been addressed. The observation that remains open recommended addressing environmental conditions for several operator actions credited outside of the Main Control Room. The actions associated with this observation are assigned Human Error Probabilities (HEPs) of 0.9 in the PRA model. It is judged that addressing this observation would have minimal impact to LERF and the ILRT risk assessment.

Following the 2009 Peer review, a self-assessment relative to the combined ASME/ANS PRA Standard (Reference 30) and the NRCs comments and clarifications contained in RG 1.200 was performed as part of the 2011 PRA update (Reference 34). The 2011 Self-Assessment used the 2009 Peer review results as input. Attachment 4 of this submittal, Table A-2 identified gaps to Category II identified in the 2011 self-assessment and the status of those gaps following the 2011 and 2014 PRA updates. Included in the last column of LAR Attachment 4, Appendix A, Table A-2, "Summary of Clinton 2006 PRA Self-Assessment Identified Enhancements," is the URE number, a significance statement and the impact to this ILRT for the gaps that have not been addressed. These gaps are judged to not have an impact on this application as justified in Table A-2.

External Hazards Although EPRI report 1018243 (Reference 20) recommends a quantitative assessment of the contribution of external events (for example, fire and seismic) where a model of sufficient quality exists, it also recognizes that the external events assessment can be taken from existing, Page 19 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE previously submitted and approved analyses or another alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. Based on this, currently available information for external events models was referenced, and a multiplier was applied to the internal events results based on the available external events information. This is further discussed in LAR Attachment 4, Section 5.7. The fire and seismic PRA Technical Adequacy are discussed in additional detail in Attachment 4, Appendix A, Section A.3, "External Hazards."

Identification of Key Assumptions The methodology employed in this risk assessment followed the EPRI guidance (Reference 20) as previously approved by the NRC. The analysis included the incorporation of several sensitivity studies and factored in the potential impacts from external events in a bounding fashion. None of the sensitivity studies or bounding analyses indicated any source of uncertainty or modeling assumption that would have resulted in exceeding the acceptance guidelines. Since the accepted process utilizes a bounding analysis approach which is mostly driven by CDF contribution that does not already lead to LERF, there are no identified key assumptions or sources of uncertainty for this application (i.e. those which would change the conclusions from the risk assessment results presented here).

3.5.2.4 Summary A PRA technical adequacy evaluation was performed consistent with the requirements of RG 1.200, Revision 2. This evaluation combined with the details of the results of this analysis demonstrate with reasonable assurance that the proposed extension to the ILRT interval for CPS Unit 1 to fifteen years satisfies the risk acceptance guidelines in RG 1.174.

The Fire PRA results and Seismic CDF inputs were used to bound external event impacts of the proposed extension. The technical adequacy of the Fire PRA Model and the Seismic CDF input were qualitatively assessed and found to be adequate to support the conclusions found in , Section 7.0 of this submittal.

3.5.3 Summary of Plant-Specific Risk Assessment Results The findings of the CPS, Unit 1 Risk Assessment contained in Attachment 4 confirm the general findings of previous studies that the risk impact associated with extending the ILRT interval from three in ten years to one in 15 years is very small.

Based on the results from Attachment 4, Section 5.0, "Results," and the sensitivity calculations presented in Attachment 4, Section 6.0 "Sensitivities," the following conclusions regarding the assessment of the plant risk are associated with permanently extending the Type A ILRT test frequency to fifteen years:

RG 1.174 (Reference 3) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines "very small" changes in risk as resulting in increases of CDF below 10-6/yr and increases in LERF below 10-7/yr.

"Small" changes in risk are defined as increases in CDF below 10-5/yr and increases in LERF below 10-6/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 Page 20 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE in the Type A ILRT interval for the base case with corrosion included is 9.81E-09/yr, which falls within the "very small" change region of the acceptance guidelines in RG 1.174.

o If the EPRI Expert Elicitation Methodology Class 3a and Class 3b failure probabilities are used, the change is estimated as 1.15E-09/yr, which falls further within the very small change region of the acceptance guidelines in RG 1.174.

The change in dose risk for changing the Type A ILRT interval from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.80E-03 person-rem/yr using the EPRI guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr.

o The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology. The change in dose risk meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr.

The increase in the conditional containment failure frequency from the three in ten-year interval to one in fifteen years including corrosion effects using the EPRI guidance is 0.44%, which is below the acceptance criteria of 1.5% identified in the NRC SER on the issue (Reference 9).

o The increase in CCFP drops to about 0.05% using the EPRI Expert Elicitation methodology. This value meets both of the related acceptance criteria for change in CCFP of less than 1.5%.

To determine the potential impact from other hazard groups, an additional bounding assessment from the risk associated with the other relevant hazard groups for CPS (i.e.,

Seismic, Internal Fire, High Winds/Tornadoes, External Floods, Transportation) utilizing the latest information from various sources was performed. The total increase in LERF due to internal events and other hazard groups is 1.11E-07/yr, which is in Region II (i.e.,

"small" changes) of the RG 1.174 acceptance guidelines. The other acceptance criteria for change in population dose and change in CCFP are also still met when the other hazard groups are considered in the analysis.

Finally, a similar bounding analysis for the other hazard groups indicates that the total LERF from both internal events and the other hazard groups is 2.06E-06/yr, which is less than the RG 1.174 limit of 1E-05/yr given that the LERF is in Region II (i.e., "small" change in risk).

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

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

Page 21 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Reducing the frequency of Type A tests (i.e., 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.

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 ILRTs 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.

Details of the CPS, Unit 1, risk assessment are contained in Attachment 4 of this submittal.

3.6 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.

3.6.1 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 other 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., C1R15) in April 2015.

During outage C1R15, 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 April 2015, there are 2,253.31 pounds of combined Degraded and Unqualified coatings.

Allowed is 20,000 lbs. per Engineering Evaluation EE-00-143, Rev. 0. Therefore, 2,253.31 lbs.

of combined Degraded and Unqualified coatings would not be of concern during a LOCA. The list was compiled from the field notes and a coatings document review.

3.6.2 Inservice Inspection Program (ISI)

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 lSI inspection intervals are displayed in Tables 3.6.2-1 and 3.6.2-2.

Table 3.6.2-1, CPS IWE Examination Schedule 2nd and 3rd Ten-Year Inspection Interval 2nd Interval 9/10/08 to 9/9/11 1st Period C1R12 Page 23 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Table 3.6.2-1, CPS IWE Examination Schedule 2nd and 3rd Ten-Year Inspection Interval 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 (1) 9/10/18 to 9/9/21 1st Period C1R19 C1R20 C1R21 3rd Interval (1) 9/10/21 to 9/9/25 2nd Period C1R22 C1R23 C1R24 C1R25 3rd Interval (1) 9/10/25 to 9/9/28 3rd Period C1R26 C1R27 C1R28 (1) The dates and outages for the 3rd ISI inspection interval are proposed as the 3rd interval inspection plan and schedule have not been developed at this time.

Table 3.6.2-2, CPS IWL Examination Schedule 2nd and 3rd Ten-Year Inspection Interval C1R12 - 2010 C1R15 - 2015 C1R20 - 2020 C1R25 - 2025 C1R30 - 2030 Page 24 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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, 2001 Edition through the 2003 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 2001 Edition through the 2003 Addenda and the ASME Section XI Code shall apply to components and items classified as ASME Code Class MC or ASME Code Class CC.

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 3.6.2-3, Class MC and/or CC Exempt Components Exam Category Item Number Description Applicability to CPS E-A E1.20 Vent System Not Applicable Accessible Surface Areas E1.30 Moisture Barriers Not Applicable L-B L2.10 Tendon Not Applicable L2.20 Wire or Strand Not Applicable L2.30 Anchorage Not Applicable Hardware and Surrounding Concrete L2.40 Corrosion Not Applicable Protection Medium L2.50 Free Water Not Applicable Page 25 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Augmented Inspections Table 3.6.2-4, 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 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 Owners Activity Report-1, 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.

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 owners 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.

3.6.3 Supplemental Inspection Requirements With the implementation of the proposed change, TS 5.5.13 will be revised by replacing the reference to RG 1.163 (Reference 1) with reference to NEI 94-01, Revision 3-A (Reference 2).

This will require 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 26 of 52

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.

For CPS, no additional inspections are required.

3.6.4 Primary Containment Leakage Rate Testing Program - Type B and Type C Testing Program CPS Types B and C testing program requires 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 6), 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 2015 for CPS has shown substantial margin between the actual As-Found (AF) and As-Left (AL) outage summations and the regulatory requirements as described below:

The As-Found minimum pathway leak rate average for CPS shows an average of 7.2% of La with a high of 9.77% La.

The As-Left maximum pathway leak rate average for CPS shows an average of 13.7% of La with a high of 18.39% La.

Table 3.6.4-1 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 3.6.4-1, CPS Type B and C LLRT Combined As-Found/As-Left Trend Summary RFO C1R11 C1R12 C1R13 C1R14 C1R15 2008 2010 2011 2013 2015 As-Found Min 11597.7 18425.3 31551.3 33798.1 35305.4 Path (sccm)

Percentage 3.21 5.1 8.73 9.35 9.77 of La As-Left Max Path 24924.6 36848.3 52689.2 66201.4 66465.1 (sccm)

Percentage 6.89 10.19 14.58 18.32 18.39 of La 3.6.5 Type B and Type C Local Leak Rate Testing Program Implementation Review Table 3.6.5-1 identifies the components that were on extended intervals and have not demonstrated acceptable performance during the previous two outages for CPS.

Table 3.6.5-1, CPS Type B and C LLRT Program Implementation Review C1R14 - 2013 Component As- Evaluation As-left Cause of Corrective Scheduled found Limit SCCM Failure Action Interval SCCM SCCM 1IA175 - IA 56000 500 1680 Seat Replaced 30 months Instrument Air (1) leakage Valve Isolation Check Valve to 1IA006 1MC057 Page 28 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE C1R15 - 2015 Component As- Evaluation As-left Cause of Corrective Scheduled found Limit SCCM Failure Action Interval SCCM SCCM 1RF021 - 22000 10000 695 Intermediate Refurbished 30 months Containment position actuator Building floor Drain indication (2)

Inboard Isolation (2)

Control Valve 1MC070 1VR006B - 24980 500 6500 Seat Replaced 30 months Continuous leakage valve.

Containment Purge (3) Accepted As-Inboard Isolation Left above Valve admin limit by 1MC113 evaluation (1) 1IA175 failed it's as-found LLRT with 56000 sccm. However, the as-found, Minimum Pathway, leakage for penetration 1MC057 was 704.25 sccm. As-left value of 1680 sccm was determined to be acceptable.

(2) 1RF021 indicated intermediate when attempting to close. When the test volume was pressurized to test pressure, air flow was felt on the outlet of the test vent valve 1RF030B. This intermediate indication was the most probable cause of the excessive leakage. Resolved intermediate position indication in 1RF021 and re-tested.

(3) 1VR006B failed its as-found LLRT with 24980 sccm. This is added to the Max-Path leakage. This is also a secondary containment bypass penetration so this leakage applies to both the 0.6La total and the 0.08La. The previous test results for this penetration was 6,904 sccm. Following maintenance, the As-Left Maximum Pathway leakage for penetration 1MC113 was tested as 6500 sccm. This value is 94.2% of the 1MC113 As-left Max Pathway leakage measured in C1R12 which had a value of 6900 sccm (6500 / 6900 = 94.2%). The leakage identified in IR 2501033 for 1MC113, was determined to be an improvement of past leakage for the penetration and secondary containment bypass as a whole. Because of this improvement Engineering accepted the as-found test results of 1MC113 and the impact on 0.08La (secondary containment bypass), and 0.6La (containment leakage). No further action was required.

The percentage of the total number of CPS Type B tested components that are on 120-month extended performance-based test intervals is 96%.

The percentage of the total number of CPS Type C tested components that are on 60-month extended performance-based test intervals is 90%.

3.7 Operating Experience During the conduct of the various examinations and tests conducted in support of the Containment related programs previously mentioned, issues that do not meet established Page 29 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE criteria or that provide indication of degradation, are identified, placed into the site's corrective action program, and corrective actions are planned and performed.

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

IN 1992-20, "Inadequate Local Leak Rate Testing" IN 2010-12, "Containment Liner Corrosion" IN 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" Each of these areas is discussed in detail in Sections 3.7.1 through 3.7.3, respectively.

3.7.1 IN-92-20, "Inadequate Local Leak Rate Testing" The issue discussed in IN-92-20, "Inadequate Local Leak Rate Testing," was based on events at four different plants, Quad Cities Nuclear Power Station, Dresden Nuclear Power Station, Perry Nuclear Plant and the Clinton Power Station. The common issue in the four events was the failure to adequately perform local leak rate testing on different penetration configurations leading to problems that were discovered during ILRT tests in the first three cases.

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

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

In the event at CPS, relief valve discharge lines that were assumed to terminate below the suppression pool minimum drawdown level were discovered to terminate at a level above that datum. These lines needed to be reconfigured and the valves should have been Type C LLRT tested. To correct this problem, CPS removed the vacuum breaker connections and the flanges and extended the pipes to ensure that a water seal would be maintained.

As for the testing of two-ply stainless steel bellows at CPS, a modification (i.e., FH-030) was installed on the Inclined Fuel Transfer System (IFTS) containment penetration in 1995 to eliminate the concern raised by IN-92-20 and to allow bellows testing to be performed using Type B test methods. The testing assembly provides a means of applying a static test pressure to the bellows to ensure containment integrity will be maintained in accordance with 10 CFR 50, Appendix J.

There are no mechanical bellows consistently exceeding the administrative limit of 500 sccm at CPS.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 3.7.2 IN 2010-12, "Containment Liner Corrosion" This IN was issued to alert plant operators to three events that occurred where the steel liner of the containment building was corroded and degraded. At Beaver Valley and Brunswick plants material had been found in the concrete which trapped moisture against the liner plate and corroded the steel. In one case it was material intentionally placed in the building and in the other case it was foreign material which had inadvertently been left in the form when the wall was poured. But the result in both cases was that the material trapped moisture against the steel liner plate leading to corrosion. In the third case, an insulating material placed between the concrete floor and the steel liner plate at Salem adsorbed moisture and led to corrosion of the liner plate.

The situation that occurred at Salem is not likely to take place at CPS. CPS does not have moisture barriers.

CPS should not experience the events that took place at Beaver Valley and Brunswick. EGC has implemented periodic examinations during refueling outages on metallic containment structures or liners in accordance with the Section XI, Subsection IWE. The applicable EGC visual examination procedure, ER-AA-335-018, requires the conditions described in the IN examples to be recorded. Conditions that may affect containment surface integrity are then required to be evaluated by engineering evaluation or repair/replacement prior to startup from refueling outages.

3.7.3 IN 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" The containment basemat metallic shell and liner plate seam welds of pressurized water reactors are embedded in a 3-to 4-feet thick concrete floor during construction and are typically covered by a leak-chase channel system that incorporates pressurizing test connections. This system allows for pressure testing of the seam welds for leak-tightness during construction and also in service, as required. A typical basemat shell or liner weld leak-chase channel system consists of steel channel sections that are fillet welded continuously over the entire bottom shell or liner seam welds and subdivided into zones, each zone with a test connection.

Each test connection consists of a small carbon or stainless steel tube (less than 1-inch diameter) that penetrates through the back of the channel and is seal-welded to the channel steel. The tube extends up through the concrete floor slab to a small steel access (junction) box embedded in the floor slab. The steel tube, which may be encased in a pipe, projects up through the bottom of the access box with a threaded coupling connection welded to the top of the tube, allowing for pressurization of the leak-chase channel. After the initial tests, steel threaded plugs or caps are installed in the test tap to seal the leak-chase volume. Gasketed cover plates or countersunk plugs are attached to the top of the access box flush with the containment floor. In some cases, the leak-chase channels with plugged test connections may extend vertically along the circumference of the cylindrical containment shell or liner to a certain height above the floor.

The CPS IWE/IWL Containment Inspection Program indicates the applicable requirements are ASME Section XI Code 2001 Edition through 2003 Addenda, and 10CFR50.55a requirements.

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Attachment 1 EVALUATION OF PROPOSED CHANGE A review of the program implementation document determined that CPS has weld leak chase design and test connections for containment liner plate seam welds. These are designed to test the enclosed seam welds for leakage. The design of the containment includes stainless steel liners. Therefore, the concern with corrosion is very unlikely and inspection is not required.

However, the seam welds are tested as part of the containment leakage test boundary per CPS 9861.01, "Integrated Leak Rate Test," by venting the leak chase channels by removal of leak chase channel plugs.

3.7.4 Results of Recent Inspections 3.7.4.1 Containment and Drywell Coatings - C1R15 - 2015 3.7.4.1.1 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 has 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. Overall the coatings at the concrete wall, liner plate, and existing steel are in good condition. In some areas, the coating on the concrete basement floor is in poor condition. A condition report, (AR 02498541) has been generated for the identified condition. Areas of mechanical damage to concrete substrate are present throughout the drywell floor. A large area (approximately 250 sq. ft.), is scheduled to be repaired refueling outage C1R17 (Spring 2017).

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.

3.7.4.1.2 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 liner plate on Elevation 737' is approximately 15 feet away. On this elevation, there are areas of Page 32 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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', one area of mechanical damage on the inner concrete wall exhibits spalling with rebar exposed. Overall for these elevations, the liner plate and inner wall are in good condition.

828' Elevation Most of the containment floor is covered with plastic for FME and to facilitate refuel services activities and as a result, it was not feasible to inspect. The dome was remotely observed from outside of the contaminated area. There were no concerns or issues about coatings degradation on this elevation.

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.

3.7.4.1.3 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.

3.7.4.2 IWE - C1R14 - 2013 3.7.4.2.1 Containment In-Service Inspections (CISI) on Containment liner 737' Elevation CISI was performed on containment liner on the 737' elevation during C1R14 for ASME Section XI IWE inspection requirements. The following observations document the C1R14 as-found reportable indications.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 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 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.

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

755' Elevation CISI were performed on containment liner on the 755' elevation during C1R14 for ASME Section XI IWE inspection requirements. The following observations document the C1R14 as-found reportable indications.

At 24 degrees, 776' elevation, 1" x 2" peeled, cracked and stained coatings 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 were performed on containment liner on the 781' elevation during C1R14 for ASME Section XI IWE inspection requirements. The following observations document the C1R14 as-found reportable indications.

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 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.

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.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 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 were performed on containment liner on the 800' elevation during C1R14 for ASME Section XI IWE inspection requirements. The following observations document the C1R14 as-found reportable indications.

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

At 85 degrees, 815' elevation, 8" x 8" area had 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 190 degrees, 823' elevation, stained coatings 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.

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

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

3.7.4.3 IWE/IWL - C1R15 - 2015 3.7.4.3.1 Inspection Scope In accordance with the 2nd 10-Year Inspection Interval, the scheduled examinations for ASME Section XI, IWE/IWL Containment were completed during outage C1R15. The scope of inspection as listed in the Second Inspection Period in the current Interval was completed.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 3.7.4.3.2 Results This inspection was conducted for all accessible areas in accordance with ER-AA-335-018, "Visual Examination of ASME /WE Class MC and Metallic Liners of IWL Class CC Components." Thirty reports were created to organize the inspections. These reports are numbered C1R15-001 to C1R15-030, and the corresponding areas in each report are listed below. Inspector observations and all recordable indications are listed in this summary. In general, the inspections revealed these common observations of minor significance:

deteriorated caulking, as identified in reports C1R15-022, -024, and -029; flaking/peeling paint, coating damage, missing coating, light rust, light corrosion. There was only one area where the inspector noted the presence of medium to heavy rust but no apparent loss of material (Report No. C1R15-029). This location corresponds to Penetration 1K3E (OD). A picture of this observation is included in the report. Deteriorated coating repairs will be implemented as required. Deteriorated coatings were identified in reports C1R15-020, -025, -028, and -029. In cases where no penetration pictures were taken, the recorded conditions do not indicate loss in material and are thereby acceptable. This statement applies to reports C1R15-021, C1R15-026 and C1R15-027.

General Visual Examinations were conducted by VT-3 examiners to assess the general condition of the surface or component. Below is a listing of individual observations noted for each area and penetrations inspected:

Report No. C1R15-001 - Containment Steam Tunnel Concrete and liner inspections were satisfactory except for this recordable indication: (1) Flaking paint, light rust, coating damaged all around penetration 1MC-45 on the liner.

Report No. C1R15-002 - Containment Liner - 828' and Dome 0 - 90 deg. Satisfactory. No recordable indication.

Report No. C1R15-003 - Containment Liner - 828' and Dome 90 - 180 deg. Satisfactory. No recordable indication.

Report No. C1R15-004 - Containment Liner - 828' and Dome 180 - 270 deg. Satisfactory. No recordable indication.

Report No. C1R15-005 - Containment Liner - 828' and Dome 270 - 360 deg. Satisfactory. No recordable indication.

Report No. C1R15-006 - OD Concrete. 707' AB, 270 - 360 deg.

Concrete coatings were satisfactory except for two areas: (1) Az. 285° El. 707' had flaking and peeling paint. (2) Az. 293° El. 710' has peeling paint.

Report No. C1 R15-007- OD Concrete. 712' FB.90-180 deg. Satisfactory. No recordable indication.

Report No. C1R15-008 - OD Concrete. Reactor Water Cleanup Mezz. 750' AB. 40 - 90 deg.

Satisfactory. No recordable indication.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Report No. C1R15-009 - OD Concrete. FB 737', 90 - 180 deg. Satisfactory. No recordable indication.

Report No. C1 R15-010 - OD Concrete. AB 737'. 270- 90 deg. Satisfactory. No recordable indication.

Report No. C1R15-011 - OD Concrete. AB 762'. 270 - 90 deg. Concrete and coatings were satisfactory except for two areas: (1) Az. 295° El. 770' had flaking paint. (2) Az. 285° El. 770' has flaking paint.

Report No. C1 R15-012-OD Concrete. FB 755', 90 -270 deg. Satisfactory. No recordable indication.

Report No. C1R15-013 - OD Concrete. AB 781', and AB Steam Tunnel. Upper Aux. Steam Tunnel concrete is satisfactory. Typical pictures above the main steam system penetrations.

Recordable indications are: (1) missing surface coating and coating damage between feedwater penetrations and around penetration 1MC-14, apparent moisture damage. (2) Flaking paint on Az. 80°, El. 780' and Az. 85°, El. 785'.

Report No. C1 R15-014-OD Concrete. FB 781', 90 -270 deg. Satisfactory. No recordable indication.

Report No. C1 R15-015 - OD Concrete. Gas Control Boundary 800' and above AB & FB 0-360 deg. Satisfactory. No recordable indication.

Report No. C1R15-016 - Containment Equipment. Hatch Satisfactory. No recordable indication.

Report No. C1 R15-017 - Containment Personnel Lock- 741'. Satisfactory. No recordable indication.

Report No. C1R15-018 - Containment Personnel Lock - 832'. Satisfactory. No recordable indication.

Report No. C1R15-019 - Fuel Transfer Tube 755' FB and 770' Containment Building The outboard side was underwater and inaccessible. The inboard side had recordable indications: (1) Light corrosion noted on weld-o-lets. No material loss. No change from previous report. (2) Missing/incomplete coating on upper transfer tube area. No change noted from previous report.

Report No. C1R15-020 - Penetrations ID & OD Above 762', 0 - 90 deg. There were no recordable indications for all ID side inspections. The OD side inspection found recordable indications: (1) Flaking paint, moderate rust, no apparent material loss for penetrations 1MC-60, 1MC-61, 1MC-64, 1MC-65, 1MC-74, 1MC-89, 1MC210, and 1MC-211.

Report No. C1R15-021 - Penetrations ID & OD Below 737', 37 - 355 deg. There were no recordable indications for all ID side inspections. The OD side inspection found recordable indications: (1) Flaking paint, light rust, no apparent material loss for penetrations 1MC-12, and Page 37 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 1MC-13. This report only contains inspector comments on the recordable indications.

Inadvertently, the inspectors did not submit pictures for this report.

Report No. C1R15-022 - Penetrations ID & OD Above 737', 270 - 360 deg. There were no recordable indications for all ID side inspections. The OD side inspection found recordable indications: (1) Flaking and cracking paint, and light rust for penetrations 1MC-17, 1MC-20, and 1MC-29. Also, penetration 1MC-20 shows signs of caulking deterioration.

Report No. C1R15-023 - Penetrations ID & OD Above 737', 0 - 90 deg. There were no recordable indications for all ID side inspections. Inspector noted that one penetration, 1MC-77, was not accessible from the OD side, and that it was previously noted in the last inspection in C1R11. The OD side inspection found recordable indication: (1) Flaking paint and light rust for penetrations 1MC-17, 1MC-20, and 1MC-29.

Report No. C1R15-024 - Penetrations OD Above 737', 90 - 180 deg. Inspections of all penetrations were satisfactory, except for two: (1) Penetration 1MC-46 (OD) had deteriorated caulking, flaking paint and missing coating, (2) Penetration 1MC-47 (OD) had flaking paint and missing coating.

Report No. C1R15-025 - Penetrations OD Above 737', 180-360 deg. Inspections of all penetrations were satisfactory, except for three: (1) Penetration 1MC-35 (OD) had flaking paint, moderate rust, but no apparent material loss. (2) Penetration 1MC-76 (OD) had flaking paint and light rust. (3) Penetration 1MC-78 (OD) had flaking paint and light rust.

Report No. C1R15-026 - Penetrations OD Above 762', 120 - 277 deg. Inspections of all penetrations were satisfactory, except for three: (1) Penetration 1MC-102 (OD) had flaking paint, light rust, but no apparent loss of material, (2) Penetration 1MC-103 (OD) had flaking paint, light rust, but no apparent loss of material, (3) Penetration 1MC-104 (OD) had flaking paint, light rust, but no apparent loss of material. This report contains inspector comments and pictures for recordable indications in three locations. Inadvertently, the inspectors did not submit picture for one location.

Report No. C1R15-027 - Penetrations OD Above 781', 90 - 180 deg. (1 Penetration both ID &

OD). Inspections of all penetrations were satisfactory, except for four: (1) Penetration 1MC-50 (OD) had light rust but no apparent loss of material. (2) Penetration 1MC-51 (OD) had light rust but no apparent loss of material. (3) Penetration 1MC-155 (OD) had flaking paint and light rust but no apparent loss of material. (4) Penetration 1MC-170 (OD) had flaking paint, light rust but no apparent loss of material. This report contains inspector comments for recordable indications on four locations. Inadvertently, the inspectors only provided picture for one location.

Report No. C1R15-028 - Penetrations OD Above 762', 180 - 360 deg. Inspections of all penetrations were satisfactory, except for six. (1) Penetration 1MC-166 (OD) had flaking and peeling paint. (2) Penetration 1MC-168 (OD) had flaking paint, medium corrosion and rust, but no apparent loss of material. (3) Penetration 1MC-171 (OD) had flaking paint, corrosion, no apparent loss of material. (4) Penetration 1MC-203 (OD) had flaking and peeling paint, corrosion, but no apparent loss of material. (5) Penetration 1MC-204 (OD) had flaking paint, corrosion, but no apparent loss of material. (6) Penetration 1MC-204 (OD) had flaking and peeling paint.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Report No. C1R15-029 - Penetrations OD Above 762' & 781', 0 - 180 deg. Inspections of all penetrations were satisfactory, except for four: (1) Penetration 1K1E (OD) had discoloration and light rust. (2) Penetration 1K3N (OD) had flaking paint and light rust but no apparent material loss. (3) Penetration 1P1E-2 (OD) had rust but no apparent material loss. (4) Penetration 1K3E (OD) had coating degradation, medium to heavy rust, but no apparent loss of material.

Report No. C1R15-030 - Penetrations OD Above 762' & 781', 180 - 360 deg. Satisfactory. No recordable indication.

In summary, Containment ISI inspections completed in C1R15 identified minor surface conditions such as flaking paint, peeling paint, rust, light corrosion, coating damage, missing coating, and incomplete coating. However, none of the examinations found any loss of metal and none were of concern to the containment function.

3.8 NRC SE Limitations and Conditions 3.8.1 Limitations and Conditions Applicable to NEI 94-01, Revision 2-A The NRC staff found that the use of NEI TR 94-01, Revision 2, was acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT surveillance interval to 15 years, provided the following conditions as listed in Table 3.8.1-1 were satisfied:

Table 3.8.1-1: NEI 94-01, Revision 2-A, Limitations and Conditions Limitation/Condition (From Section 4.0 of SE) CPS Response For calculating the Type A leakage rate, the CPS will utilize the definition in NEI 94-01 licensee should use the definition in the NEI Revision 3-A, Section 5.0. This definition has TR 94-01, Revision 2, in lieu of that in remained unchanged from Revision 2-A to ANSI/ANS-56.8-2002. (Refer to SE Revision 3-A of NEI 94-01.

Section 3.1.1.1.)

The licensee submits a schedule of Reference Section 3.6.2 and Tables 3.6.2-1 containment inspections to be performed and 3.6.2-2.

prior to and between Type A tests. (Refer to SE Section 3.1.1.3.)

The licensee addresses the areas of the Reference Section 3.6.2 of this submittal.

containment structure potentially subjected to degradation. (Refer to SE Section 3.1.3.)

The licensee addresses any tests and There are no major modifications planned.

inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4.)

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Attachment 1 EVALUATION OF PROPOSED CHANGE Table 3.8.1-1: NEI 94-01, Revision 2-A, Limitations and Conditions Limitation/Condition (From Section 4.0 of SE) CPS Response The normal Type A test interval should be CPS will follow the requirements of NEI 94-01 less than 15 years. If a licensee has to utilize Revision 3-A, Section 9.1. This requirement the provision of Section 9.1 of NEI TR 94-01, has remained unchanged from Revision 2-A Revision 2, related to extending the ILRT to Revision 3-A of NEI 94-01.

interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an In accordance with the requirements of NEI unforeseen emergent condition. (Refer to SE 94-01 Revision 2-A, SER Section 3.1.1.2, Section 3.1.1.2.) CPS will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

For plants licensed under 10 CFR 52, Not applicable. CPS was not licensed under applications requesting a permanent 10 CFR 52.

extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No.

1009325, Revision 2, including the use of past containment ILRT data.

3.8.2 Limitations and Conditions Applicable to NEI 94-01, Revision 3-A The NRC staff found that the guidance in NEI TR 94-01, Revision 3, was acceptable for referencing by licensees in the implementation for the optional performance-based requirements of Option B to 10 CFR 50, Appendix J. However, the NRC staff identified two conditions on the use of NEI TR 94-01, Revision 3 (Reference NEI 94-01 Revision 3-A, NRC SE 4.0, Limitations and Conditions):

Topical Report Condition 1 NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI TR 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR MSIVs), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Response to Condition 1 Condition 1 presents three separate issues that are required to be addressed. They are as follows:

ISSUE 1 - The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.

ISSUE 2 - In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.

ISSUE 3 - Use of the allowed 9-month extension for eligible Type C valves is only authorized for non-routine emergent conditions with exceptions as detailed in NEI 94-01, Revision 3-A, Section 10.1.

Response to Condition 1, ISSUE 1 The post-outage report shall include the margin between the Type B and Type C Minimum Pathway Leak Rate (MNPLR) summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.60 La.

Response to Condition 1, ISSUE 2 When the potential leakage understatement adjusted Type B and C MNPLR total is greater than the CPS leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the CPS leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.

Response to Condition 1, ISSUE 3 CPS will apply the 9-month allowable interval extension period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.

Topical Report Condition 2 The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending Page 41 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 refueling outages. Type C tests involve valves, which in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for.

Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.

When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total leakage, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Response to Condition 2 Condition 2 presents two separate issues that are required to be addressed. They are as follows:

ISSUE 1 - Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.

ISSUE 2 - When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Response to Condition 2, ISSUE 1 The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25%

in the LLRT periodicity. As such, CPS will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate, which will increase the As-Left leakage total for each Type C component currently on greater than a 60-month test interval up to the 75-month extended test interval. This will result in a combined conservative Page 42 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Type C total for all 75-month LLRTs being "carried forward" and will be included whenever the total leakage summation is required to be updated (either while on line or following an outage).

When the potential leakage understatement adjusted leak rate total for those Type C components being tested on greater than a 60-month test interval up to the 75-month extended test interval is summed with the non-adjusted total of those Type C components being tested at less than or equal to a 60-month test interval, and the total of the Type B tested components, if the MNPLR is greater than the CPS leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the CPS leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

Response to Condition 2, ISSUE 2 If the potential leakage understatement adjusted leak rate MNPLR is less than the CPS leakage summation limit of 0.50 La, then the acceptability of the greater than a 60-month test interval up to the 75-month LLRT extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

In addition to Condition 1, ISSUES 1 and 2, which deal with the MNPLR Type B and C summation margin, NEI 94-01, Revision 3-A, also has a margin related requirement as contained in Section 12.1, "Report Requirements."

A post-outage report shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B and Type C tests, if performed during that outage. The technical contents of the report are generally described in ANSI/ANS-56.8-2002 and shall be available on-site for NRC review. The report shall show that the applicable performance criteria are met, and serve as a record that continuing performance is acceptable. The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit. Adverse trends in the Type B and Type C leakage rate summation shall be identified in the report and a corrective action plan developed to restore the margin to an acceptable level.

At CPS, in the event an adverse trend in the aforementioned potential leakage understatement adjusted Type B and C summation is identified, then an analysis and determination of a corrective action plan shall be prepared to restore the trend and associated margin to an acceptable level. The corrective action plan shall focus on those components which have contributed the most to the adverse trend in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

At CPS an adverse trend is defined as three (3) consecutive increases in the final pre-mode change Type B and C MNPLR leakage summation values, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 3.9 Conclusion NEI 94-01, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, describe an NRC-accepted approach for implementing the performance-based requirements of 10 CFR 50, Appendix J, Option B. It incorporated the regulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals to 15 years and Type C test intervals to 75 months. NEI 94-01, Revision 3-A delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies. CPS is adopting the guidance of NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, for the CPS 10 CFR 50, Appendix J testing program plan.

Based on the previous ILRTs conducted at CPS, it may be concluded that the permanent extension of the containment ILRT interval from 10 to 15 years represents minimal risk to increased leakage. The risk is minimized by continued Type B and Type C testing performed in accordance with Option B of 10 CFR 50, Appendix J, Drywell Inspections and the overlapping inspection activities performed as part of the following CPS inspection programs:

• Inservice Inspection Program IWE/IWL

• Containment Inspections per TS SR 3.6.1.1.1

• Protective Coatings Program This experience is supplemented by risk analysis studies, including the CPS risk analysis provided in Attachment 4. The risk assessment concluded that increasing the ILRT interval on a permanent basis to a one-in-fifteen year frequency is not considered to be significant since it represents only a small change in the CPS risk profile.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. 10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR 50, "Leakage Rate Testing of Containment of Water Cooled Nuclear Power Plants." Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.

The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B and Type C testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of 10 CFR 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that Page 44 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type C test frequencies will not directly result in an increase in containment leakage.

EPRI TR-1009325, Revision 2, provided a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance.

NEI 94-01, Revision 3-A, Section 9.2.3.1 states that Type A ILRT intervals of up to 15 years are allowed by this guideline. The Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI Report 1018243 (Formerly TR-1009325, Revision 2) indicates that, in general, the risk impact associated with ILRT interval extensions for intervals up to 15 years is small. However, plant-specific confirmatory analyses are required.

The NRC staff reviewed NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2.

For NEI TR 94-01, Revision 2, the NRC staff determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163. The NRC staff finds that the Type A testing methodology as described in ANSI/ANS-56.8-2002, and the modified testing frequencies recommended by NEI TR 94-01, Revision 2, serves to ensure continued leakage integrity of the containment structure. Type B and Type C testing ensures that individual penetrations are essentially leak tight. In addition, aggregate Type B and Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.

For EPRI Report No. 1009325, Revision 2, a risk-informed methodology using plant-specific risk insights and industry ILRT performance data to revise ILRT surveillance frequencies, the NRC staff finds that the proposed methodology satisfies the key principles of risk-informed decision making applied to changes to TSs as delineated in RG 1.177 and RG 1.174. The NRC staff, therefore, found that this guidance was acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.2 of the Safety Evaluation (SE).

The NRC staff reviewed NEI TR 94-01, Revision 3, and determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, as modified by the conditions and limitations summarized in Section 4.0 of the associated Safety Evaluation. This guidance included provisions for extending Type C LLRT intervals up to 75 months. Type C testing ensures that individual containment isolation valves are essentially leak tight. In addition, aggregate Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The NRC staff, therefore, found that this guidance, as modified to include two limitations and conditions, was acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing. Any applicant may reference NEI TR 94-01, Revision 3, as modified by the associated SE and approved by the NRC, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, in a licensing action to satisfy the requirements of Option B to 10 CFR 50, Appendix J.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 4.2 Precedent This LAR is similar in nature to the following license amendments to extend the Type A Test Frequency to 15 years and the Type C test frequency to 75 months as previously authorized by the NRC:

Surry Power Station, Unit 1 (Reference 24)

Donald C. Cook Nuclear Plant, Unit 1 (Reference 25)

Beaver Valley Power Station, Unit Nos. 1 and 2 (Reference 26)

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (Reference 27)

Peach Bottom Atomic Power Station, Units 2 and 3 (Reference 28)

Comanche Peak Nuclear Power Plant, Units 1 and 2 (Reference 40) 4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (EGC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed activity involves the extension of the Clinton Power Station (CPS), Unit 1, Type A containment test interval to 15 years, and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.

The change in dose risk for changing the Type A Integrated Leak Rate Test (ILRT) interval from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.80E-03 person-rem/yr using the EPRI guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr. The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology. The change in dose risk meets both of the related acceptance for change in population dose of less than 1.0 person-rem/yr or less Page 46 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE than 1% person-rem/yr. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

In addition, as documented in NUREG-1493, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The CPS, Unit 1 Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) activity based, and, (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance.

Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME) Section XI, and Technical Specifications (TS) requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for CPS. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how CPS is operated.

Therefore, the proposed change does not result in 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 amendment to the TS 5.5.13, "Primary Containment Leakage Rate Testing Program," involves the extension of the CPS, Unit 1 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plants ability to mitigate the consequences of an accident.

The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment.

The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for CPS. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how CPS is operated.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed amendment to TS 5.5.13 involves the extension of the CPS, Unit 1 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves the extension of the interval between Type A containment leak rate tests and Type C tests for CPS, Unit 1. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained.

The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.

The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for CPS, Unit 1. This exception was for an activity that has taken place; therefore, the deletion is solely an administrative action and does not change how CPS is operated and maintained. Thus, there is no reduction in any margin of safety.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed Page 48 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

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

September 1995

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

3. Regulatory Guide 1.174, Revision 2, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis," May 2011

4. Regulatory Guide 1.200, Revision 2, "An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities," March 2009

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

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

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

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

9. 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 Page 49 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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

10. 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

11. 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

12. Clinton PRA Peer Review Report, October 2000

13. 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

14. Letter from D. Pickett (NRC) to M. Lyon (CPS), Clinton Power Station Unit 1 - Issuance of Amendment 106 Regarding Revision of Technical Specifications for the Drywell to Permit Bypass Testing On a 10-year Frequency (TAC NO. M94889), dated September 4, 1996

15. Letter from D. Pickett (NRC) to J. Skolds (AmerGen), Clinton Power Station Unit 1 -

Issuance of Amendment 160 Regarding The One-Time Technical Specification Change To Extend The Test Interval For The Next Appendix J Type A Test And The Next Drywell Bypass Leakage Rate Test From 10 To 15 Years (TAC NO. MB7675), dated January 8, 2004

16. Letter from K. Jabbour (NRC) to C. Crane (AmerGen), Clinton Power Station Unit 1 -

Issuance of Amendment 167 Regarding the Application of Alternative Source Term Methodology (TAC NO. MB8365), dated September 19, 2005

17. Letter from K. Jabbour (NRC) to C. Crane (AmerGen), Clinton Power Station Unit 1 -

Issuance of Amendment 173 Regarding the Revision Of Secondary Containment Bypass Leakage Surveillance Requirement (TAC NO. MC6488), dated March 21, 2006

18. Letter from J. Hopkins (NRC) to J. Sipek (CPS), Clinton Power Station, Unit 1 - Issuance of Amendment 121 Regarding the Deferral of The Next Scheduled Local Leak Rate Test for Valve 1MC-042 Until The Seventh Refueling Outage (TAC NO. MA3754), dated March 8, 1999

19. Letter from J. Hopkins (NRC) to O. Kingsley (Exelon), Clinton Power Station, Unit 1 -

Issuance of Amendment 145 Regarding the Replacement of Individual Main Steamline Leakage Limits With An Aggregate Leakage Limit, Revising Technical Specification Surveillance Requirement 3.6.1.3.9, dated March 26, 2002 Page 50 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE

20. EPRI Report 1018243, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325, dated October 2008

21. Letter from K. Jury (AmerGen) to NRC, Clinton Power Station, Unit 1 - Request for Amendment to Technical Specification 3.6.5.1, "Drywell" and 5.5.13, "Primary Containment Leakage rate Testing Program," dated November 16, 2006

22. Letter from P. Simpson (AmerGen) to NRC, Clinton Power Station, Unit 1 - Withdrawal of Request for Amendment to Technical Specification 3.6.5.1, "Drywell" and 5 .5.13, "Primary Containment Leakage Rate Testing Program," dated April 30, 2007

23. NEI 00-02, "Probabilistic Risk Assessment Peer Review Process Guidance," Rev. A3, dated March 2000

24. Letter to D. Heacock from S. Williams (NRC), Surry Power Station, Unit 1 - Issuance of Amendment Regarding the Containment Type A and Type C Leak Rate Tests, dated July 3, 2014 (ML14148A235)

25. Letter to L. Weber from A. Dietrich (NRC), Donald C. Cook Nuclear Plant, Unit 1 -

Issuance of Amendments Re: Containment Leakage Rate Testing Program, dated March 30, 2015 (ML15072A264)

26. Letter to E. Larson from T. Lamb (NRC), Beaver Valley Power Station, Unit Nos. 1 and 2

- Issuance of Amendment Re: License Amendment Request to Extend Containment Leakage Rate Test Frequency, dated April 8, 2015 (ML15078A058)

27. Letter to G. Gellrich from A. Chereskin (NRC), Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Issuance of Amendments Re: Extension of Containment Leakage Rate Testing Frequency, dated July 16, 2015 (ML15154A661)

28. Letter to B. Hanson from R. Ennis (NRC), Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Extension of Type A and Type C Leak Rate Test Frequencies (TAC NOS. MF5172 AND MF5173), dated September 8, 2015 (ML15196A559)

29. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-S-2002, New York, New York, April 2002

30. ASME/American Nuclear Society, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009, March 2009

31. Clinton Power Station 2009 PRA Peer Review Report, April 2010.

32. 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, dated March 27, 2002 (Accession Number ML020920100)

Page 51 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE

33. U.S. Nuclear Regulatory Commission (NRC), Memorandum to Brian W. Sheron (Director Office of Nuclear Regulatory Research), From Patrick Hiland, (Chairman, Safety/Risk Assessment Panel for Generic Issue 199), "Safety/Risk Assessment Results for Generic Issue 199, 'Implications of Updated Probabilistic Seismic Hazard Estimates In Central and Eastern United States on Existing Plants'," dated September 2, 2010 (Accession Number ML11356A034)

34. CPS 2011 Self-Assessment of the PRA Against the ASME PRA Standard Requirements Notebook, CPS-PSA-016, Revision 1, December 2011

35. CPS Station Internal Events PRA, Model of Record 2006C

36. Letter from N. DiFrancesco (NRC) to M. Pacilio (Exelon), Clinton Power Station, Unit 1 -

Issuance of Amendment 192 Re: Technical Specification Change for The Relocation of Specific Surveillance Frequency Requirements Based On Technical Specification Task Force (TSTF)-425 (TAC NO. ME3332), dated February 15, 2011

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

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

39. BWROG PSA Peer Review Certification Implementation Guidelines, January 1997.

40. Letter from B. Singal (NRC) to R. Flores (Luminant), Comanche Peak Nuclear Power Plant, Units 1 and 2 - Issuance of Amendments Re: Technical Specification Change For Extension of the Integrated Leak Rate Test Frequency From 10 to 15 Years (CAC Nos.

MF5621 and MF5622), dated December 30, 2015 Page 52 of 52

ATTACHMENT 2 Mark-up of Technical Specifications Page TS 5.5.13 Page 5.0-16a

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

S.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 so, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, *Performance-Based Containment Leak-Test Program,* dated September 1995,as .

modified by the *following exceptions: (1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Type A tests, and (2) NEI 94 1995, Section 9.2.3: The first Type A test performed after November 23, 1993 shall be performed no.

later than November 23, 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 9.0 psig.

• The maximum allowable primary containment leakage rate L.,, at P8 ,

sha~l be 0.65\ of primary containment air weight*per day.

Leakage Rate acceptance c~iteria are:

• a. Primary containment leakage rate acceptance criterion is S 1.0 L.. During the first unit startup following testing in accord~ce with this program, the leak rate acceptance criteria are S 0.60 La for the Type B and Type C tests and S o.75 L. for Type A tests;

b. Air lock testing acceptance criteria are:

1) overall air lock leakage rate is s s scfh when teste~ at

~ Par

2) For each door, leakage rate is S 5 scfh when the gap.l between door seals is pressurized ~ P**

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.

NEI 94-01, "Industry Guideline for Implementing Performance-Based (continued)

Option of 10 CFR Part 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, I I

~

CLINTON S.0-16a .Amendment No. 16 0

ATTACHMENT 3 Mark-up of Technical Specification Bases Page (Provided for Information Only)

TS Bases 3.6.1.1 Page B 3.6-5

Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.1 (continued)

REQUIREMENTS (continued) With regard to leakage rate values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument indication uncertainties (Ref. 7).

SR 3.6.1.1.2 With respect to primary containment integrated leakage rate testing, the primary containment hydrogen recombiners (located outside the primary containment) are considered extensions of the primary containment boundary. This requires the smaller of the leakage from the PCIVs that isolate the primary containment hydrogen recombiner, or from the piping boundary outside containment, to be included in the ILRT results. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

With regard to leakage rate values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument indication uncertainties (Ref. 7).

REFERENCES 1. USAR, Section 6.2.

2. USAR, Section 15.6.5.

3. 10 CFR 50, Appendix J, Option B.

4. USAR, Section 6.2.1. 3-A

5. NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J." 2002

6. ANSI/ANS-56.8-1994, "American National Standard for Containment System Leakage Testing Requirement."

7. Calculation IP-0-0056.

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

CLINTON B 3.6-5 Revision No. 4-6

ATTACHMENT 4 Risk Assessment for CPS Regarding the ILRT (Type A)

Permanent Extension Request

Risk Management Team Clinton Power Station (CPS)

Risk Assessment for CPS Regarding the ILRT (Type A) and DWBT Permanent Extension Request CL-LAR-07 Revision 0

Risk Management Team Risk Management Team Revisions:

REV. DESCRIPTION PREPARER/DATE REVIEWER/DATE APPROVER/DATE C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 METHEDOLOGY............................................................................................... 2-1 3.0 GROUND RULES.............................................................................................. 3-1 4.0 INPUTS ............................................................................................................. 4-1 4.1 General Resources Available ................................................................. 4-1 4.1.1 GGNS DWBT Method ........................................................... 4-11 4.1.2 RBS DWBT Method .............................................................. 4-12 4.1.3 CPS DWBT Method .............................................................. 4-12 4.2 Plant Specific Inputs ............................................................................. 4-19 4.3 CPS Population Dose Derivation .......................................................... 4-25 4.4 Impact of Extension on Detection of Component Failures That Lead to Leakage (Small and Large)................................................................... 4-43 4.5 Impact of Extension on Detection of Steel Corrosion that Leads to Leakage ................................................................................................ 4-46 4.6 Impact of DWBT Interval Extension of Release Categories .................. 4-53 4.6.1 DWBT Data Analysis ............................................................ 4-54 5.0 RESULTS .......................................................................................................... 5-1 5.1 Step 1 - Quantify the Base-Line Risk in Terms of Frequency per Reactor Year ........................................................................................................ 5-3 5.2 Step 2 - Develop Plant-Specific Person-REM Dose (Population Dose) per Reactor Year .................................................................................... 5-8 5.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval From 10-to-15 Years ...................................................................................... 5-11 5.4 Step 4 - Determine the Change in Risk in Terms of Large Early Release Frequency ............................................................................................. 5-14 5.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability ............................................................................................. 5-14 5.6 Summary of Internal Events Results ..................................................... 5-15 5.7 ContributionS from Other Hazard Groups ............................................. 5-17 i C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE OF CONTENTS (cont'd)

Section Page 6.0 SENSITIVITIES ................................................................................................. 6-1 6.1 Sensitivity to Corrosion Impact Assumptions .......................................... 6-1 6.2 EPRI Expert Elicitation Sensitivity ........................................................... 6-3 6.3 DWBT Data Sensitivity............................................................................ 6-6

7.0 CONCLUSION

S ................................................................................................ 7-1

8.0 REFERENCES

.................................................................................................. 8-1 APPENDIX A PRA TECHNICAL ADEQUACY ............................................................A-1 ii C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 permanent extension of the Clinton Power Station (CPS) containment Type A integrated leak rate test (ILRT) interval from ten years to fifteen years. 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 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 [3]. Additionally, consistent with other previous ILRT extension requests for BWR Mark III containments, the risk assessment also includes an assessment for extending the Drywell Bypass Test (DWBT) interval from ten years to fifteen years. 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.

1.2 BACKGROUND

Revisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing requirements from three-in-ten years to at least once per ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage was less than the normal containment leakage of 1.0La (allowable leakage).

1-1 C0467100013-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The basis for a 10-year test interval is provided in Section 11.0 of NEI 94-01 [1],

Revision 0, and was established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493 [6],

Performance-Based Containment Leak Test Program, provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project Report TR-104285 [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 percent per day to 5 percent per day leads to a barely perceptible increase in total population exposure, and increasing the leak rate to 50 percent per day increases the total population exposure by less than 1 percent. 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 NRC Regulatory Guide 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/DWBT interval extension risk assessment for CPS employs the EPRI 1018243 1-2 C0467100013-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval methodology, with the affected System, Structure, or Component (SSC) being the primary containment boundary.

1.3 ACCEPTANCE CRITERIA The acceptance guidelines in RG 1.174 [4] are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in core damage frequency (CDF) less than 1.0E-06 per reactor year and increases in large early release frequency (LERF) less than 1.0E-07 per reactor year. Note that CDF is not impacted by the proposed change for CPS. Therefore, since the Type A test does not impact CDF for CPS, 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 C0467100013-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval (when compared against the baseline interval of 3 tests per 10 years), 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 permanently extending the ILRT/DWBT surveillance interval to 15 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 and the DWBT are not expected to 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. 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 percent 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 would require that the increase in CCFP be less than or equal to 1.5 percentage points.

1-4 C0467100013-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 2.0 METHEDOLOGY A simplified bounding analysis approach consistent with the EPRI methodology [3] is used for evaluating the permanent change in risk associated with increasing the test interval to fifteen 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/DWBT interval to fifteen 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 Conditional Containment Failure Probability (CCFP)

6. Evaluate the sensitivity of the results to assumptions in the steel corrosion analysis and to variations in the fractional contributions of large isolation failures (due to corrosion) to LERF.

Furthermore,

• 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 containment failure probability (CCFP) for defense-in-depth considerations, and change in LERF to demonstrate that the acceptance guidelines from RG 1.174 are met.

2-1 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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].

2-2 C467150095-12660-12/1/15

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

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

• The CPS Level 1 and Level 2 internal events PRA models provide representative core damage frequency and release category frequency distributions to be utilized in this analysis.

• It is appropriate to use the CPS internal events PRA model as a gauge to effectively describe the risk change attributable to the ILRT/DWBT extension. It is reasonable to assume that the impact from the ILRT/DWBT extension (with respect to percent increases in population dose) will not substantially differ if other hazard groups were to be included in the calculations; however, other hazard groups (e.g., internal fires, seismic) have been accounted for in the analysis based on the available information for CPS [8, 9] as described in Section 5.7.

• Dose results for the containment failures modeled in the 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.

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

[31] and the Illinois Department of Public Health (IDPH) [30] are appropriate for this analysis.

• The representative containment leakage for Class 1 sequences is 1 La.

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

• The representative containment leakage for Class 3a is 10 La 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 utilized 35La for the Class 3b sequences.

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

3-1 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

• 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.

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

• An evaluation of the risk impact of the ILRT on shutdown risk is addressed using the generic results from EPRI TR-105189 [12].

• The methodology to evaluate the impact of concurrently extending the DWBT interval is performed consistent with previous one-time ILRT/DWBT extensions for BWR Mark III containment types [19, 20, 21],

including CPS, which have been approved by the NRC.

3-2 C467150095-12660-12/1/15

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

• Section 4.1 General Resources Available as input

• Section 4.2 Plant Specific Resources Required

• Section 4.3 CPS Population Dose Derivation

• Section 4.4 Details on the EPRI Methodology that is followed

• Section 4.5 Details of the Calvert Cliffs corrosion analysis method that is also used as a sensitivity for this assessment

• 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.

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

• NUREG/CR-3539 [13]

• NUREG/CR-4220 [14]

• NUREG-1273 [15]

• NUREG/CR-4330 [16]

• EPRI TR-105189 [12]

• NUREG-1493 [6]

• EPRI TR-104285 [2]

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

• Calvert Cliffs corrosion analysis [5]

• EPRI 1018243 [3]

• NRC Final Safety Evaluation Report [7]

• Prior Mark III ILRT/DWBT Extension Risk Assessments [19, 20, 21]

4-1 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval NUREG/CR-3539 [13]

Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leak rates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 [22] as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

The study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and to be included in the model.

NUREG/CR-4220 [14]

NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985. The study reviewed over two thousand LERs, ILRT reports and other related records to calculate the unavailability of containment due to leakage. It assessed the large containment leak probability to be in the range of 1E-3 to 1E-2, with 5E-3 identified as the point estimate based on 4 events in 740 reactor years and conservatively assuming a one-year duration for each event.

The study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident.

NUREG-1273 [15]

A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREG/CR-4220 database. This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected. In addition, this study noted that local leak rate tests can detect essentially all potential degradations of the containment isolation system.

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval NUREG/CR-4330 [16]

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

NUREG/CR-4330 is a study that examined the risk impacts associated with increasing the allowable containment leakage rates. The details of this report have no direct impact on the modeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals. However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies:

the effect of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of containment.

EPRI TR-105189 [12]

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

The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessment because this EPRI study provides insight regarding the impact of containment testing on shutdown risk. This study performed a quantitative evaluation (using the EPRI ORAM software) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk.

The result of the study concluded that a small but measurable safety benefit (i.e.,

shutdown CDF reduced by 1.0E-8/yr to 1.0E-7/yr) is realized from extending the test intervals from 3 per 10 years to 1 per 10 years.

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

NUREG-1493 is the NRCs cost-benefit analysis for proposed alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The NRC conclusions are consistent with other similar containment leakage risk studies:

• Reduction in ILRT frequency from 3 per 10 years to 1 per 20 years results in an imperceptible increase in risk.

• Given the insensitivity of risk to the containment leak rate and the small fraction of leak paths detected solely by Type A testing, increasing the interval between integrated leak rate tests is possible with minimal impact on public risk.

EPRI TR-104285 [2]

Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study), the EPRI TR-104285 study is a quantitative evaluation of the impact of extending Integrated Leak Rate Test (ILRT) and (Local Leak Rate Test) LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with NUREG-1150 [17] Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 [6] in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-104285 used a simplified Containment Event Tree to subdivide representative core damage sequences into eight categories of containment response to a core damage accident:

1. Containment intact and isolated

2. Containment isolation failures due to support system or active failures

3. Type A (ILRT) related containment isolation failures

4. Type B (LLRT) related containment isolation failures

5. Type C (LLRT) related containment isolation failures

6. Other penetration related containment isolation failures

7. Containment failure due to core damage accident phenomena

8. Containment bypass 4-4 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Consistent with the other containment leakage risk assessment studies, this study concluded:

These study results show that the proposed CLRT [containment leak rate tests] frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms...

Release Category Definitions The EPRI methodology [2,3] defines accident classes that may be used in the ILRT/DWBT extension evaluation. These containment failure classifications, reproduced in Table 4.1-1, are used in this analysis to determine the risk impact of extending the Containment Type A ILRT and DWBT intervals as described in Section 5 of this report.

NUREG-1150 [23] and NUREG/CR 4551 [17]

NUREG-1150 and the technical basis, NUREG/CR-4551, provide an ex-plant consequence analysis for a spectrum of accidents including a severe accident with the containment remaining intact (i.e., Tech Spec leakage). This ex-plant consequence analysis is calculated for the 50-mile radial area surrounding Grand Gulf, another plant with a Mark III containment. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the CPS Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent CPS. (The meteorology and site differences other than population are assumed not to play a significant role in this evaluation.)

Calvert Cliffs Steel Corrosion Analysis [5]

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

4-5 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting steel liner corrosion, and the corresponding change in risk. The methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms was factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner. CPS also has a concrete cylinder and dome and a concrete basemat, each with a steel liner. The drywell is a Class 1 seismic structure and features reinforced concrete walls and floor in a vertical right cylinder geometry. The ceiling is also reinforced concrete with a removable steel dome known as the drywell head. The floor is common with the primary containment basemat. Additional details are contained in Table 4.1-2. The function of the drywell is to maintain a pressure boundary that forces steam from a loss of coolant accident (LOCA) through the 102 horizontal vents in the drywell wall into the suppression pool.

The corrosion analysis for Calvert Cliffs is used for the CPS containment vessel with slight variations made to account for the design differences.

EPRI 1018243 [3]

EPRI 1018243 presents a risk impact assessment for extending integrated leak rate test (ILRT) surveillance intervals to 15 years and provides the results of an expert elicitation process to determine the relationship between pre-existing containment leakage probability and magnitude.

EPRI 1018243 complements the previous EPRI report 104285 [2]. The earlier report considered changes to local leak rate testing intervals as well as changes to ILRT testing intervals. The original risk impact assessment [2] considers the change in risk based on population dose, whereas the revision [3] considers dose as well as large early release frequency (LERF) and conditional containment failure probability (CCFP).

This report deals with changes to ILRT testing intervals and is intended to provide bases for supporting changes to industry and regulatory guidance on ILRT surveillance intervals.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The risk impact assessment using the Jeffreys Non-Informative Prior statistical method is further supplemented with a sensitivity case using expert elicitation performed to address conservatisms. The expert elicitation is used to determine the relationship between pre-existing containment leakage probability and magnitude. The results of the expert elicitation process from this report are used as a separate sensitivity investigation for the CPS analysis presented here in Section 6.2.

NRC Safety Evaluation Report [7]

This SER documents the NRC staffs evaluation and acceptance of NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, subject to the limitations and conditions identified in the SER and summarized in Section 4.0 of the SER. These limitations (associated with the ILRT Type A tests) were addressed in the Revision 2-A of NEI 94-01 which are also included in Revision 3-A of NEI 94-01 [1] and the final version of the updated EPRI report [3], which was used for this application. Additionally, the SER clearly defined the acceptance criteria to be used in future Type A ILRT extension risk assessments as delineated previously in the end of Section 1.3.

Previous ILRT/DWBT Extension Risk Assessments for Mark III Plants [19, 20, 21]

Reference is made to other extension requests for Mark III containments that considered extensions to the ILRT interval and the DWBT interval.

Consistent with other previous ILRT extension requests for BWR Mark III containments, the risk assessment also includes an assessment for extending the Drywell Bypass Test (DWBT) interval from ten years to fifteen years. 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 4-7 C467150095-12660-12/1/15

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

The DWBT verifies that pre-existing drywell bypass leakage does not exceed the maximum allowed leakage. For CPS, the DWBT acceptance criterion in the Tech Spec SR 3.6.5.1.3 [34] is <10% of the analyzed design limit. The design bypass limit is used to establish the timing of automatic initiation of the containment spray system following a LOCA. If a leakage path were to exist between the drywell and the containment, the leaking steam would produce pressurization of the containment. To mitigate the consequences of any steam which bypasses the suppression pool, a high containment pressure signal will automatically initiate the containment spray system any time after LOCA + 10 minutes. The allowable bypass leakage is defined as the amount of steam which could bypass the suppression pool without exceeding the design containment pressure. The acceptable leakage rates are 10% of the maximum allowable leakage rate.

The Design Basis and Test Leakage criteria are found in Section 6.2 of the USAR [18].

The DWBT thus affects the likelihood of suppression pool bypass in the Level 1 and Level 2 PRA analyses.

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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval It is assumed in this augmented 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 containment has a different arrangement compared to BWR Mark I/II containments or PWR containments. The difference is that the 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 further discussed below, this leads to at least nine combinations of drywell and containment leakage sizes (refer to Figure 4.1-1). Each combination will have an impact on radionuclide releases that corresponds approximately to one of the original containment failure categories.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 4-10 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The different combinations of the drywell and containment leakage sizes can result in different accident classes. To address this issue, the Grand Gulf Nuclear Station (GGNS) and the Clinton Power Station (CPS) DWBT methods have some slight differences, which are discussed in more detail in the following two sub-sections.

Following that, the approach used for the updated CPS DWBT method is presented.

The Mark III and CPS plant-specific data utilized for the DWBT portion of this risk assessment is then provided in Section 4.6.

4.1.1 GGNS DWBT Method In the GGNS assessment [20], the assignment of each of these combinations to an original containment failure category depends on the consideration of the availability of the containment spray system. If containment sprays are available, the combination of drywell and containment leakage is categorized based on the containment leakage category. If containment sprays are not available, the combination of drywell and containment leakage is assumed to result in containment failure (Class 7) except for the combinations with base drywell bypass leakage. The combinations with base drywell leakage (DWLb) are assumed to have the same categories as the base case ILRT evaluation. Table 4.1-3 summarizes the classification of combinations into the EPRI accident classes used in the GGNS assessment.

The probability for each combination in Table 4.1-3 is determined by multiplying the conditional probabilities for DWBT and ILRT category by each other. For those cases where containment spray is a factor the probability of the combination of DWBT and ILRT is multiplied by the probability that containment spray is available or is not available as applicable.

The other change in the methodology to address the DWBT is the need to increase the containment failure due to phenomenology class (Class 7) frequency for the extended test frequencies. This is done in a manner similar to the method applied to Class 3a and 3b. That is, the Class 1 frequency is also adjusted downward for the Class 7 frequency increase in order to maintain the same total CDF.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 4.1.2 RBS DWBT Method For the most part, the GGNS methodology for DWBT extension evaluation was previously used for RBS [21]. The main modifications to the GGNS methodology were as follows:

• RBS credited the containment unit coolers to mitigate the adverse effects of the increased drywell leakages instead of the containment spray credited in the GGNS evaluation. Containment spray has dual functions by reducing the containment pressure and scrubbing the fission products from the containment atmosphere while containment unit coolers were designed mainly to reduce containment pressure. However, the GGNS method conservatively does not credit the containment spray for scrubbing. Thus the effects of crediting containment unit coolers and containment spray are the same.

• The RBS base cases for DWBT extension evaluation used EPRI Class 1 frequency to calculate the Class 3a, Class 3b and additional Class 7 frequencies. The GGNS method base cases used the total CDF for the calculation, which was conservative since more Class 1 frequency would be re-categorized into Class 3a, 3b or Class 7 frequencies. Such a conservative approach was not appropriate for the RBS evaluation since the RBS Class 1 frequency only consisted of about 10% of the total CDF, and as such, the calculated Class 3a, 3b and additional Class 7 frequencies could exceed the Class 1 frequency if using the total CDF for calculations. Therefore, only the CDF portion that does not lead to a more severe release category in the Level 2 analysis is re-categorized to Class 3a, 3b, or 7. This exclusion of a portion of the CDF that is impacted by the DWBT extension is similar to the allowed exclusion of LERF contributors per the accepted EPRI methodology for ILRT extension assessments.

4.1.3 CPS DWBT Method The CPS DWBT methodology [19] is very similar to that used for GGNS except the assignment of the nine combinations of drywell and containment leakage sizes to an original containment failure category. Unlike the GGNS approach which conservatively assumed all combinations without containment spray available would contribute to accident class 7, CPS used a Clinton specific MAAP 4.0 model to determine the effects of the increased drywell leakages that could lead to higher containment pressure. The MAAP run results along with other considerations found that containment failure induced by containment pressurization aggravated by the drywell bypass leakage 4-12 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 is considered a negligible contributor to CDF.

CPS assigned the equivalent EPRI category and LERF characterization as shown in Table 4.1-4. Note, these assignments are consistent with GGNS and RBS assignments with the exception that leakage combinations leading to EPRI Class 7 assignments are not included and CPS combination CA1 is conservatively classified as 3a EPRI release, consistent with the previous CPS ILRT extension analysis [19] while GGNS assigned this combination to EPRI class 1.

No credit for the availability of containment spray was taken in the CPS analyses.

Similar to GGNS method, 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. As previously noted GGNS and RBS conservatively assumed all combinations without containment spray (GGNS) or Unit Coolers (RBS) available would contribute to accident class 7. The assumed normal leakage rate DWLb is sufficiently low, that a 100DWLb drywell release would be a negligible contributor to accident class

7. See section 4.6 for additional detail.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-1 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 La, 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-2 CPS CONTAINMENT AND DRYWELL STRUCTURAL DESIGN FEATURES DESIGN FEATURE CPS CONTAINMENT DESIGN CPS DRYWELL DESIGN Design Pressure 15 psig 30 psig Construction Reinforced concrete with carbon steel liner. Reinforced concrete with Stainless steel in suppression pool area. steel plate covering internal surfaces of drywell cylinder and top slab.

Stainless steel in suppression pool area.

Drywell vents below suppression pool surface connect drywell and containment.

Liner Thickness Nominally 1/4 to 1/2 Nominally 1/2 Concrete Wall 3-0 walls, 2-6 dome, 9-8 base mat 5-0 walls, 6-0 top slab, Thickness 9-8 base mat Equipment Hatch Dished welded steel hatch bolted onto flange on Same as containment.

hatch barrel. Flange connection is double gasketed.

Personnel Airlocks 2 Personnel airlocks each with interlocked 1 Personnel airlock (same double doors. Constructed of welded steel. as containment).

Doors have double gasketed flanges.

Piping Penetrations Three types: Same as containment.

Type 1 for high-energy lines, guard pipes enclose these lines to direct the energy into the 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.

Other Mechanical Inclined Fuel Transfer Tube consists of a 3/4 thick Drywell head (24.7 feet carbon steel rolled plate pipe sleeve of 40 inside diameter elliptical dome diameter with a 36 standard flange on the with double seal) is bolted containment side. onto drywell head flange.

Serves as boundary between drywell and upper containment pools during non-refueling periods.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-2 CPS CONTAINMENT AND DRYWELL STRUCTURAL DESIGN FEATURES DESIGN FEATURE CPS CONTAINMENT DESIGN CPS DRYWELL DESIGN Electrical Penetrations Dual header plate type electrical penetrations are Cables penetrating the used. drywell wall pass through penetrations that are filled with sealant.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-3 GGNS [20] DWBT AND ILRT LEAKAGE COMBINATION ACCIDENT CLASSES EPRI LEAKAGE DW BYPASS CONTAINMENT CLASSIFICATION COMBINATIONS LEAKAGE LEAKAGE ASSIGNMENT AA 1 DWLb 1 La 1 AB 1 DWLb 10 La 3a (2)

AC 1 DWLb 35 La 3b BA1 CS Available 10 DWLb 1 La 1 (1)

BA2 CS Not Available CF CF 7 BB1 CS Available 10 DWLb 10 La 3a BB2 CS Not Available CF CF 7 BC1 CS Available 10 DWLb 35 La 3b BC2 CS Not Available CF CF 7 (2)

CA1 CS Available 35 DWLb 1 La 1 CA2 CS Not Available CF CF 7 CB1 CS Available 35 DWLb 10 La 3a CB2 CS Not Available CF CF 7 CC1 CS Available 35 DWLb 35 La 3b CC2 CS Not Available CF CF 7 Notes to Table 4.1-3:

(1)

CF = Containment failure assumed to occur.

(2)

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-4 CPS [19] DWBT AND ILRT LEAKAGE COMBINATION ACCIDENT CLASSES EPRI LEAKAGE DW BYPASS CONTAINMENT CLASSIFICATION COMBINATIONS LEAKAGE LEAKAGE ASSIGNMENT AA 1 DWLb 1 La 1 (Non-LERF)

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

(1)

AC 1 DWLb 35 La 3b (LERF)

BA1 10 DWLb 1 La 1 (Non-LERF)

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

BC1 10 DWLb 35 La 3b (LERF)

(1)

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

CB1 35 DWLb 10 La 3b (LERF)

CC1 35 DWLb 35 La 3b (LERF)

Note to Table 4.1-4:

(1)

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

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

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

• Population within a 50-mile radius

• Reactor Power Level [18]

• Allowable Containment Leakage [18]

CPS Internal Events Core Damage Frequency The current CPS Internal Events 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 Report for the 2014 PRA Interim Update [24], the mean value of the internal events core damage frequency (CDF) is 2.13E-06/yr. (truncation limit 5E-13/yr.). Core Damage Frequency by Class is provided in Table 4.2-1.

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 4.2-2 summarizes the pertinent CPS results in terms of release category, taken from Table 3.4-4 of the CPS PRA Summary Notebook [24]. The total Large Early Release Frequency (LERF) which corresponds to the H/E release category in Table 4.2-2 was calculated to be 1.16E-7/yr (truncation limit 5E-14). The total release frequency is 1.30E-06/yr., with a total CDF of 2.23E-06/yr (using a 5E-14 truncation limit.). This corresponds to an OK release (i.e.,

intact containment limited to normal leakage) of 9.30E-7/yr.

Based on the Level 1 and LERF PRA model results described above, Table 4.2-3 lists the relevant EPRI release category frequencies pertinent for the ILRT/DWBT extension risk assessment, including the delineation of LERF and non-LERF frequencies for Class 7.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The non-LERF frequencies for Class 7 are determined by reviewing the top 114 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. The phenomena-induced containment failures (non-LERF) sequences contribute 9.16E-07/yr. as shown in the Table 4.2-3.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.2-1 2014 CLINTON LEVEL 1 CDF RESULTS(3)

(1)

CLASS DESCRIPTION CDF (/YR)  % OF CDF IA/IC Loss of Makeup at 4.9E-07 22.9%

High RPV Pressure (Transient Initiators)

IBE Early Station 1.4E-07 6.6%

Blackout (less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)

IBL Late Station 6.5E-07 30.4%

Blackout (greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)

ID Loss of Makeup at 1.0E-07 4.7%

Low RPV Pressure (Transient Initiators)

II Loss of Containment 4.2E-07 19.6%

Heat Removal IIIA Excessive LOCA 1.0E-09 <0.1%

IIIB Loss of Makeup at 3.7E-08 1.7%

High RPV Pressure (LOCA Initiators)

IIIC Loss of Makeup at 8.5E-08 4.0%

Low RPV Pressure (LOCA Initiators)

IIID Loss of Vapor 8.3E-09 0.4%

Suppression (LOCA Initiators)

(2)

IV Loss of Adequate 2.0E-07 9.6%

Reactivity Control (ATWS)

V Containment Bypass 1.6E-09 0.1%

Total - 2.1E-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)

Class IVL included in Class IV.

(3)

From Table 7.3-1 of The CPS PRA Level 2 Notebook [32].

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.2-2

SUMMARY

OF CONTAINMENT EVALUATION(3)

INPUT OUTPUT LEVEL 1 PSA CET EVALUATION RELEASE CORE DAMAGE CHARACTERIZE FREQUENCY FREQUENCY RELEASE RELEASE BIN (PER YEAR)

(1) 2.23E-6/year Little or No Release OK 9.30E-7

@ 5E-14 truncation (Intact)

(2)

LL & Late 2.00E-8 LL & I Low Public LL & E 1.16E-09 Risk Impact (2)

L & Late 9.88E-8 L&I L&E 2.17E-7 (2)

Moderate M & Late 2.02E-7 Release M&I M&E 2.00E-7 (2)

H & Late 4.45E-7 High Release H&I H&E 1.16E-7 4-22 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Notes to Table 4.2-2:

(1)

Level 1 CDF result when using a quantification truncation limit of 5E-14/yr. This result is only provided when comparing to the Level 2 results.

(2)

Sum of release frequencies in right column (not including Release Bin OK) is 1.30E-06/yr.

(3)

From Table 7.2-1 of The CPS PRA Level 2 Notebook [32].

(4)

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 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sy, 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 />.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.2-3 RELEVANT LEVEL 2 RELEASE CATEGORY FREQUENCIES FOR CPS EPRI RELEASE CATEGORY FREQUENCY/YR SOURCE 1: No Containment Failure 9.30E-07 Table 4.2-2 CDF contribution of sequences (top 99%) that (1) included failure of the 2: Containment Isolation Failure 2.68E-07 containment isolation function (Event Tree Node IS=F)

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

(LERF) 1.14E-07 Containment Bypass (Class V)

Table 4.2-2 and sequence evaluation 2.23E-06 CDF -

7: Phenomena-induced containment failures - 9.30E-07 (non-LERF) OK (EPRI Class 1) -

- 1.16E-07 LERF (EPRI

- 2.68E-07 Class 7 LERF & 8) -

= 9.16E-07 Cont. Isolation (EPRI Class 2)

Class V 8: Containment Bypass 1.55E-09 (ISLOCA + BOC Sequences)

Total: 2.23E-06 Note to Table 4.2-3:

(1)

Not all Containment Isolation Failure sequences are found to be H/E (LERF) in the CPS PRA based on MAAP calculations. Never-the-less, 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 affects just the total dose and % change metrics.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 utilized. From the EPRI guidance it is noted that 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]. 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.

• 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.

• 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/DWBT interval extension submittal [19].

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

COLLAPSED APB NUMBER DESCRIPTION 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 4-27 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.3-2 GRAND GULF NUREG/CR-4551 [17] 50-MILE RADIUS POPULATION DOSE APB FRACTIONAL APB 50-MILE APB CONTRIBUTION RADIUS DOSE APB 50-MILE FREQUENCY TO 50-MILE RISK RADIUS DOSE APB APB (PER RADIUS TOTAL (PERSON- (PERSON-(1) (2) (3) (4) (5)

1. DEFINITION YEAR) DOSE RISK REM/YEAR) REM)

CD, VB, Early 6.46E-7 .268 0.139 2.15E+5 1 CF, Early SP Bypass, CS Not Available CD, VB, Early 2.00E-7 .056 0.029 1.45E+5 2 CF, Early SP Bypass, CS Available CD, VB, Early 2.86E-8 .011 5.7E-3 1.99E+5 3 CF, Late SP Bypass CD, VB, Early 8.92E-7 .267 0.139 1.56E+5 4 CF, 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 4-29 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The APBs described above can then be 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/DWBT frequency and are therefore (per the EPRI guidance) not included in the assignment process.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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) 4-31 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 substantial differences in reactor power level, population density, 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 utilizes census data to calculate population counts for user defined sector segments. The EPRI methodology does not specifically specify population projection to a future date. For this risk assessment the CPS population was projected to the year 2030 to represent the average population density over the next two 15 year extension intervals.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.3-5 SECPOP CODE POPULATION ESTIMATES SECPOP (1)(2) (1)(2)

RADIUS 2000 2010 2020 2030 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 (1)

The 2020 and 2030 estimates are made assuming the 5.79%

population increase experienced in the 50-mile radius region during the decade from July 2000 to July 2010 continues to occur each of the next two decades.

(2)

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 4.3-5b 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). See map and table below.

Illinois Counties Within 50 Miles of CPS Shown in Map Above 4-33 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.3-5B IDPH [30] POPULATION PROJECTION ESTIMATES AREA WITHIN 2020 2025 50 MILE 2010 POPULATION POPULATION COUNTY RADIUS POPULATION ESTIMATE ESTIMATE Champaign 99% 201,370 217,735 225,626 Christian 40% 34,804 33,152 32,345 Coles 20% 53,945 56,851 58,405 De Witt 100% 16,583 15,832 15,495 Douglas 75% 19,976 19,767 19,709 Ford 75% 14,074 13,450 13,244 Iroquois <5% 29,657 27,687 26,816 Livingston 40% 38,882 39,390 39,596 Logan 100% 30,272 30,380 30,441 Macon 100% 110,757 105,401 103,126 Mason 25% 14,627 12,841 12,074 McLean 100% 169,838 188,341 197,855 Menard 40% 12,708 12,867 12,913 Moultrie 90% 14,846 14,715 14,706 Piatt 100% 16,722 16,205 16,000 Sangamon 35% 197,822 203,501 207,194 Shelby 25% 22,339 21,496 21,118 Tazewell 75% 135,439 136,051 136,436 Vermillion <5% 81,588 77,965 76,441 Woodford 50% 38,664 40,350 41,360 (1)(2)

Total -- 1,255,013 -- 1,300,900 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%

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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:

• Grand Gulf Containment Volume [29] = 1.40E+6 ft3

• 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:

(0.73%

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 (1)

APB # REM) 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 4.2-3 for each relevant EPRI release category class from Table 4.1-1, and then 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 4-36 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 3 frequencies are derived later and the Class 4, 5, and 6 frequencies are not utilized in the EPRI methodology for the ILRT extension risk assessment).

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 class which differs 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 the top 114 sequences of the Clinton Level 2 model (which provide more than 99.0% of Level 2 release frequency) and to assign each sequence into one of the collapsed Accident Progression Bins (APBs) from NUREG/CR-4551. The CPS Level 2 model (i.e., containment event tree structure) contains a significantly larger amount of information about the accident sequences than what is used 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 utilized (or did not contribute) to the APB assignment performed here for the ILRT assessment. Additionally, it should be noted that 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 Accident Progression Bins from NUREG/CR-4551 as described in Section 4.2. The Class 7 Sequences are divided into two categories.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Class 7 LERF is defined to consist of LERF sequences (excluding the BOC, 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 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 1 CX - No Energetic (highest dose) is assigned.

Containment Failure DL - Drywell If the drywell is not isolated or Suppression Pool Scrubbing fails, it is Isolation assumed that an un-scrubbed release to containment occurs as SP - Suppression soon as the vessel is breached. If Containment Spray (CS) fails the Pool scrubbing 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 to Pool Failure Above 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 assigned to APB 5.

Isolation For Class IIE sequences with late GE declaration, containment SP - Suppression failure is assumed to occur in the early time frame. If the drywell is Pool scrubbing 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.

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

POPULATION DOSE RISK (50 RELEASE POPULATION MILES)

ACCIDENT CLASS FREQUENCY / DOSE (50 MILES) (PERSON-REM /

(APB NUMBER) YR PERSON-REM (1) YR) (2) 7 LERF (APB 1) 1.14E-07 5.48E+05 6.27E-02 7 non-LERF 7a (APB 1) 4.72E-08 5.48E+05 2.59E-02 7b (APB 3) 2.29E-09 5.08E+05 1.16E-03 7c (APB 4) 3.35E-08 3.97E+05 1.33E-02 7d (APB 5) 6.27E-07 3.21E+05 2.01E-01 7e (APB 6) 2.27E-07 3.34E+05 7.57E-02 7f (APB 8) 7.14E-09 1.38E+05 9.88E-04 (4)

Class 7 non-LERF Total 9.44E-07 3.37E+05 3.18E-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.

(3)

Total non-LERF release frequency is shown as 9.16E-07 in Table 4.3-9. Release frequency above used a summing of CAFTA sequences contributions using Fussell-Vesely contributions from a cutset report. Sequence tagging in the recovery led to a small percentage of cutsets being double counted (counted as contributing to two sequence endstates). URE CL2015-014 is tracking resolution. This issue has no impact to CDF/LERF values. Table 4.2-9 calculation used a different methodology to estimate 7 non-LERF contributions. Negligible impact to ILRT/DWBT results.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 9.30E-07 2.71E+03 2.52E-03 2: Containment Isolation Failure 2.68E-07 5.48E+05 1.47E-01 7: Phenomena-induced 1.14E-07 5.48E+05 6.27E-02 containment failures (LERF) 7: Phenomena-induced 9.16E-07 3.37E+05 3.09E-01 containment failures (non-LERF) 8: Containment Bypass 1.55E-09 5.48E+05 8.49E-04 Total: 2.23E-06 5.22E-01 4-42 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 test 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 latest 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 latest available EPRI data [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 delta 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 delta LERF. EPRI describes ways to demonstrate that, using plant-specific calculations, the delta 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 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 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), and the average time that a leak could exist without detection for a ten-year interval is 5 years (10 yr / 2). This interval change would lead to a non-detection probability that is a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRT testing, given a 10-year vs. a 3-yr (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 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.

4-44 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval interval. Correspondingly, an extension of the ILRT interval to fifteen years can be estimated to lead to about a factor of 5.0 (7.5/1.5) increase in the non-detection probability of a leak.

CPS Past ILRT Results The surveillance frequency for Type A testing in NEI 94-01 [1] under option B criteria is at least once per ten years based on an acceptable performance history (i.e., two consecutive periodic Type A tests at least 24 months apart) where the calculated performance leakage rate was less than 1.0La, and in compliance with the performance factors in NEI 94-01, Section 11.3. Based on the successful completion of two (1) consecutive ILRTs at CPS , the current ILRT interval is once per ten years. Note that the probability of a pre-existing leakage due to extending the ILRT interval is based on the industry-wide historical results as noted in the EPRI guidance document [3].

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/DWBT interval to fifteen 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].

(1)

Per email from F. Sarantakos, Appendix J Program Engineer, the previous ILRTs were performed in 1993 and 2008 and both ILRTs met Tech Spec requirements (Leakage rates were below TS allowed leakage rate). DWBTs were also performed in 1993 and 2008 and also met TS requirements.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

5. Determine the impact on the Conditional Containment Failure Probability (CCFP).

6. Evaluate the sensitivity of the results to assumptions in the steel corrosion analysis and to variations in the fractional contributions of large isolation failures (due to corrosion) to LERF.

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 conditional containment failure probability (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.

4.5 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.

4-46 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 would not be 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 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 4-47 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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:

• Differences between the containment basemat and the containment cylinder and dome

• The historical flaw likelihood due to concealed corrosion

• The impact of aging

• The corrosion leakage dependency on containment pressure

• The likelihood that visual inspections will be effective at detecting a flaw Assumptions

• Consistent with the Calvert Cliffs analysis, a half failure is assumed for the basemat concealed liner corrosion due to lack of identified failures (see Table 4.5-1, Step 1).

• 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 4-48 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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).

• 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). Sensitivity studies are included that address doubling this rate every two years and every ten years.

• 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. Sensitivity studies in Section 6 are included that increase and decrease the probabilities by an order of magnitude (See Table 4.5-1, Step 4).

• Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the basemat region is considered to be less likely than the containment walls. (See Table 4.5-1, Step 4.)

• 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 very 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. Sensitivity studies are included in Section 6 that evaluate total detection failure likelihoods as low as 5%

and as high as 15%, respectively.

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 2/(70

• 5.5) = 5.2E-3 0.5/(70

• 5.5) = 1.3E-3 with 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 avg 5-10 5.2E-3 avg 5-10 1.3E-3 failure rate doubles every five years (14.9% increase per 15 1.4E-2 15 3.5E-3 year). The average for 5th to 10th year is set to the historical 15 year average = 6.27E-3 15 year average = 1.57E-3 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 4.06% (1 to 10 years) 1.02% (1 to 10 years)

Uses age adjusted flaw 9.40% (1 to 15 years) 2.35% (1 to 15 years) likelihood (Step 2), assuming failure rate doubles every five (Note that the Calvert Cliffs (Note that the Calvert Cliffs years (consistent with Calvert analysis presents the delta analysis presents the delta Cliffs analysis - See Table 6 of between 3 and 15 years of between 3 and 15 years of Reference [5]). 8.7% to utilize in the estimation 2.2% to utilize in the of the delta-LERF value. For estimation of the delta-LERF this analysis, the values are value. For this analysis, calculated based on the 3, 10, however, values are and 15 year intervals.) calculated based on the 3, 10, and 15 year intervals.)

4 Likelihood of Breach in 1% 0.1%

Containment Given Steel Flaw The failure probability of the containment cylinder and dome is assumed to be 1%

(compared to 1.1% in the Calvert Cliffs analysis). The basemat failure probability is assumed to be a factor of ten less, 0.1% (compared to 0.11%

in the Calvert Cliffs analysis).

4-51 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.5-1 STEEL CORROSION BASE CASE CONTAINMENT CYLINDER STEP DESCRIPTION AND DOME CONTAINMENT BASEMAT 5 Visual Inspection Detection 10% 100%

Failure Likelihood 5% failure to identify visual Cannot be visually inspected.

Utilize assumptions consistent flaws plus 5% likelihood that with 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.

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) 0.00406% (at 10 years) 0.00102% (at 10 years)

=4.06%

• 1%

• 10% =1.02%

• 0.1%

• 100%

0.0094% (at 15 years) 0.00235% (at 15 years)

=9.40%

• 1%

• 10% =2.35%

• 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 10 years: 0.00406% + 0.00102% = 0.00508%

At 15 years: 0.0094% + 0.00235% = 0.0118%

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 4.6 IMPACT OF DWBT INTERVAL EXTENSION OF RELEASE CATEGORIES Similar to the prior CPS ILRT/DWBT interval extension risk assessment, Table 4.6-1 provides the release categories that are utilized in this assessment for the different combinations of drywell bypass leakage and containment leakages. These classifications are consistent with the CPS ILRT/DWBT risk assessment [19] except for using the multiplier of 100 (per the updated EPRI guidance) rather than a multiplier of 35.

TABLE 4.6-1 CPS DWBT AND ILRT LEAKAGE COMBINATION ACCIDENT CLASSES EPRI LEAKAGE DW BYPASS CONTAINMENT CLASSIFICATION COMBINATIONS LEAKAGE LEAKAGE ASSIGNMENT AA 1 DWLb 1 La 1 AB 1 DWLb 10 La 3a AC 1 DWLb 100 La 3b BA1 10 DWLb 1 La 1 BB1 10 DWLb 10 La 3a BC1 10 DWLb 100 La 3b CA1 100 DWLb 1 La 3a CB1 100 DWLb 10 La 3b CC1 100 DWLb 100 La 3b Note to Table 4.6-1:

(1)

CF = Containment failure assumed to occur.

Again, 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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.

Clinton Continuous Monitoring Capability Clinton has the ability to continuously monitor DW leakage. The latest drywell leakage test, performed in 2008, found drywell leakage to be 20.18 psi at 3 psi (2). Leakage during operation is less as differential pressure between the drywell and containment is

< 1 psid. However, small airline leaks cause the drywell to pressurize at a rate of approximately 0.03 psi/hr and operators must vent the drywell approximately once per (1)

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

(2)

Drywell Leakage provided by T. Hable, Clinton SRME to J. Steinmetz via email on 10/07/2015.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval day.(1) The following paragraphs are from the 2003 ILRT 15 year one time LAR submittal [19]. This description below continues to reflect the current operating practices and drywell performance:

Due to the demonstrated leaktight performance of the drywell, CPS is able to monitor the integrity of the drywell during normal plant operation. This is possible due to the normal operation of pneumatic controls and operators in the drywell that pressurize the drywell, plus the existence of small instrument air system leaks. These effects create a differential pressure between the drywell and primary containment that is monitored, and periodic operation action is required to vent the drywell.

For example, in 1994, the drywell was being pressurized at a rate of approximately 0.04 psi/hr. The drywell was being vented approximately once per day when pressure approached the upper TS limit of 1.0 psid.

Based on application of the ideal gas law and known data, such as the drywell pressurization rate and the drywell leakage measured during the fourth refueling outage (RF-4), the total amount of instrument air in-leakage was calculated to be between 21.5 and 22.5 scfm. The rate of drywell pressurization remained essentially constant since drywell closeout from RF-4. Pressurization rates following subsequent refuelings have also remained consistent with those observed following RF-4.

This steady drywell pressurization rate allows qualitative monitoring of the drywell leakage rate. An increase in this rate would be indication of an increase in the instrument air system leakage into the drywell since it is improbable that the drywell would become more leaktight. Conversely, a decrease in this rate would be evidence of a larger drywell leakage area.

The maximum drywell leakage rate that would still maintain a differential pressure between the drywell and wetwell must be less than the instrument air in-leakage rate (which after RF-4 was 23 scfm). The A/k for a 23 scfm leak at 0.2 psid is 0.0025 ft2 or 0.2% of the allowable leakage area.

Because of this large margin to the allowable drywell leakage rate, it has been concluded that as long as the drywell continues to pressurize, regardless of the rate, drywell integrity is always assured. This ability to qualitatively assess the integrity of the drywell during normal plant operation provides further support to extending the DWBT interval.

In order to provide added assurance that the drywell has not seriously degraded between the performances of DWBTs, a qualitative assessment of the drywell leak tightness is performed at least once per operating cycle.

The first assessment was performed prior to Operating Cycle 7. By (1)

The pressurization rate and frequency of venting are derived from graphs found in the 2009 to 2015 periodic drywell leakage assessments documented under Clinton AR 400694 to meet programmatic commitments found in NRC approval of License Amendment 160 [11] and License Request to extend Drywell testing to 10 years [10].

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval checking for gross leakage, this assessment will provide an indication of the ability of the drywell to perform its design function. As a check for gross leakage, the assessment may not identify drywell leakage that is masked by plant conditions, or identify leakage through systems that are not communicating with the drywell atmosphere at the time of the assessment.

For example, minor increases in drywell bypass leakage could be masked by a small leak in the instrument air system inside the drywell. The assessment is not detailed enough to account for such minor changes.

However, as demonstrated above, as long as the drywell continues to pressurize, regardless of the rate, drywell integrity is always assured.

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

[27] and updated with the latest (Feb-08) Clinton Power Station 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.

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 4-56 C467150095-12660-12/1/15

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

SITE TEST DATE (SCFM) 300 SCFM 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 then 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 6

• 1 DWLb and all test data is below 10 DWLb. 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.

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 4-57 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval These represent very 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).

FIGURE 4.6-1 MARK III DWBT RESULTS COMPARED TO 300 SCFM X Axis = Test # from Table 4.6-2 Y Axis = Test Leakage ÷ 300 SCFM (1)

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

• 10 = 36,540 scfm) 4-58 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Only 2 of the data points are measurably in the 6-10 range, and all are below the 10 Lb 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 one-time 15 year ILRT LAR [19]. 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 NEI/EPRI approach is reflective of drywell failures. The body of plant experience used considered the older BWR containment designs.

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

• 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.

• 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.

• 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.

4-59 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.6-3 DW AND WW LEAKAGE PROBABILITIES DW DW WW WW LEAKAGE LEAKAGE PROBABILITY LEAKAGE LEAKAGE SIZE (LB) SIZE (LA) PROBABILITY (BASE)

(1) (1) 1Lb 0.9885 1La 0.9885 10Lb 0.0092 10La 0.0092 100Lb 0.0023 100La 0.0023 (1)

The probability of assumed normal drywell leakage (1La) is

[1-(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:

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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) 0.99 1 0.0115 3a 0.0023 3b A sensitivity case increasing the probability of a small (10DWLb) and large (100DWLb) leakage rates by a factor of 10 is used in the assessment as described in Section 6.3.

This sensitivity case is not considered representative of Clinton because Clinton has a daily check on the drywell integrity via the observed daily pressurization of the drywell.

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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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.

The assessment documented the following:

• 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.

• The pressure suppression capability of the containment is robust.

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

• 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 very high.

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The pressurization issue was addressed in the Safety Evaluation Report (SER) that was part of the NRC letter [11] approving the onetime 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, 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, the following conclusion is reached for the request for a permanent 15 interval risk assessment:

There is no change in CDF due to the small increases in drywell bypass leakage associated with the DWBT interval extension.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 5.0 RESULTS The application of the approach based on EPRI Guidance [3] has led 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 performed 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:

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

• 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).

• 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 5-1 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Guidance, this class is not specifically examined since it will not significantly influence the results of this analysis.

• 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.

• 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 risk impact of extending Type A test interval from 3 to 15 and 10 to 15 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 Conditional Containment Failure Probability (CCFP).

Step 6 Evaluate the sensitivity of the results to assumptions in the steel corrosion analysis and to variations in the fractional contributions of large isolation failures (due to corrosion) to LERF.

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 5.1 STEP 1 - QUANTIFY THE BASE-LINE RISK IN TERMS OF FREQUENCY PER REACTOR YEAR The CPS PRA Level 2 Model [25] 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.

For the assessment of the impact on the risk profile due to the ILRT/DWBT 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 1.55E-09/yr, as documented in Table 4.2-3.

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.

5-3 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 is 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 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.

5-4 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Class 2 Sequences This group consists of large containment isolation failures. The failure frequency for Class 2 sequences is 2.68E-07/yr, as documented in Table 4.2-3. Note that this frequency is not affected by the ILRT/DWBT 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 for relatively large flaws, consistent with the EPRI methodology [3].

The respective frequencies per year are determined as follows:

PROBClass_3a = probability of small pre-existing containment leakage

= 0.0092 (see Section 4.4)

PROBClass_3b = probability of large pre-existing containment leakage

= 0.0023 (see Section 4.4)

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 * [2.23E (1.14E-07 + 1.60E-09 + 4.20E-07 + 6.50E-07)]

= 1.20E-08/yr 5-5 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Class_3b = 0.0023 * [CDF - (EPRI Class 7 LERF + EPRI Class 8 + Class II +

Class IBL)]

= 0.0023 * [2.23E (1.14E-07 + 1.60E-09 + 4.20E-07 + 6.50E-07)]

= 2.42E-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 latest EPRI methodology [3]. The probability of Class 3a and Class 3b leakages are combined leakage probabilities from Table 4.6-4.

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 is9.16E-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)

= 2.23E (2.68E-07 (class 2) + 1.20E-08 (3a) + 2.42E-09 (3b) + 1.14E-07 (7 LERF) + 9.16E-07 (7-Non-LERF) + 1.55E-09 (Class 8))

= 9.16E-07/yr 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 9.16E-07 2 Large Isolation Failures (Failure to Close) 2.68E-07 3a Small Isolation Failures 1.20E-08 3b Large Isolation Failures 2.42E-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) 1.14E-07 7 non-LERF Failures Induced by Phenomena (non-LERF) 9.16E-07 8 Containment Bypass 1.55E-09 CDF All CET End states (including intact case) 2.23E-06 5-7 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 = 3.37E+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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 (1 La) 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 5.48E+05 (LERF) 7 non-LERF Failures Induced by Phenomena (non-3.37E+05 LERF) 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.

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

EPRI METHODOLOGY CORROSION ACCIDENT PERSON- CHANGE DUE TO CLASSES REM PERSON- PERSON- CORROSION (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 9.16E-07 2.49E-03 9.16E-07 2.49E-03 -2.52E-08 2 Large Isolation Failures 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 1.20E-08 3.25E-04 1.20E-08 3.25E-04 --

3b Large Isolation Failures 2.71E+05 2.42E-09 6.58E-04 2.43E-09 6.60E-04 2.52E-06 7 LERF Failures Induced by 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 --

CDF All CET end states 2.23E-06 0.523 2.23E-06 0.523 2.49E-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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 5.3 STEP 3 - EVALUATE RISK IMPACT OF EXTENDING TYPE A TEST INTERVAL FROM 10-TO-15 YEARS The next step is to evaluate the risk impact of extending the test interval from its current ten-year value to fifteen-years. To do this, an evaluation must first be made of the risk associated with the ten-year interval since the base case applies to a 3-year interval (i.e., a simplified representation of a 3-in-10 year interval).

Risk Impact Due to 10-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). However, as noted previously, the DWBT tests also impact the Class 7 sequences. Thus, the frequency of Class 3a, 3b, and Class 7 sequences are impacted by the ILRT/DWBT interval extension. The risk contribution is changed based on the EPRI guidance as described in Section 4.4 by a factor of 3.33 compared to the base case values. The results of the calculation for a 10-year interval are presented in Table 5.3-1.

Risk Impact Due to 15-Year Test Interval The risk contribution for a 15-year interval is calculated in a manner similar to the 10-year interval. The difference is in the increase in probability of not detecting a leak in Classes 3a and 3b for the ILRT Type A tests, and for Class 7 for the DWBT tests. For this case, the value used in the analysis is a factor of 5.0 compared to the 3-year interval value, as described in Section 4.4. The results for this calculation are presented in Table 5.3-2.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 5.3-1 CPS ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT/DWBT FREQUENCY EPRI METHODOLOGY PLUS CHANGE DUE TO (2)

EPRI METHODOLOGY CORROSION CORROSION OR ACCIDENT PERSON- DWBT 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 8.82E-07 2.39E-03 8.82E-07 2.39E-03 -1.44E-07 2 Large Isolation Failures 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 3.98E-08 1.08E-03 3.98E-08 1.08E-03 --

3b Large Isolation Failures 2.71E+05 8.07E-09 2.19E-03 8.12E-09 2.20E-03 1.44E-05 7 LERF Failures Induced by 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 --

CDF All CET end states 2.23E-06 0.525 2.23E-06 0.525 1.43E-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/DWBT Interval TABLE 5.3-2 CPS ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT/DWBT FREQUENCY EPRI METHODOLOGY PLUS CHANGE DUE TO (2)

EPRI METHODOLOGY CORROSION CORROSION OR ACCIDENT PERSON- DWBT 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 8.58E-07 2.33E-03 8.58E-07 2.33E-03 -3.35E-07 2 Large Isolation Failures 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 5.98E-08 1.62E-03 5.98E-08 1.62E-03 --

3b Large Isolation Failures 2.71E+05 1.21E-08 3.29E-03 1.22E-08 3.32E-03 3.35E-05 7 LERF Failures Induced by 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 --

CDF All CET end states 2.23E-06 0.526 2.23E-06 0.526 3.32E-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/DWBT 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 core damage frequency (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/DWBT 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 the original 3-in-10 year test interval assessment from Table 5.2-2, the Class 3b frequency is 2.43E-09/yr, which includes the corrosion effect of containment. Based on a ten-year test interval from Table 5.3-1, the Class 3b frequency is 8.12E-09/yr; and, based on a fifteen-year test interval from Table 5.3-2, it is 1.22E-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 3 to 15 years (including corrosion effects) is 9.81E-09/yr.

Similarly, the increase due to increasing the interval from 10 to 15 years (including corrosion effects) is 4.12E-09/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 year results to the original 3-in-10 year requirement.

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 conditional containment failure probability (CCFP). The change in CCFP is indicative of the effect of the ILRT/DWBT 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 5-14 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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).

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 CCFP CCFP15-3 CCFP15-10 3 IN 10 YRS 1 IN 10 YRS 1 IN 15 YRS 58.41% 58.66% 58.84% 0.44% 0.18%

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

The change in CCFP of approximately 0.5% as a result of extending the test interval to 15 years from the original 3-in-10 year requirement is judged to be relatively 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 3-in-10 years to 1-in-15 years all are below the acceptance criteria defined in Section 1.3, namely:

1. Change in LERF = 9.81E-9/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 3.80E-3 person-rem/yr (0.73%), which is less than 1.0 person-rem/year or 1% of the total population dose.

3. Change in CCFP is 0.44%, which is less than 1.5%.

5-15 C467150095-12660-12/1/15

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

BASE, 3 TO 10, AND 3 TO 15 YR EXTENSIONS (INCLUDING AGE ADJUSTED STEEL CORROSION LIKELIHOOD)

BASE CASE EXTEND TO EXTEND TO 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARS EPRI DOSE PERSON- PERSON- PERSON-CLASS PER-REM CDF (1/YR) REM/YR CDF (1/YR) REM/YR CDF (1/YR) REM/YR 1 2.71E+03 9.16E-07 2.49E-03 8.82E-07 2.39E-03 8.58E-07 2.33E-03 2 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 2.68E-07 1.47E-01 3a 2.71E+04 1.20E-08 3.25E-04 3.98E-08 1.08E-03 5.98E-08 1.62E-03 3b 2.71E+05 2.43E-09 6.60E-04 8.12E-09 2.20E-03 1.22E-08 3.32E-03 7 LERF 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 1.14E-07 6.27E-02 7 non-LERF 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 9.16E-07 3.09E-01 8 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 1.55E-09 8.49E-04 Total 2.23E-06 0.523 2.23E-06 0.525 2.23E-06 0.526 ILRT Dose Rate from 3a 9.85E-04 3.29E-03 4.95E-03 and 3b Delta From 3 yr --- 2.21E-03 3.80E-03 Total Dose From 10 yr --- --- 1.59E-03 Rate (1) 3b Frequency (LERF) 2.43E-09 8.12E-09 1.22E-08 Delta 3b From 3 yr --- 5.69E-09 9.81E-09 LERF From 10 yr --- --- 4.12E-09 CCFP % 58.41% 58.66% 58.84%

Delta From 3 yr --- 0.26% 0.44%

CCFP %

From 10 yr --- --- 0.18%

Note to Table 5.6-1:

1. The overall difference in total dose rate is less than the difference of only the 3a and 3b categories between two testing intervals. This is because the overall total dose rate includes contributions from other categories that do not change as a function of time, e.g., the EPRI Class 2 and 8 categories, and also due to the fact that the Class 1 person-rem/yr decreases when extending the IRLT/DWBT frequency.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 5.7 CONTRIBUTIONS FROM OTHER HAZARD GROUPS 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. Appendix A provides a technical adequacy assessment of the Seismic Core Damage Frequency (SCDF) and the Fire PRA. For this ILRT/DWBT risk assessment, contributions from other hazard groups are addressed using a CDF multiplier approach. This approach has been used in previous ILRT submittals [28, 39]

Internal Fire Risk [8]

CPS has a Fire PRA model that was updated in 2014. The total Fire PRA CDF reported is 6E-06/yr [8], which is a factor of 1.8 higher than the Fire CDF (i.e., 3.26E-06/yr) calculated in the CPS Individual Plant Examination of External Events (IPEEE)

[37]. The Fire PRA LERF reported is 9.21E-07/yr. In addition to modeling limitations, the Fire PRA may be subject to more modeling uncertainty than the internal events PRA evaluations. While the 2014 Fire PRA is based on the guidance of NUREG/CR-6850

[38] and is generally self-consistent within its calculational framework, the Fire PRA CDF results do not compare well with internal events PRAs because of the number of conservative assumptions that have been included in the Fire PRA process. Therefore, direct use of the Fire PRA results as a reflection of CDF may be inappropriate, and the actual fire CDF may be overestimated. In any event, the reported Fire CDF value from the 2014 Fire PRA Update is used as a bounding value for this calculation.

Seismic Risk [9]

A quantifiable seismic PRA model for Clinton has not yet been approved for general use in risk applications. However, a Clinton seismic risk analysis was performed as part of the CPS IPEEE [37]. Clinton performed a seismic margins assessment (SMA) following the guidance of NUREG-1407 [35] and EPRI NP-6041 [36]. The SMA is a deterministic evaluation process that does not calculate risk on a probabilistic basis. No core 5-17 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval damage frequency sequences were quantified as part of the IPEEE seismic risk evaluation.

The conclusions of the Clinton IPEEE seismic risk analysis are as follows:

No improvements to the plant were identified as a result of the Seismic Margins Assessment the plant was determined to be fully capable of attaining safe shutdown conditions after the Review Level Earthquake (RLE).

However, more recent information is available from the NRC. A Risk Assessment for NRC GI-199 Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States (CEUS) on Existing Plants, [9], Table D-1 lists the postulated core damage frequencies using the updated 2008 USGS Seismic Hazard Curves. For Clinton, two reference sources are provided for the seismic hazard (i.e.

NUREG/CR-0098 and UHS). A seismic CDF value of 1.7E-5/yr is selected for this ILRT/DWBT risk assessment because it is the higher identified value (i.e. Clinton (UHS)) in Reference [9]. The NRC study did not calculate LERF.

Other External Events In addition to internal fires and seismic events, the CPS IPEEE Submittal analyzed a variety of other external hazards:

• High Winds/Tornadoes

• External Flooding

• Transportation and Nearby Facility Accidents

• 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 5-18 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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/DWBT interval extension risk assessment.

Other Hazard Group Contributor Summary The method chosen to account for external events contributions is similar to that used in the other ILRT interval extension analyses [28, 39] in which a multiplier is applied to the internal events results. The contributions of the external events from various CPS analysis are summarized in Table 5.7-1.

TABLE 5.7-1 OTHER HAZARD GROUP CONTRIBUTOR

SUMMARY

OTHER HAZARD INITIATOR GROUP CDF (1/YR)

Seismic [9] 1.7E-05 Internal Fire [8] 6.0E-06 High Winds/Tornadoes Screened External Floods Screened Transportation and Nearby Facility Accidents Screened Total (for initiators with CDF available) 2.3E-05 Internal Events CDF 2.23E-06 (1)

External Events Multiplier 10.31 Note to Table 5.7-1:

(1)

The multiple for seismic alone is 7.62.

The EPRI Category 3b frequency for the 3-per-10 year, 1-per-10 year, and 1-per-15 year ILRT/DWBT intervals are shown in Table 5.6-1 as 2.43E-09/yr, 8.12E-09/yr, and 1.22E-08/yr, respectively. Using the other hazard group multiplier of 10.31 for CPS, the change in the LERF risk measure due to extending the ILRT/DWBT from 5-19 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 3-per-10 years to 1-per-15 years, including both internal events and other measurable hazard groups hazards risk, is estimated as shown in Table 5.7-2.

TABLE 5.7-2 CPS 3B (LERF) AS A FUNCTION OF ILRT/DWBT FREQUENCY FOR INTERNAL AND EXTERNAL EVENTS (INCLUDING AGE ADJUSTED STEEL CORROSION LIKELIHOOD) 3B 3B FREQUENCY 3B FREQUENCY (3-PER-10 FREQUENCY (1-PER-15 YEAR (1-PER-10 YEAR YEAR LERF (1)

ILRT/DWBT) ILRT/DWBT) ILRT/DWBT) INCREASE Internal Events Contribution 2.43E-09 8.12E-09 1.22E-08 9.81E-09 Other Hazard Group Contribution (Internal 2.51E-08 8.38E-08 1.26E-07 1.01E-07 Events CDF x 10.31)

Combined 2.75E-08 9.19E-08 1.38E-07 1.11E-07 Note to Table 5.7-2:

(1)

Associated with the change from the baseline 3-per-10 year frequency to the proposed 1-per-15 year frequency.

Thus, the total increase in LERF (measured from the baseline 3-per-10 year ILRT interval to the proposed 1-per-15 year frequency) due to the combined internal and external events contribution is estimated as 1.11E-07/yr, which includes the age adjusted steel corrosion likelihood.

The other metrics for the ILRT/DWBT interval extension risk assessment can be similarly derived using the multiplier approach. The results between the 3-in-10 year interval and the 15 year interval compared to the acceptance criteria are shown in Table 5.7-3. As can be seen, the impacts from including the other hazard group contributors are as follows:

1. Change in LERF = 1.11E-7/yr, which is slightly above the 1.0E-7/yr upper boundary for the very small risk increase as defined in RG 1.174, but at the bottom of the band for small risk increase.

2. Change in population dose rate is 4.30E-2 person-rem/yr (0.73%), which is less than 1.0 person-rem/year or 1% of the total population dose.

3. Change in CCFP is 0.44%, which is less than 1.5%.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Thus, the inclusion of external events does not change the conclusion of the risk assessment. That is, the acceptance criteria are all met such that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years has been demonstrated to be small. Note that a bounding analysis for the total LERF contribution follows Table 5.7-3 to demonstrate that the total LERF value for CPS is less than 1.0E-5/yr consistent with the requirements for a Small Change in risk of the RG 1.174 acceptance guidelines.

TABLE 5.7-3 COMPARISON TO ACCEPTANCE CRITERIA INCLUDING OTHER HAZARD GROUPS CONTRIBUTION FOR CPS LERF PERSON-REM/YR CCFP (1) (1)

CONTRIBUTOR CPS Internal Events 9.81E-9/yr 3.80E-03/yr (0.73%) 0.44%

CPS Other Hazard 1.01E-7/yr 3.92E-02/yr (0.73%) 0.44%

Groups CPS Total 1.11E-7/yr 4.30E-02/yr (0.73%) 0.44%

Acceptance Criteria <1.0E-6/yr <1.0 person-rem/yr or 1.5%

<1.0%

Notes to Table 5.7-3:

(1)

The EPRI Class (1, 2, 7, 8) release Person-Rem/yr are assumed to be the same percentage relative to base risk (0.73%) for internal and external events.

(2)

The Probability of DW and WW leakage due to the ILRT/DWBT extension is assumed the same for both Internal and External Events, therefore the percentage change for CCFP remains constant (0.44%).

The 1.11E-07/yr increase in LERF due to the combined hazard events from extending the CPS ILRT/DWBT frequency from 3-per-10 years to 1-per-15 years falls within Region II between 1E-7 to 1E-6 per reactor year (Small Change in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when the calculated increase in LERF due to the proposed plant change is in the Small Change range, the risk assessment must also reasonably show that the total LERF is less than 1E-5/yr. Similar bounding assumptions regarding the external event contributions that were made above are used for the total LERF estimate.

5-21 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval From Table 4.2-2, the total LERF due to postulated internal event accidents is the sum of the LERF release categories, which is 1.16E-7/yr. For Fire, the total LERF is 9.21E-07/yr [8]. The base LERF due to seismic is assumed to be in the same proportion as the internal events contribution. The total LERF value for CPS is then shown in Table 5.7-4.

TABLE 5.7-4 IMPACT OF 15-YR ILRT EXTENSION ON LERF (3B)

FOR CPS Internal Events LERF 1.16E-07/yr Internal Fire LERF 9.21E-07/yr Other Hazard Group LERF 8.84E-07/yr (Internal Events LERF x 7.62)

Internal Events LERF due to ILRT (1) 1.22E-08/yr (Class 3b) at 15 years Other Hazard group LERF due to (1) 1.26E-07/yr ILRT at 15 years Total 2.06E-06/yr Note to Table 5.7-4:

(1)

Including age adjusted steel corrosion likelihood.

As can be seen, the estimated LERF for CPS using the Fire LERF and CDF based multiplier approach for seismic is 2.1E-06/yr, which is less than the RG 1.174 required value of 1E-5/yr.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 6.0 SENSITIVITIES 6.1 SENSITIVITY TO CORROSION IMPACT ASSUMPTIONS The results in Tables 5.2-2, 5.3-1, and 5.3-2 show that including corrosion effects calculated using the assumptions described in Section 4.5 does not significantly affect the results of the ILRT/DWBT extension risk assessment. Sensitivity cases were developed to gain an understanding of the sensitivity of the results to the key parameters in the corrosion risk analysis. The time for the flaw likelihood to double was adjusted from every five years to every two and every ten years. The failure probabilities for the cylinder, dome and basemat were increased and decreased by an order of magnitude. The total detection failure likelihood was adjusted from 10% to 15%

and 5%. The results are presented in Table 6.1-1. In every case, the impact from including the corrosion effects is minimal. Even the upper bound estimates with conservative assumptions for all of the key parameters yield increases in LERF due to corrosion of only 3.63E-09/yr. The results indicate that even with conservative assumptions, the conclusions from the base analysis would not change.

6-1 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 6.1-1 STEEL CORROSION SENSITIVITY CASES INCREASE IN CLASS 3B FREQUENCY (LERF)

FOR ILRT/DWBT EXTENSION VISUAL FROM 3 IN 10 TO 1 IN 15 YEARS INSPECTION & (PER YEAR)

CONTAINMENT NON-VISUAL AGE BREACH FLAWS (STEP 3 IN THE (STEP 4 IN THE (STEP 5 IN THE CORROSION CORROSION CORROSION TOTAL INCREASE DUE TO ANALYSIS) ANALYSIS) ANALYSIS) INCREASE CORROSION Base Case Base Case Base Case (1.0% Cylinder- (10% Cylinder-Doubles every Dome, Dome, 9.81E-09 1.14E-10 5 yrs 0.1% Basemat) 100% Basemat)

Doubles every Base Base 9.95E-09 2.59E-10 2 yrs Doubles every Base Base 9.79E-09 9.56E-11 10 yrs 15% Cylinder-Base Base 9.85E-09 1.59E-10 Dome 5% Cylinder-Base Base 9.76E-09 6.88E-11 Dome 10% Cylinder-Base Dome, 1% Base 1.08E-08 1.14E-09 Basemat 0.01% Cylinder-Base Dome, 0.001% Base 9.70E-09 1.14E-11 Basemat Lower Bound 0.1% Cylinder- 5% Cylinder-Doubles every Dome, 0.001% Dome 100% 9.70E-09 5.74E-12 10 yrs Basemat Basemat Upper Bound 10% Cylinder- 15% Cylinder-Doubles every Dome, 1% Dome 1.33E-08 3.63E-09 2 yrs Basemat 100% Basemat 6-2 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 6.2 EPRI EXPERT ELICITATION SENSITIVITY An expert elicitation was performed to reduce excess conservatisms in the data associated with the probability of undetected leaks within containment [3]. Since the risk impact assessment of the extensions to the ILRT interval is sensitive to both the probability of the leakage as well as the magnitude, it was decided to perform the expert elicitation in a manner to solicit the probability of leakage as a function of leakage magnitude. In addition, the elicitation was performed for a range of failure modes which allowed experts to account for the range of failure mechanisms, the potential for undiscovered mechanisms, inaccessible areas of the containment as well as the potential for detection by alternate means. The expert elicitation process has the advantage of considering the available data for small leakage events, which have occurred in the data, and extrapolate those events and probabilities of occurrence to the potential for large magnitude leakage events.

The basic difference in the application of the ILRT interval methodology using the expert elicitation is a change in the probability of pre-existing leakage within containment. The base case methodology uses the Jeffreys non-informative prior for the large leak size and the expert elicitation sensitivity study uses the results from the expert elicitation. In addition, given the relationship between leakage magnitude and probability, larger leakage that is more representative of large early release frequency can be reflected.

For the purposes of this sensitivity, the same leakage magnitudes that are used in the base case methodology (i.e., 10La for small and 100La for large) are used here. Table 6.2-1 illustrates the magnitudes and probabilities of a pre-existing leak in containment associated with the base case and the expert elicitation statistical treatments. These values are used in the ILRT interval extension for the base methodology and in this sensitivity case. Details of the expert elicitation process, including the input to expert elicitation as well as the results of the expert elicitation, are available in the various appendices of EPRI TR-1018243 [3].

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 6.2-1 EPRI EXPERT ELICITATION RESULTS EXPERT ELICITATION MEAN PROBABILITY PERCENT LEAKAGE SIZE (La) BASE CASE OF OCCURRENCE [3] REDUCTION 10 9.2E-03 3.88E-03 58%

100 2.3E-03 2.47E-04 89%

The summary of results using the expert elicitation values for probability of containment leakage is provided in Table 6.2-2. As mentioned previously, probability values are those associated with the magnitude of the leakage used in the base case evaluation (10La for small and 100La for large). The expert elicitation process produces a relationship between probability and leakage magnitude in which it is possible to assess higher leakage magnitudes that are more reflective of large early releases; however, these evaluations are not performed in this particular study.

The net effect is that the reduction in the multipliers shown above has the same impact on the calculated increases in the LERF values. The increase in the overall value for LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is 1.15E-09/yr. Similarly, the increase due to increasing the interval from 10 to 15 years is 5.01E-10/yr. As such, if the expert elicitation mean probabilities of occurrence are used instead of the non-informative prior estimates, the change in LERF for CPS is much further within the range of a very small change in risk when compared to the current 1-in-10, or baseline 3-in-10 year requirement. The results of this sensitivity study are judged to be more indicative of the actual risk associated with the ILRT extension than the results from the assessment as dictated by the values from the EPRI methodology [3], and yet are still conservative given the assumption that all of the Class 3b contribution is considered to be LERF.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 6.2-2 CPS ILRT/DWBT CASES:

3 IN 10 (BASE CASE), 1 IN 10, AND 1 IN 15 YR INTERVALS (ILRT LEAKAGE BASED ON EPRI EXPERT ELICITATION PROBABILITIES)

BASE CASE EXTEND TO EXTEND TO 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARS EPRI DOSE PERSON- PERSON- PERSON-CLASS PER-REM CDF (1/YR) REM/YR CDF (1/YR) REM/YR CDF (1/YR) REM/YR 1 2.71E+03 9.23E-07 2.51E-03 9.08E-07 2.46E-03 8.96E-07 2.43E-03 2 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 2.68E-07 1.47E-01 3a 2.71E+04 6.44E-09 1.75E-04 2.15E-08 5.82E-04 3.22E-08 8.74E-04 3b 2.71E+05 2.67E-10 7.25E-05 9.12E-10 2.48E-04 1.41E-09 3.84E-04 7 LERF 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 1.14E-07 6.27E-02 7 non-LERF 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 9.16E-07 3.09E-01 8 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 1.55E-09 8.49E-04 Total 2.23E-06 0.522 2.23E-06 0.522 2.23E-06 0.523 ILRT Dose Rate from 3a 2.47E-04 8.30E-04 1.26E-03 and 3b Delta From 3 yr --- 5.38E-04 9.37E-04 Total Dose From 10 yr --- --- 3.99E-04 Rate (1) 3b Frequency (LERF) 2.67E-10 9.12E-10 1.41E-09 Delta 3b From 3 yr --- 6.45E-10 1.15E-09 LERF From 10 yr --- --- 5.01E-10 CCFP % 58.31% 58.34% 58.36%

Delta From 3 yr --- 0.03% 0.05%

CCFP %

From 10 yr --- --- 0.02%

Note to Table 6.2-2:

(1)

The overall difference in total dose rate is less than the difference of only the 3a and 3b categories between two testing intervals. This is because the overall total dose rate includes contributions from other categories that do not change as a function of time, e.g., the EPRI Class 2 and 8 categories, and also due to the fact that the Class 1 person-rem/yr decreases when extending the IRLT/DWBT frequency.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 6.3 DWBT DATA SENSITIVITY An additional sensitivity is included related to the interpretation of the DWBT data used for the base case assessment. The probability of small and large drywell failures is increased by a factor of 10. The base case applied the containment failure probabilities to the drywell failure probabilities. The base case probabilities are considered conservative 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 NEI/EPRI approach is reflective of drywell failures. The body of plant experience used considered the older BWR containment designs.

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

• 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 4.1-1. As this comparison shows, the basic designs are much the same and therefore would be expected to have much the same leakage failure mechanisms.

• 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.

TABLE 6.3-1 DW LEAKAGE X10 PROBABILITY SENSITIVITY DW SENSITIVITY DW WW WW LEAKAGE DW LEAKAGE PROBABILITY LEAKAGE LEAKAGE SIZE (LB) LEAKAGE PROBABILITY SIZE (LA)

(BASE) PROBABILITY 10Lb 0.0092 0.092 10La 0.0092 100Lb 0.0023 0.023 100Lb 0.0023 Note: WW Leakage (10La and 100La remains constant) 6-6 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The summary of results using the revised values for probability of drywell bypass leakage is provided in Table 6.3-2. The results indicate increases to the population dose and to the CCFP values compared to the base risk assessment, but the results are all still within the acceptance criteria of less than 1.0 person-rem/yr or less than 1.0% person-rem/yr, and less than 1.5% change in CCFP.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 6.3-2 CPS ILRT/DWBT CASES:

3 IN 10 (BASE CASE), 1 IN 10, AND 1 IN 15 YR INTERVALS (DWBT 10LB, 100LB LEAK PROB. INCREASED BY A FACTOR OF 10)

BASE CASE EXTEND TO EXTEND TO 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARS EPRI DOSE PERSON- PERSON- PERSON-CLASS PER-REM CDF (1/YR) REM/YR CDF (1/YR) REM/YR CDF (1/YR) REM/YR 1 2.71E+03 8.94E-07 2.43E-03 8.11E-07 2.20E-03 7.51E-07 2.04E-03 2 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 2.68E-07 1.47E-01 3a 2.71E+04 3.31E-08 8.99E-04 1.10E-07 2.99E-03 1.66E-07 4.50E-03 3b 2.71E+05 2.63E-09 7.14E-04 8.78E-09 2.38E-03 1.32E-08 3.59E-03 7 LERF 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 1.14E-07 6.27E-02 7 non-LERF 3.37E+05 9.16-07 3.09E-01 9.16E-07 3.09E-01 9.16E-07 3.09E-01 8 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 1.55E-09 8.49E-04 Total 2.23E-06 0.523 2.23E-06 0.527 2.23E-06 0.530 ILRT Dose Rate from 3a 1.61E-03 5.38E-03 8.09E-03 and 3b Delta From 3 yr --- 3.54E-03 6.09E-03 Total Dose From 10 yr --- --- 2.55E-03 Rate (1) 3b Frequency (LERF) 2.63E-09 8.78E-09 1.32E-08 Delta 3b From 3 yr --- 6.15E-09 1.06E-08 LERF From 10 yr --- --- 4.45E-09 CCFP % 58.41% 58.69% 58.89%

Delta From 3 yr --- 0.28% 0.48%

CCFP %

From 10 yr --- --- 0.20%

Note to Table 6.3-2:

(1)

The overall difference in total dose rate is less than the difference of only the 3a, and 3b categories between two testing intervals. This is because the overall total dose rate includes contributions from other categories that do not change as a function of time, e.g., the EPRI Class 2 and 8 categories, and also due to the fact that the Class 1 person-rem/yr decreases when extending the IRLT/DWBT frequency.

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7.0 CONCLUSION

S Based on the results from Section 5 and the sensitivity calculations presented in Section 6, the following conclusions regarding the assessment of the plant risk are associated with permanently extending the Type A ILRT test frequency and the DWBT frequency to fifteen years:

• 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 10-6/yr and increases in LERF below 10-7/yr. Small changes in risk are defined as increases in CDF below 10-5/yr and increases in LERF below 10-6/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 and the DWBT interval for the base case with corrosion included is 9.81E-09/yr (see Table 6.1-1), which falls within the very small change region of the acceptance guidelines in Reg.

Guide 1.174.

If the EPRI Expert Elicitation methodology Class 3a and Class 3b failure probabilities are used, the change is estimated as 1.15E-09/yr (see Table 6.2-2), which falls further within the very small change region of the acceptance guidelines in Reg. Guide 1.174.

• The change in dose risk for changing the Type A ILRT interval and the DWBT interval from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.80E-03 person-rem/yr using the EPRI guidance with the base case corrosion included (see Table 5.6-1). This change meets both of the related acceptance criteria identified in Section 1.3 for change in population dose of less than 1.0 person-rem/ year or less than 1% person-rem/yr.

The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology (see Table 6.2-2). The change in dose risk meets both of the related acceptance criteria identified in Section 1.3 for change in population dose of less than 1.0 person-rem/ year or less than 1% person-rem/yr.

• The increase in the conditional containment failure frequency from the three in ten year interval to one in fifteen years including corrosion effects using the EPRI guidance (see Table 5.6-1) is 0.44%, which is below the acceptance criteria of 1.5% identified in the NRC SER on the issue [7] as discussed in Section 1.3.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The increase in CCFP drops to about 0.05% using the EPRI Expert Elicitation methodology (see Table 6.2-2). This value meets both of the related acceptance criteria identified in Section 1.3 for change in CCFP of less than 1.5%.

• To determine the potential impact from other hazard groups, an additional bounding assessment from the risk associated with the other relevant hazard groups for CPS utilizing the latest information from various sources was performed. As shown in Table 5.7-3, the total increase in LERF due to internal events and other hazard groups is 1.11E-07/yr, which is in Region II of the Reg. Guide 1.174 acceptance guidelines. As also shown in Table 5.7-3, the other acceptance criteria for change in population dose and change in CCFP are also still met when the other hazard groups are considered in the analysis.

• Finally, as shown in Table 5.7-4, a similar bounding analysis for the other hazard groups indicates that the total LERF from both internal events and the other hazard groups is 2.06E-06/yr, which is less than the Reg. Guide 1.174 limit of 1E-05/yr given that the LERF is in Region II (small change in risk).

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

Previous Assessments The NRC in NUREG-1493 [6] has previously concluded the following:

• 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.

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The findings for CPS confirm these general findings on a plant specific basis for the ILRT/DWBT interval extension considering the severe accidents evaluated for CPS, the CPS containment failure modes, and the local population surrounding CPS.

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8.0 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 2, May 2011.

[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 2014 Fire PRA Summary and Quantification Notebook, CPS-PSA-021-06, Revision 1, December 2014.

[9] U.S. Nuclear Regulatory Commission (NRC), Memorandum to Brian W. Sheron, Director Office of Nuclear Regulatory Research, From Patrick Hiland, Chairman, Safety/Risk Assessment Panel for Generic Issue 199: Safety/Risk Assessment Results for Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates In Central and Eastern United States on Existing Plants, Accession Number ML11356A034, September 2, 2010.

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[10] Letter U-602549 from Wilfred Connell (Vice President CPS) to the NRC, Subject Clinton Power Station Proposed Amendment of Facility Operating License No.

NPF-62 (LS-96-001), February 22, 1996.

[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.

[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. 16, January, 2014.

[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 8-2 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 (2014 PRA Interim Update), CPS PSA-013, Revision 4, March 2014.

[25] CPS PRA Quantification Notebook, CPS PSA-014, Revision 6, March 2014.

[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] Letter from Thomas P. Kirwin (Entergy, Palisades Nuclear Plant) to U.S. Nuclear Regulatory Commission, License Amendment Request to Extend the Containment Type A Leak Rate Test Frequency to 15 Years, Accession Number ML110970616, April 6, 2011.

[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 2, March 2014.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

[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] CPS Technical Specification Section 3.6 of Amendment No. 158.

[35] U.S. Nuclear Regulatory Commission, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, NUREG-1407, June 1991.

[36] Electric Power Research Institute, A Methodology for Assessment of Nuclear Power Plant Seismic Margin. EPRI NP-6041-LS, Rev. 1, August 1991.

[37] Clinton Power Station Individual Plant Examination For External Events, September 1995.

[38] EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, EPRI 1011989, NUREG/CR-6850, September 2005.

[39] Letter from James Barstow (Exelon Generation Company, LLC) 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" (Peach Bottom Nuclear Station), Accession Number ML14315A084, November 7, 2014.

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy APPENDIX A PRA TECHNICAL ADEQUACY C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A

A.1 OVERVIEW A technical Probabilistic Risk Assessment (PRA) analysis is presented in this report to help support an extension of the Clinton Unit 1 containment Type A test integrated leak rate test (ILRT) interval to fifteen years. The analysis follows the guidance provided in Regulatory Guide 1.200, Revision 2 [A-3], An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities. The guidance in RG-1.200 indicates that the following steps should be followed to perform this study:

1. Identify the parts of the PRA used to support the application SSCs, operational characteristics affected by the application and how these are implemented in the PRA model.

A definition of the acceptance criteria used for the application.

2. Identify the scope of risk contributors addressed by the PRA model If not full scope (i.e. internal and external), identify appropriate compensatory measures or provide bounding arguments to address the risk contributors not addressed by the model.

3. Summarize the risk assessment methodology used to assess the risk of the application Include how the PRA model was modified to appropriately model the risk impact of the change request.

4. Demonstrate the Technical Adequacy of the PRA Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the application.

Document peer review findings and observations that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.

Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the Regulatory Guide. Provide justification to show that where specific requirements in the standard are not met, it will not unduly impact the results.

Identify key assumptions and approximations relevant to the results used in the decision-making process.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy Items 1 through 3 are covered in the main body of this report. The purpose of this appendix is to address the requirements identified in item 4 above. Each of these items (plant changes not yet incorporated into the PRA model, relevant peer review findings, consistency with applicable PRA standards and the identification of key assumptions) are discussed in the following sections.

The risk assessment performed for the ILRT extension request is based on the current Level 1 and Level 2 PRA model. Note that for this application, the accepted methodology involves a bounding approach to estimate the change in the PRA risk metric of LERF from extending the ILRT interval. Rather than exercising the PRA model itself, it involves the establishment of separate evaluations that are linearly related to the plant CDF contribution. Consequently, a reasonable representation of the plant CDF that does not result in a LERF does not require that Capability Category II be met in every aspect of the modeling if the Category I treatment is conservative or otherwise does not significantly impact the results.

A discussion of the Exelon model update process, the peer reviews performed on the Clinton PRA model, the results of those peer reviews and the potential impact of peer review findings on the ILRT extension risk assessment are provided in Section A.2.

Section A.3 provides a qualitative assessment of External Hazards Fire and Seismic inputs. Section A.4 provides an assessment of key assumptions and approximations used in this assessment and Section A.5 briefly summarizes the results of the PRA technical adequacy assessment with respect to this application.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.2 PRA Model Evolution and Peer Review Summary A.2.1 Introduction The 2014A version of the CPS PRA model is the most recent evaluation of the Unit 1 risk profile at CPS for internal event challenges. The CPS PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the CPS PRA is based on the event tree/fault tree methodology, which is a well-known methodology in the industry.

Exelon Generation Company, LLC (Exelon) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the 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. The following information describes this approach as it applies to the CPS PRA.

PRA Maintenance and Update The Exelon risk management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plant. This process is defined in the Exelon Risk Management program, which consists of a governing procedure and subordinate implementation procedures. The PRA model update procedure delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating Exelon nuclear generation sites. The overall Exelon Risk Management program defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following activities are routinely performed:

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• Design changes and procedure changes are reviewed for their impact on the PRA model.

• Maintenance unavailabilities are captured, and their impact on CDF is trended.

• Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated approximately every four years.

In addition to these activities, Exelon risk management procedures provide the guidance for particular risk management maintenance activities. This guidance includes:

• Documentation of the PRA model, PRA products, and bases documents.

• The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.

• Guidelines for updating the full power, internal events PRA models for Exelon nuclear generation sites.

• Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10 CFR 50.65(a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately 4-year cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant.

The 2014A model was completed in March of 2014.

As indicated previously, RG 1.200 also requires that additional information be provided as part of the LAR submittal to demonstrate the technical adequacy of the PRA model used for the risk assessment. Each of these items (plant changes not yet incorporated into the PRA model, relevant peer review findings, and consistency with applicable PRA Standards) will be discussed in turn in this section.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.2.2 Plant Changes Not Yet Incorporated into the PRA Model A PRA updating requirements evaluation (URE) documented in the Clinton PRA model update tracking database is created for all issues that are identified that could impact the PRA model. The URE database includes the identification of those plant changes that could impact the PRA model.

A review of the open UREs indicates that there are no plant changes that have not yet been incorporated into the PRA model that would affect this application. UREs are evaluated for potential impact to applications and to the PRA base model results and are classified as High, Medium or Low priority. High priority items could significantly impact applications. Medium priority items are items that are assessed as potentially important to applications and Low priority items area items that are assessed as not important to applications and likely to have minimal or no numeric impact. There are no open High priority UREs and seventeen UREs identified as Medium priority. The remaining open UREs are low priority, having little or no impact to the PRA results.

Medium priority UREs were found not to impact this application. Low priority UREs were also reviewed and none were found that would impact this application.

A.2.3 Consistency with Applicable PRA Standards The Clinton FPIE PRA model has undergone several reviews including a BWROG Peer Review in 2000 [A-5, 6, and 7]. UREs were created and extensive changes were made to the PRA model in updates through 2006. As a result of the extensive changes made to the model a full Peer review was again performed in 2009 [A-12]. The results of the 2009 Peer review and the actions taken to address gaps identified, best represent the consistency of the model to current PRA standards.

The 2009 Peer Review was conducted using the 2009 ASME/ANS PRA Standard [A-1]

and the NRCs comments and clarifications contained in RG 1.200 [A-3]. The Peer Review was conducting using the CPS 2006C FPIE PRA model [A-4]. The general objective for the CPS PRA is to meet Capability Category II. The findings and A-5 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy observations (F&Os) that were identified in the Peer Review were between the 2006C CPS PRA and the requirements for Capability Category II. These F&Os (i.e., both findings and suggestions) were entered into the CPS Updating Requirements Evaluation (URE) database for tracking purposes. These UREs were used for scoping of the Clinton 2011 PRA update.

All Findings from the 2009 Peer Review were addressed as part of the 2011 PRA update. The 2009 CPS Peer Review observations were incorporated into the CPS URE database for tracking. All but one of the Suggestion observations have been addressed.

Following the 2009 Peer review, a self-assessment relative to the combined ASME/ANS PRA Standard [A-1] and the NRCs comments and clarifications contained in RG 1.200.

[A-3] was performed as part of the 2011 PRA update. The 2011 Self-Assessment used the 2009 Peer review results as input. Table A-2 lists gaps to Category II identified in the 2011 self-assessment and the status of those gaps following the 2011 and 2014 PRA updates. Included in the last column of Table A-2 is the URE number, a significance statement and the impact to this ILRT/DWBT for the gaps that have not been addressed. These gaps are judged to not have an impact on this application as justified in Table A-2.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy Table A-2

SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 1 Review initiating event precursors in identifying the IE-A7 Completed: URE CL2009-006 has been closed. This initiating events to be modeled. documentation aspect has been incorporated into the CPS PRA Initiating Event Notebook. This work included review of hundreds of events INPO SENs, SOERs, A rigorous explicit assessment of all the events in NUREG-SERs, and NRC SECY letters on precursors, as well as 1275 could be pursued (if determined that this is the true CPS specific experience records. No new initiating intent of SR IE-A7); however, such an effort is judged not event categories were identified.

to provide much benefit to the CPS IE analysis.

2 Loss of switchgear room cooling assumptions should be IE-C4 Deferred: Switchgear room cooling calculations were supported by room cooling calculation. not performed prior to completion of the 2014 PRA Update..This item is deferred and maintained here for future consideration.

Tracked under URE L2010-012. The current modeling assumptions regard for the need for room cooling in the long term are judged realistic. Performance of loss of cooling calculations is not expected to impact the model or results. No impact to the CPS ILRT/DWBT application.

3 Complete CPS URE 2001-055 (recommendation to AS-B3, Completed: URE CL2001-055 has been closed. The perform loss of room cooling calculations to support PRA SC-B2, documentation exists in the PRA Dependency Notebook success criteria assumptions). SC-C1, (CPS PSA-006). Room cooling assumptions were SC-C2, confirmed with the system manager during system SY-A19, manager interviews and are judged to be reasonable.

SY-B7 4 Complete CPS URE 2001-061 (concerning SY-A17, Completed: URE CL2001-061 has been closed. The recommendation to specifically perform loss of room SY-B8 documentation exists in the PRA Dependency Notebook cooling calculation for switchgear rooms). (CPS PSA-006). Room cooling assumptions were confirmed with the system manager during system manager interviews and are judged to be reasonable.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy Table A-2

SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 5 Complete URE 2001-144 (concerning recommendations to SY-B8 Completed: URE 2001-144 has been closed and the enhance containment isolation documentation). documentation enhanced.

6 To meet the requirements of SR HR-A1, the following HR-A1, Closed: The CPS 2011 updated involved re-assessing would be developed as supporting documentation for CPS: HR-A2, each PRA system for pre-initiator HEPs. The detailed

• A list of the PRA systems to consider for HR-A3, pre-IE identification process is described and test and maintenance actions HR-C2, documented in Appendix J of the HRA notebook (CPS-

• Rules for identifying and screening test HR-C3 PSA-004).

and maintenance actions from the PRA

• A list of procedures reviewed, the potential test and maintenance actions associated with the procedures, and the disposition of the action (screened or evaluated).

• Identify T&M activities that require realignment of the system outside its normal operational or stand by status.

However, performing this task is judged not to have significant impact on the PRA model and results.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy Table A-2

SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 7 Complete URE 2001-084 (concerning use of screening HR-B1, Closed: URE CL2001-084 was closed as part of the values for dominant pre-initiator HEPs). HR-B2 2011 PRA update. Pre-initiators were identified as part of a detailed system analysis described in Appendix J of the Clinton HRA notebook (PSA PSA-004). The pre-Leading candidates for risk significant pre-initiators were initiator HEPs that were identified in App. J were identified through review of the BNL study on sensitivity of calculated using the EPRI HRA Calculator and plant PRA to HEPs, review of other PRAs, and review of plant-specific procedures were available.

specific operating experience. Non-risk significant pre-initiators were screened out. The approach taken is one that reflects the best use of resources by excluding pre-initiators for which data is unavailable and overall contribution is insignificant. To meet the requirements of SR HR-B1, the following would be developed as supporting documentation for CPS:

• A list of the PRA systems to consider for pre-initiator actions.

• Rules for identifying and screening pre-initiator actions from the PRA.

• A list of procedures reviewed, the potential pre-initiator actions associated with the procedures, and the disposition of the action (screened or evaluated).

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy Table A-2

SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 8 Complete URE 2001-084. Although this will not HR-D1, Closed: URE CL2001-084 was closed as part of the significantly impact the HRA results, future PRA updates HR-D2, 2011 PRA update. Pre-initiators were identified as part should include an assessment of the quality of plant written HR-D3, of a detailed system analysis described in Appendix J of procedures and administrative controls as well as human- HR-D4 the Clinton HRA notebook (PSA PSA-004). The pre-machine interface for both pre-initiator and post-initiator initiator HEPs that were identified in App. J were human actions. calculated using the EPRI HRA Calculator and plant specific procedures were available.

Alternative:

Possible upgrade to the pre-initiator HRA to include specific quantifications for each pre-initiator HEP would be strict compliance with the standard. This is not considered necessary for most applications. It is recommended that CPS await further ASME clarification on this item before proceeding. This can be confirmed for each application in lieu of performing the quantifications.

9 Failure data development using surveillance test data DA-C10 Deferred: Current industry PRA efforts and PRA peer should fulfill the requirements of DA-C10, and should be reviews are having difficulty understanding the full intent documented appropriately. Review surveillance test of this SR. Future updates of the CPS PRA will procedures and identify all failure modes that are fully consider enhancement to the documentation and tested by the procedures. Include data for the failure investigation of the plant failure data implied by this SR.

modes that are fully tested. The results of unplanned The impact on the overall CDF or LERF values is demands on equipment should also be accounted for. judged to be non-significant.

Tracked by URE CL2009-008, Primarily a documentation/process issue. Impact on overall CDF and LERF expected to be negligible. No impact to the CPS ILRT/DWBT application.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy Table A-2

SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 10 As needed in maintenance unavailability determination, DA-C13 Completed: All the risk significant maintenance events perform interviews of maintenance staff for equipment with in the CL11a PRA, with the exception of two, are based incomplete or limited maintenance information and on current plant-specific data and are judged document appropriately. reasonable. Two risk significant maintenance events, 1ADASDIV2SRVSM-- and 1APTR-RATSVC-M--, are based on plant-specific data from the previous PRA revision and are judged reasonable. Interviews not warranted.

11 Identify significant basic events that contribute to the QU-D5a Closed: The support system initiating event fault trees significant initiating events whose frequencies are have been incorporated into the single-top CPS 2011 quantified using fault tree methods. model. Basic event importance measures are Otherwise, importance measures calculated and assessed calculated for all SSIE basic events that appear in the to ensure results make logical sense. cutsets.

12 Strict reading of SR QU-F2 would indicate that the QU-F2 Closed: Item (b) has been incorporated into the following enhancements to the documentation of the CPS Quantification Notebook of the CPS 2011 update.

PRA would need to be made to comply with the Standard:

a) Provide a list of human actions and equipment Deferred: Items (a) and (c) are documentation failures (significant basic events) that cause enhancements for the base PRA and are maintained for accidents to be non-dominant. consideration for future updates. Non-significant b) Bases for the elimination of mutually exclusive documentation item.

events from the model need to be added.

c) Include cutsets segregated by accident sequence Tracked by URE CL2009-011. Considered in the documentation. This is available but may documentation enhancements. No impact to the CPS not be needed in the formal documentation. This ILRT/DWBT application.

should await further ASME/ANS clarification before extensive resources are committed.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.3 EXTERNAL HAZARDS Although EPRI report 1018243 [A-10] recommends a quantitative assessment of the contribution of external events (for example, fire and seismic) where a model of sufficient quality exists, it also recognizes that the external events assessment can be taken from existing, previously submitted and approved analyses or another alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. Based on this, currently available information for external events models was referenced, and a multiplier was applied to the internal events results based on the available external events information. This is further discussed in Section 5.7 of the risk assessment. The fire and seismic PRA Technical Adequacy are discussed in additional detail.

A.3.1 FIRE PRA TECHNICAL ADEQUACY The Clinton Power Station (CPS) Fire Probabilistic Risk Assessment (FPRA) is an update/upgrade of the original fire risk assessment performed as part of the plants Individual Plant Examination of External Events (IPEEE) and as previously updated.

The 2014 FPRA update generally uses the fire scenarios developed for the 2008 FPRA update and incorporates them into the 2014 Full Power Internal Events (FPIE) model.

This FPRA is an interim implementation of NUREG/CR-6850 [A-8]. That is, not all tasks identified in NUREG/CR-6850 are completely addressed or implemented in this update due to the limited scope of the current incremental update and due to the changing state-of-the-art of the industry at the time of the 2014 FPRA development. The 2014 FPRA has therefore not received a Peer Review.

The methodologies employed in the FPRA analysis are consistent with those provided in NUREG/CR-6850 [A-8] and Supplement 1 [A-9]; however, no commitment to compliance with the NUREG is made or implied. At this time, there is no commitment for A-12 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy CPS to transition to 50.49(c), Performance based / Risk Informed Fire Protection, or NFPA 805.

A.3.2 Fire PRA Limitations This 2014 Fire PRA Update is an interim implementation of NUREG/CR-6850 [A-8].

That is, not all tasks identified in NUREG/CR-6850 are completely addressed or implemented in this update due to limited scope of the current incremental update or due to the changing state-of-the-art of industry at the time of the 2014 CPS FPRA development. Limitations and other precautions regarding the development of the 2014 FPRA, in terms of the tasks identified in NUREG/CR-6850 [A-8], are as follows:

• Confirmatory Walkdowns (Tasks 1-3, 6) - Confirmatory ignition source and scenario walkdowns for the 2014 PRA update were not conducted due to the limited scope of the project. The results and insights from the 2008 FPRA walkdowns were retained for the 2014 update. These are judged adequate for the ILRT/DWBT risk evaluation.

• Instrumentation Review (Task 2) - The requirements of NUREG/CR-6850 regarding the explicit identification and modeling of instrumentation required to support PRA credited operator actions is not fully addressed.

The instrumentation review from the FPIE PRA is retained for the FPRA update. Available instrumentation is generally redundant and diverse such that it is judged unlikely that more detailed treatment would significantly impact the FPRA, especially for use in the ILRT/DWBT risk evaluation.

• Balance of Plant (Task 2) - The BOP (PCS) is not fully treated. BOP support system failure is conservatively assumed in locations where related components and cables may exist. This represents a potential conservatism in the model, and therefore potential conservativism in the Fire CDF used in the ILRT/DWBT risk evaluation.

• Limited Analysis Iterations (Tasks 9-12) - The process of conducting a FPRA is iterative, identifying conservative assumptions and high risk compartments and performing analyses to refine the assumptions and reduce those compartment risks. The ability to conduct iterations is limited based on resources. The scenarios included in the 2014 CPS FPRA may benefit from additional refinement. This reflects a potential conservativism in the Fire CDF used in the ILRT/DWBT risk evaluation.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy

• Multi-Compartment Review (Task 11) - This subtask reviews the fire analysis compartment boundaries to ensure they are sufficiently robust to prevent the spread of fire between FPRA Physical Analysis Units (PAUs) or that such propagations are adequately addressed by the developed scenarios (i.e., multi-compartment fire scenarios). Based on the 2008 FPRA walkdowns, the compartment boundaries abilities to contain the effects of fires were deterministically considered as part of scenario development. Probabilistic failure potential (e.g., failure of fire door due to fire impacts), while not currently included in the FPRA in terms of multi-compartment fire scenarios is judged to be a very small contributor to the overall Fire CDF that is far outweighed by other FPRA conservatisms (e.g., limited credit for balance of plant due to limited cable data). The design and plant layout of CPS makes fire propagation to multiple compartments unlikely compared to the fire risk in individual compartments. Therefore, the FPRA is judged adequate for use in the ILRT/DWBT risk evaluation.

• Seismic Fire Interactions (Task 13) - This task reviews previous assessments to identify any specific interaction between suppression system and credited components or adverse impact of fire protection system interactions that should be accounted for in the FPRA. The seismic fire interactions assessment is typically qualitative in nature and therefore would not impact the ILRT/DWBT results.

• Uncertainty and Sensitivity Analysis (Task 15) - This task explores the impacts of possible variation of input parameters used in the development of the model and the inputs to the analysis on the FPRA results. This task is considered qualitatively for the 2014 FPRA update. These analyses do not impact the ILRT/DWBT risk evaluation.

Given the above, the 2014 Clinton FPRA model is judged to provide a meaningful representation of Fire CDF and LERF contribution, and is appropriate for use in risk-informed decision-making, to the extent that these limitations are recognized and addressed in each application, as appropriate (as performed above for the ILRT/DWBT assessment).

A.3.3 Fire PRA Technical Adequacy for Clinton ILRT/DWBT The conservative methodology described in Section 5.7 that uses a multiplier to obtain ILRT/DWBT impact is considered in the evaluation of Fire PRA Technical Adequacy. A review of limitations reported in the Fire PRA Summary and Quantification Notebook A-14 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy

[A-2] is also considered. The Fire PRA technical adequacy is judged to be adequate to support the conclusions found in Section 7.0 of the main body of this report.

A.3.4 Seismic Core Damage Frequency Estimate Technical Adequacy for Clinton ILRT/DWBT As described in NRC memorandum, Safety/Risk Assessment Results for Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants [A-12]:

In accordance with Management Directive (MD) 6.4, Generic Issues Program, a Safety/Risk Assessment panel was established to:

• Determine, on a generic basis, if the risk associated with Generic Issue (GI) 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States (CEUS) on Existing Plants, warrants further investigation for potential imposition as a cost-justified backfit.

• Provide a recommendation regarding the next step (i.e., should the issue continue to the Regulatory Assessment Stage for identification and evaluation of potential generic, cost justified backfits, be dropped due to low risk, or have other actions taken outside the Generic Issues Program

[GIP]).

The panel reviewed available information, including IPEEE information. Also, as noted in Reference [A-12]:

Approximate SCDF estimates were developed using a method which includes integrating the mean seismic hazard curve and the mean plant-level fragility curve for each NPP. This method, developed by Kennedy (1997), is discussed in Section 10.8.9 of AMSE/ANS RA-Sa-2009 and has previously been used by the staff in the resolution of GI-194, Implications of Updated Probabilistic Seismic Hazard Estimates, and during reviews of various risk-informed license amendments. This approach was discussed with EPRI under an NRC-EPRI seismic research memorandum of understanding. EPRI agreed that this is a reasonable approach for evaluating GI-199.

A-15 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy The calculated Clinton seismic CDF was derived using two inputs: The NUREG-0098 Review Level Earthquake (RLE) spectrum and a site-specific uniform hazard spectrum (UHS). The following eight SCDF estimates were developed from each set of seismic hazard curves (NUREG-0098 and UHS) [A-12]:

1. SCDFpga - integration of the pga-based seismic hazard and plant-level fragility curves.

2. SCDF10 - integration of the 10-Hz seismic hazard and plant-level fragility curves.

3. SCDF5 - integration of the 5-Hz seismic hazard and plant-level fragility curves.

4. SCDF1 - integration of the 1-Hz seismic hazard and plant-level fragility curves.

5. SCDFmax - maximum of the SCDFpga, SCDF10, SCDF5, and SCDF1 estimates.

6. SCDFavg - simple average of the SCDFpga, SCDF10, SCDF5, and SCDF1 estimates.

7. SCDFIPEEE - weighted average of the SCDFpga, SCDF10, SCDF5, and SCDF1 estimates, where the weights were obtained from Appendix A of NUREG-1407 (SCDFpga was weighted by one-seventh and the other SCDF estimates were weighted by two-sevenths).

8. SCDFwl - SCDF estimate based on the weakest link model described in Appendix A Clinton Seismic CDF results for the eight categories and two hazard spectrums (16 SCDF values) ranged from 1.1E-06 to 1.7E-05. To conservatively estimate the Clinton Seismic CDF for the ILRT/DWBT Other Hazards impact evaluation (Section 5.7 of the Main Body), the highest SCDF of 1.7E-05 was chosen. This appears to be the most recent and conservative SCDF representing Clinton Stations seismic risk. The Seismic CDF technical adequacy is judged to be adequate to support the conclusions found in Section 7.0 of the main body of this report.

A-16 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.4 IDENTIFICATION OF KEY ASSUMPTIONS The methodology employed in this risk assessment followed the EPRI guidance [A-10]

as previously approved by the NRC. The analysis included the incorporation of several sensitivity studies and factored in the potential impacts from external events in a bounding fashion. None of the sensitivity studies or bounding analyses indicated any source of uncertainty or modeling assumption that would have resulted in exceeding the acceptance guidelines. Since the accepted process utilizes a bounding analysis approach which is mostly driven by CDF contribution that does not already lead to LERF, there are no identified key assumptions or sources of uncertainty for this application (i.e. those which would change the conclusions from the risk assessment results presented here).

A.5

SUMMARY

A PRA technical adequacy evaluation was performed consistent with the requirements of RG-1.200, Revision 2. This evaluation combined with the details of the results of this analysis demonstrate with reasonable assurance that the proposed extension to the ILRT/DWBT interval for CPS Unit 1 to fifteen years satisfies the risk acceptance guidelines in RG 1.174.

The Fire PRA results and Seismic CDF inputs were used to bound external event impacts of the proposed extension. The technical adequacy of the Fire PRA Model and the Seismic CDF input were qualitatively assessed and found to be adequate to support the conclusions found in Section 7.0 of this document.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.6 REFERENCES

[A-1] ASME/American Nuclear Society, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009, March 2009.

[A-2] CPS 2014 Fire PRA Summary and Quantification Notebook, CPS-PSA-021-06, Revision 1, December 2014.

[A-3] NRC Regulatory Guide, RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Rev. 2, March 2009.

[A-4] CPS Station Internal Events PRA, Model of Record 2006C.

[A-5] NEI 00-02, Probabilistic Risk Assessment Peer Review Process Guidance, Rev. A3, March 2000.

[A-6] BWROG PSA Peer Review Certification Implementation Guidelines, January 1997.

[A-7] Clinton PRA Peer Review Report, October, 2000.

[A-8] EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, EPRI 1011989 - NUREG/CR-6850, September 2005.

[A-9] Fire Probabilistic Risk Assessment Methods Enhancements: Supplement 1 to NUREG/CR-6850 and EPRI 1011989, EPRI and NRC, EPRI 1019259, December 2009.

[A-10] Electric Power Research Institute, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325. EPRI TR-1018243, October 2008.

[A-11] Clinton Power Station 2009 PRA Peer Review Report, April, 2010.

[A-12] U.S. Nuclear Regulatory Commission (NRC), Memorandum to Brian W. Sheron, Director Office of Nuclear Regulatory Research, From Patrick Hiland, Chairman, Safety/Risk Assessment Panel for Generic Issue 199: Safety/Risk Assessment Results for Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates In Central and Eastern United States on Existing Plants, Accession Number ML11356A034, September 2, 2010.

[A-13] CPS 2011 Self-Assessment of the PRA Against the ASME PRA Standard Requirements Notebook, CPS-PSA-016, Revision 1, December 2011 A-18 C467150095-12660-12/1/15

300 W'nl1Plcf ,c.1<1 l/v.11 1~1!1" IL 60'>" 'i Exelon Generation ~0 7 tOOO 011 ce RS-16-015 10 CFR 50.90 January 25, 2016 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

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 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 is a request to revise TS 5.5.13, "Primary Containment Leakage Rate Testing Program" to allow for the permanent extension of the Type A Integrated Leak Rate Testing (ILRT) and Type C Leak Rate Testing frequencies.

Specifically, the proposed change will revise CPS TS 5.5.13, by replacing the references to Regulatory Guide (AG) 1.163, "Performance-Based Containment Leak-Test Program," and 10 CFR 50, Appendix J, Option B with a reference to NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, as the documents used by CPS to implement the performance-based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J. This license amendment request (LAA) also proposes an administrative change to TS 5.5.13 to delete the information regarding the performance of the next CPS Type A test to be performed no later than November 2008 as this Type A test has already occurred.

Attachment 1 contains the evaluation of the proposed changes. Attachment 2 provides the marked up TS pages. The marked up TS Bases page is provided in Attachment 3 for information only.

The proposed amendment is risk-informed and follows the guidance in Regulatory Guide 1.174, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis," Revision 2. EGC has performed a CPS-specific evaluation to assess the risk impact of the proposed amendment. A copy of the risk assessment is provided in Attachment 4.

January 25, 2016 U. S. Nuclear Regulatory Commission Page2 The proposed change has been reviewed by the CPS Plant Operations Review Committee, and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed amendment by January 31, 2017 in order to support the extension of the CPS Unit 1 ILRT, which is required to be performed during the outage in the Spring of 2017. Once approved, this amendment shall be implemented within 30 days.

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.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Timothy Byam at (630) 657-2818.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 251h day of January 2016.

t Patrick R. Simpson Manager - Licensing Exelon Generation Company, LLC Attachments:

1) Evaluation of Proposed Change

2) Markup of Technical Specifications Page

3) Markup of Technical Specifications Bases Page

4) Risk Assessment for CPS Regarding the ILRT (Type A) Permanent Extension Request cc: NRC Regional Administrator, Region Ill 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 - Revise Technical Specification Section 5.5.13 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies 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 ECCS Net Positive Suction Head (NPSH) Analysis 3.4 Justification for the Technical Specification Change 3.5 Plant Specific Confirmatory Analysis 3.6 Non-Risk Based Assessment 3.7 Operating Experience 3.8 NRC SE Limitations and Conditions 3.9 Conclusion

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

Page 1 of 52

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 NPF-62 for Clinton Power Station (CPS) Unit 1. The proposed change is a request to revise TS 5.5.13, "Primary Containment Leakage Rate Testing Program" to allow the following:

Increase in the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with Nuclear Energy Institute (NEI) Technical Report NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A and the conditions and limitations specified in NEI 94-01, Revision 2-A.

Adopt an extension of the containment isolation valve (CIV) leakage rate testing (Type C) frequency from the 60 months currently permitted by 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors,"

Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.

Adopt the use of ANSI/ANS 56.8-2002, "Containment System Leakage Testing Requirements."

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

Specifically, the proposed change contained herein would revise CPS TS 5.5.13, by replacing the references to Regulatory Guide (RG) 1.163, "Performance-Based Containment Leak-Test Program," (Reference 1) and 10 CFR 50, Appendix J, Option B with a reference to NEI 94-01, Revision 3-A (Reference 2), and the conditions and limitations specified in NEI 94-01, Revision 2-A (Reference 8), as the documents used by CPS to implement the performance-based leakage testing program in accordance with Option B of 10 CFR 50, Appendix J. This license amendment request (LAR) also proposes an administrative change to TS 5.5.13 to delete the information regarding the performance of the next CPS Type A test to be performed no later than November 2008 as this Type A test has already occurred.

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 Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exceptions:

(1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Page 2 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Type A tests, and (2) NEI 94-01 1995, Section 9.2.3: The first Type A test performed after November 23, 1993 shall be performed no later than November 23, 2008.

The proposed changes to CPS TS 5.5.13 will replace the reference to RG 1.163 with a reference to NEI Topical Report NEI 94-01 Revisions 2-A and 3-A.

The proposed change will revise TS 5.5.13 to state, 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 modified by the following exception: (1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Type A tests."

Markup of TS 5.5.13 is provided in Attachment 2.

Markup of TS Bases 3.6.1.1 is provided in Attachment 3 for information only. contains the plant specific risk assessment conducted to support this proposed change. This risk assessment followed the guidelines of NRC RG 1.174, Revision 2 (Reference

3) and NRC RG 1.200, Revision 2 (Reference 4). The risk assessment concluded that increasing the ILRT on a permanent basis to one-in-fifteen year frequency is considered to represent a small change in the CPS risk profile.

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 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 Page 3 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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 37) 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 38).

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.

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.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 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 Technical Specifications.

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:

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 inch (max.).

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Attachment 1 EVALUATION OF PROPOSED CHANGE 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 drywell is not normally entered during operation, but access is possible during a hot shutdown with the reactor subcritical.

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.

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.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 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 200oF 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.

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.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 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 ECCS Net Positive Suction Head (NPSH) Analysis NPSH available to the ECCS pumps has been determined in accordance with RG 1.1, "Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps." Pressure drop across the suction strainer is based on results from testing and conservative analysis. The vapor pressure for suppression pool water used in NPSH calculations for events where significant debris generation is expected is based on a suppression pool bulk water temperature of 185oF, which is the maximum design temperature of the containment. Analyses show maximum suppression pool temperatures to be less than the containment design temperature of 185oF. For events in which no significant debris generation is expected, NPSH will continue to be evaluated for 212oF suppression pool water temperature.

Containment pressure is assumed to be atmospheric in accordance with RG 1.1 requirements.

3.4 Justification for the Technical Specification Change 3.4.1 Chronology of Testing Requirements of 10 CFR 50, Appendix J The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the TS. Title 10 CFR 50, 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.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50, Appendix J refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.

Also in 1995, RG 1.163 (Reference 1) was issued. The RG endorsed NEI 94-01, Revision 0, (Reference 5) with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) test from three tests in 10 years to one test in 10 years. This relaxation was based on an NRC risk assessment contained in NUREG-1493, (Reference 6) and Electric Power Research Institute (EPRI) TR-104285 (Reference 7) both of which showed that the risk increase associated with extending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval, provisions for extending the test interval an additional 15 months 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 8), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC Safety Evaluation (SE) 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 1). 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.

In 2012, NEI 94-01, Revision 3-A (Reference 2), was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J and includes provisions for extending Type A ILRT intervals to up to 15 years. NEI 94-01 has been endorsed by RG 1.163 and NRC SEs of June 25, 2008 (Reference 9) and June 8, 2012 (Reference 10) as an acceptable methodology for complying with the provisions of Option B in 10 CFR 50, Appendix J. The regulatory positions stated in RG 1.163 as modified by References 9 and 10 are incorporated in 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. Extensions of Type B and Type C test intervals are allowed based upon completion of two consecutive periodic as-found tests where the results of each test are within a licensees allowable administrative limits. Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment airlocks) and up to a maximum of 75 months for Type C tests. If a licensee considers extended test intervals of greater than 60 months for Type B or Type C tested components, the review should include the additional considerations of as-found tests, schedule and review as described in NEI 94-01, Revision 3-A, Section 11.3.2.

The NRC has provided 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:

Page 9 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Section 9.2.3, NEI TR 94-01, Revision 2, states, "Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 15 years based on acceptable performance history." However, Section 9.1 states that the "required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent conditions but should not be used for routine scheduling and planning purposes." The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

NEI 94-01, Revision 3-A, Section 10.1 concerning the use of test interval extensions in the deferral of Type B and Type C LLRTs past intervals of up to 120 months for the recommended surveillance frequency for Type B testing and up to 75 months for Type C testing, states:

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

Notes: For routine scheduling of tests at intervals over 60 months, refer to the additional requirements of Section 11.3.2.

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

The NRC 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.

3.4.2 Current CPS ILRT Requirements 10 CFR 50, Appendix J was revised, effective October 26, 1995, to allow licenses to choose containment leakage testing under either Option A, "Prescriptive Requirements," or Option B, "Performance-Based Requirements." On June 21, 1996 the NRC approved Amendment 105 for CPS (Reference 13) 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:

Page 10 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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.

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.

Current TS 5.5.13 requires that a program be established to comply with the containment leakage rate testing requirements of 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The program is required to be in accordance with the guidelines contained in RG 1.163. RG 1.163 endorses, with certain exceptions, NEI 94-01, Revision 0, as an acceptable method for complying with the provisions of Appendix J, Option B.

RG 1.163, Section C.1 states that licensees intending to comply with 10 CFR 50, Appendix J, Option B, should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01 (Reference 5) rather than using test intervals specified in ANSI/ANS 56.8-1994. NEI 94-01, Section 11.0 refers to Section 9, which states that Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per ten years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage was less than 1.0La (where La is the maximum allowable leakage rate at design pressure). Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine performance shall be at least 24 months.

Adoption of the Option B performance based containment leakage rate testing program altered the frequency of measuring primary containment leakage in Types A, B, and C tests but did not alter the basic method by which Appendix J leakage testing is performed. The test frequency is based on an evaluation of the "as found" leakage history to determine a frequency for leakage testing which provides assurance that leakage limits will not be exceeded. The allowed frequency for Type A testing as documented in NEI 94-01 is based, in part, upon a generic evaluation documented in NUREG-1493. The evaluation documented in NUREG-1493 included a study of the dependence or reactor accident risks on containment leak tightness for differing types of containment types, including a post tensioned, shallow domed concrete containment similar to CPSs containment structures. NUREG-1493 concluded in Section 10.1.2 that reducing the frequency of Type A tests from the original three (3) tests per 10 years to one (1) test per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Types B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements. Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A Page 11 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE testing, NUREG-1493 concluded that increasing the interval between ILRTs is possible with minimal impact on public risk.

3.4.3 CPS 10 CFR 50, Appendix J, Option B Licensing History June 21, 1996 The NRC issued Amendment 105 on June 21, 1996 (Reference 13). The amendment revised the Operating License and TS to implement 10 CFR 50, Appendix J - Option B, by referring to Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program." Specifically, changes were made to paragraph 2.D of the Operating License; TS Section 1.1, "Definitions;"

TS 3.6.1.1, "Primary Containment;" TS 3.6.1.2, "Primary Containment Air Locks;" TS 3.6.1.3, "Primary Containment Isolation Valves (PCIVs);" and TS Section 5.5, "Programs and Manuals."

September 4, 1996 The NRC issued Amendment 106 on September 4, 1996 (Reference 14). The amendment revised Technical Specifications for the drywell to permit bypass testing on a 10-year frequency with increased testing if performance degrades, changed the drywell air lock testing and surveillance requirements, deleted action notes for the drywell air lock and drywell isolation valves when the bypass leakage limit is not met, and deleted the specific leakage limits for the drywell air lock seal.

March 8, 1999 The NRC issued Amendment 121 on March 8, 1999 (Reference 18). The amendment allowed the deferral of the next scheduled local leak rate test for penetration 1MC-042 until the seventh refueling outage.

March 26, 2002 The NRC issued Amendment 145 on March 26, 2002 (Reference 19). The amendment replaced individual main steamline leakage limits with an aggregate leakage limit, revising technical specification surveillance requirement 3.6.1.3.9, which provides leakage rate limits applicable to the main steamline isolation valves.

January 8, 2004 The NRC issued Amendment 160 on January 8, 2004 (Reference 15). The amendment proposed a one-time Technical Specification change to extend the test interval for the next Appendix J Type A test and the next drywell bypass leakage rate test from 10 to 15 years.

September 19, 2005 The NRC issued Amendment 167 on September 19, 2005 (Reference 16). This amendment supported the application of an alternative source term (AST) methodology, in accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.67, "Accident Source Term,"

with the exception that Technical Information Document (TID) 14844, "Calculation of Distance Page 12 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Factors for Power and Test Reactor Sites," used as the radiation dose basis for equipment qualification at CPS.

March 21, 2006 The NRC issued Amendment 173 on March 21, 2006 (Reference 17). This amendment revised TS SR 3.6.1.3.8 to exclude the containment purge valve leakage rates from the summation of secondary containment bypass leakage rates.

November 16, 2006 On November 16, 2006, AmerGen requested an amendment (Reference 21) to TS 3.6.5.1, "Drywell" and 5.5.13, "Primary Containment Leakage Rate Testing Program," to delay the performance of the next primary containment Type A ILRT from the current requirement of "no later than November 23, 2008" to "prior to startup from the C1R12 refueling outage."

April 30, 2007 On April 30, 2007, AmerGen withdrew the Request for Amendment (Reference 22) to TS 3.6.5.1, "Drywell" and 5.5.13, "Primary Containment Leakage Rate Testing Program," dated November 16, 2006.

3.4.4 Integrated Leakage Rate Testing History As noted previously, CPS TS 5.5.13 currently requires Type A, B, and C testing in accordance with RG 1.163, which endorses the methodology for complying with Option B. Since the adoption of Option B, the performance leakage rates are calculated in accordance with NEI 94-01, Section 9.1.1 for Type A testing. Table 3.4.4-1 lists the past Periodic Type A ILRT results for CPS.

Table 3.4.4-1, 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 13 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE (1) Subsequent to the ILRT, a hole was discovered through the containment liner plate. The hole was identified on NCR-40031 and CR 1-86-01-119. This hole was evidently in existence during the ILRT and was created by the removal of temporary Attachment No.

CL-J-12-4. After repair the hole was retested utilizing a leak chase channel and bubbler per procedure XTP-00-07. The retest showed zero leakage. ILRT was re-performed in November 1986.

The results of the last two Type A ILRTs for CPS were less than the maximum allowable containment leakage rate of 0.65 weight%/day. As a result, since both tests were successful, CPS has been placed on an extended ILRT frequency. The current ILRT interval frequency for CPS is 10 years.

3.4.5 Drywell Bypass Leakage Rate Test (DBLRT) History The leaktightness of the drywell is periodically verified by performance of the DBLRT. This test ensures that the measured drywell bypass leakage is bounded by the safety analysis assumptions. The drywell integrity is further verified by a number of additional tests, including drywell airlock door seal leakage tests, overall drywell airlock leakage tests, drywell isolation valve tests and periodic visual inspections of exposed accessible interior and exterior drywell surfaces. Additional confidence that significant degradation in the drywell integrity has not developed is provided by the periodic qualitative assessment of drywell performance. This assessment was credited in the NRC's acceptance of the current performance-based surveillance frequency of 120 months, approved with TS Amendment 106 for CPS (Reference 14).

The DBLRT Surveillance Frequency is controlled under the Surveillance Frequency Control Program (SFCP). The scheduling of TS SR 3.6.5.1.3 in accordance with the SFCP was approved as part of TS Amendment 192 (Reference 36). As defined in CPS TS 5.5.16, "Surveillance Frequency Control Program," changes to the DBLRT frequency listed in the SCFP shall be made in accordance with NEI 04-10, "Risk-Informed Method for control of Surveillance Frequencies," Revision 1. As such, any changes to the DBLRT frequency do not require NRC review and approval. However, the DBLRT has been historically associated with the ILRT frequency because the plant line-ups are similar, the same equipment is used to perform both tests, and EGC intends to extend the frequency associated with the DBLRT as well in accordance with the SFCP. Therefore, in support of this assumption the risk assessment presented in Attachment 4 of this submittal also includes an assessment for extending the DBLRT interval to once in 15 years.

3.5 Plant Specific Confirmatory Analysis 3.5.1 Methodology An evaluation has been performed to provide an assessment of the risk associated with implementing a permanent extension of the CPS containment Type A ILRT interval from ten years to fifteen years. The risk assessment follows the guidelines from NEI 94-01 (Reference 2), the methodology outlined in EPRI TR-104285 (Reference 7), as updated by the EPRI Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (EPRI TR-1018243)

(Reference 11), the NRC regulatory guidance on the use of Probabilistic Risk Assessment Page 14 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE (PRA) findings and risk insights in support of a request for a plants licensing basis as outlined in RG 1.174 (Reference 3), 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. The format of this document is consistent with the intent of the Risk Impact Assessment Template for evaluating extended ILRT intervals provided in the EPRI TR-1018243 (Reference 11).

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 percent per day to 5 percent per day leads to a barely perceptible increase in total population exposure, and increasing the leak rate to 50 percent per day increases the total population exposure by less than 1 percent. 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 methodology to perform the risk assessment. In October 2008, EPRI TR-1018243 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 NRC RG 1.174. 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/DWBT interval extension risk assessment for CPS employs the EPRI 1018243 methodology, with the affected System, Structure, or Component (SSC) being the primary containment boundary.

In the SE issued by NRC letter dated June 25, 2008 (Reference 9), the NRC concluded that the methodology in EPRI TR-1009325, Revision 2, was acceptable for referencing by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.0 of the SE. Table 3.5.1-1 addresses each of the four limitations and conditions for the use of EPRI TR-1009325, Revision 2.

Table 3.5.1-1, EPRI Report No. 1009325 Revision 2 Limitations and Conditions Limitation/Condition (From Section 4.2 of SE) CPS Response

1. The licensee submits documentation CPS PRA technical adequacy is addressed in indicating that the technical adequacy of Section 3.5.2 of this LAR and Attachment 4, their PRA is consistent with the "Risk Impact Assessment of Extending the requirements of RG 1.200 relevant to the CPS ILRT/DWBT Interval," Appendix A, "PRA ILRT extension. Technical Adequacy."

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Attachment 1 EVALUATION OF PROPOSED CHANGE Table 3.5.1-1, EPRI Report No. 1009325 Revision 2 Limitations and Conditions Limitation/Condition (From Section 4.2 of SE) CPS Response 2.a The licensee submits documentation Because the ILRT does not impact CDF, the indicating that the estimated risk increase relevant criterion is LERF. The increase in associated with permanently extending the internal events LERF resulting from a change ILRT surveillance interval to 15 years is in the Type A ILRT interval for the base case small, and consistent with the clarification with corrosion included is 9.81E-09/yr, which provided in Section 3.2.4.5 of this SE. falls within the "very small" change region of the acceptance guidelines in RG 1.174.

If the EPRI Expert Elicitation methodology Class 3a and Class 3b failure probabilities are used, the change is estimated as 1.15E-09/yr, which falls further within the very small change region of the acceptance guidelines in RG 1.174.

2.b Specifically, a small increase in The change in dose risk for changing the population dose should be defined as an Type A ILRT interval from three-per-ten years increase in population dose of less than or to once-per-fifteen-years, measured as an equal to either 1.0 person-rem per year or increase to the total integrated dose risk for 1% of the total population dose, whichever all accident sequences, is 3.80E-03 person-is less restrictive. rem/yr using the EPRI guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr.

The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology. The change in dose risk meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1%

person-rem/yr.

2.c In addition, a small increase in CCFP The increase in CCFP from the three in ten should be defined as a value marginally year interval to one in fifteen years including greater than that accepted in a previous corrosion effects using the EPRI guidance is one-time 15-year ILRT extension requests. 0.44%, which is below the acceptance criteria This would require that the increase in of 1.5%. The increase in CCFP drops to CCFP be less than or equal to 1.5 about 0.05% using the EPRI Expert Elicitation percentage point. methodology.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Table 3.5.1-1, EPRI Report No. 1009325 Revision 2 Limitations and Conditions Limitation/Condition (From Section 4.2 of SE) CPS Response

3. The methodology in EPRI Report No. The representative containment leakage for 1009325, Revision 2, is acceptable except Class 3b sequences used by CPS is 100 La, for the calculation of the increase in based on the recommendations in the latest expected population dose (per year of EPRI report (Reference 20) and as reactor operation). In order to make the recommended in the NRC SE on this topic methodology acceptable, the average leak (Reference 9). It should be noted that this is rate accident case (accident case 3b) more conservative than the earlier previous used by the licensees shall be 100 La industry ILRT extension requests, which instead of 35 La. utilized 35 La for the Class 3b sequences.

4. A licensee amendment request (LAR) is For CPS containment over-pressure is not required in instances where containment relied upon for ECCS performance.

over-pressure is relied upon for ECCS Reference Section 3.3 of this Attachment for performance. details.

3.5.2 Technical Adequacy of the PRA The PRA Technical Adequacy Evaluation is presented in Appendix A, "PRA Technical Adequacy" of Attachment 4 of this submittal. The following is a summary of that evaluation.

3.5.2.1 Demonstrate the Technical Adequacy of the PRA The guidance provided in RG 1.200, Section 4.2 "License Submittal Documentation," indicates that the following items be addressed in documentation submitted to the NRC to demonstrate the technical adequacy of the PRA:

Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the application.

Document peer review findings and observations that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.

Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the RG. Provide justification to show that where specific requirements in the standard are not met, it will not unduly impact the results.

Identify key assumptions and approximations relevant to the results used in the decision-making process.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 3.5.2.2 Technical Adequacy of the PRA Model The risk assessment performed for the ILRT extension request is based on the current Level 1 and Level 2 PRA model. Note that for this application, the accepted methodology involves a bounding approach to estimate the change in the PRA risk metric of LERF from extending the ILRT interval. Rather than exercising the PRA model itself, it involves the establishment of separate evaluations that are linearly related to the plant Core Damage Frequency (CDF) contribution. Consequently, a reasonable representation of the plant CDF that does not result in a LERF does not require that Capability Category II be met in every aspect of the modeling if the Category I treatment is conservative or otherwise does not significantly impact the results.

3.5.2.3 PRA Model Evolution and Peer Review Summary The 2014A version of the CPS PRA model is the most recent evaluation of the Unit 1 risk profile at CPS for internal event challenges. The CPS PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events.

The PRA model quantification process used for the CPS PRA is based on the event tree/fault tree methodology, which is a well-known methodology in the industry.

EGC employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating EGC nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the CPS PRA.

PRA Maintenance and Update The EGC risk management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plant. This process is defined in the Exelon Risk Management program, which consists of a governing procedure and subordinate implementation procedures. The PRA model update procedure delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating EGC nuclear generation sites. The overall Exelon Risk Management program defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.), and for controlling the model and associated computer files.

Plant Changes Not Yet Incorporated into the PRA Model A PRA updating requirements evaluation (URE) documented in the CPS PRA model update tracking database is created for all issues that are identified that could impact the PRA model.

The URE database includes the identification of those plant changes that could impact the PRA model.

A review of the open UREs indicates that there are no plant changes that have not yet been incorporated into the PRA model that would affect this application. UREs are evaluated for potential impact to applications and to the PRA base model results and are classified as High, Medium or Low priority. High priority items could significantly impact applications. Medium priority items are items that are assessed as potentially important to applications and Low Page 18 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE priority items area items that are assessed as not important to applications and likely to have minimal or no numeric impact. There are no open High priority UREs and seventeen UREs identified as Medium priority. The remaining open UREs are low priority, having little or no impact to the PRA results. Medium priority UREs were found not to impact this application.

Low priority UREs were also reviewed and none were found that would impact this application.

Consistency with Applicable PRA Standards The CPS Full Power Internal Events (FPIE) PRA model has undergone several reviews, including a BWROG Peer Review in 2000 (References 12, 23 and 39), UREs were created, and extensive changes to the PRA model were made in updates through 2006. As a result of the extensive changes made to the model, a full Peer review was again performed in 2009 (Reference 31). The results of the 2009 Peer review and the actions taken to address "gaps" identified, best represent the consistency of the model to current PRA standards.

The 2009 Peer Review was conducted using the 2009 ASME/ANS PRA Standard (Reference

30) and the NRCs comments and clarifications contained in RG 1.200, Revision 2. The Peer Review was conducted using the CPS 2006C FPIE PRA model (Reference 35). The general objective for the CPS PRA is to meet Capability Category II. The findings and observations (F&Os) that were identified in the Peer Review were between the 2006C CPS PRA and the requirements for Capability Category II. These F&Os (i.e., both findings and suggestions) were entered into the CPS URE database for tracking purposes. These UREs were used for scoping of the CPS 2011 PRA update.

All "Findings" from the 2009 Peer Review were addressed as part of the 2011 PRA update. The 2009 CPS Peer Review observations were incorporated into the CPS URE database for tracking. All but one of the "Suggestion" observations have been addressed. The observation that remains open recommended addressing environmental conditions for several operator actions credited outside of the Main Control Room. The actions associated with this observation are assigned Human Error Probabilities (HEPs) of 0.9 in the PRA model. It is judged that addressing this observation would have minimal impact to LERF and the ILRT risk assessment.

Following the 2009 Peer review, a self-assessment relative to the combined ASME/ANS PRA Standard (Reference 30) and the NRCs comments and clarifications contained in RG 1.200 was performed as part of the 2011 PRA update (Reference 34). The 2011 Self-Assessment used the 2009 Peer review results as input. Attachment 4 of this submittal, Table A-2 identified gaps to Category II identified in the 2011 self-assessment and the status of those gaps following the 2011 and 2014 PRA updates. Included in the last column of LAR Attachment 4, Appendix A, Table A-2, "Summary of Clinton 2006 PRA Self-Assessment Identified Enhancements," is the URE number, a significance statement and the impact to this ILRT for the gaps that have not been addressed. These gaps are judged to not have an impact on this application as justified in Table A-2.

External Hazards Although EPRI report 1018243 (Reference 20) recommends a quantitative assessment of the contribution of external events (for example, fire and seismic) where a model of sufficient quality exists, it also recognizes that the external events assessment can be taken from existing, Page 19 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE previously submitted and approved analyses or another alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. Based on this, currently available information for external events models was referenced, and a multiplier was applied to the internal events results based on the available external events information. This is further discussed in LAR Attachment 4, Section 5.7. The fire and seismic PRA Technical Adequacy are discussed in additional detail in Attachment 4, Appendix A, Section A.3, "External Hazards."

Identification of Key Assumptions The methodology employed in this risk assessment followed the EPRI guidance (Reference 20) as previously approved by the NRC. The analysis included the incorporation of several sensitivity studies and factored in the potential impacts from external events in a bounding fashion. None of the sensitivity studies or bounding analyses indicated any source of uncertainty or modeling assumption that would have resulted in exceeding the acceptance guidelines. Since the accepted process utilizes a bounding analysis approach which is mostly driven by CDF contribution that does not already lead to LERF, there are no identified key assumptions or sources of uncertainty for this application (i.e. those which would change the conclusions from the risk assessment results presented here).

3.5.2.4 Summary A PRA technical adequacy evaluation was performed consistent with the requirements of RG 1.200, Revision 2. This evaluation combined with the details of the results of this analysis demonstrate with reasonable assurance that the proposed extension to the ILRT interval for CPS Unit 1 to fifteen years satisfies the risk acceptance guidelines in RG 1.174.

The Fire PRA results and Seismic CDF inputs were used to bound external event impacts of the proposed extension. The technical adequacy of the Fire PRA Model and the Seismic CDF input were qualitatively assessed and found to be adequate to support the conclusions found in , Section 7.0 of this submittal.

3.5.3 Summary of Plant-Specific Risk Assessment Results The findings of the CPS, Unit 1 Risk Assessment contained in Attachment 4 confirm the general findings of previous studies that the risk impact associated with extending the ILRT interval from three in ten years to one in 15 years is very small.

Based on the results from Attachment 4, Section 5.0, "Results," and the sensitivity calculations presented in Attachment 4, Section 6.0 "Sensitivities," the following conclusions regarding the assessment of the plant risk are associated with permanently extending the Type A ILRT test frequency to fifteen years:

RG 1.174 (Reference 3) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines "very small" changes in risk as resulting in increases of CDF below 10-6/yr and increases in LERF below 10-7/yr.

"Small" changes in risk are defined as increases in CDF below 10-5/yr and increases in LERF below 10-6/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 Page 20 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE in the Type A ILRT interval for the base case with corrosion included is 9.81E-09/yr, which falls within the "very small" change region of the acceptance guidelines in RG 1.174.

o If the EPRI Expert Elicitation Methodology Class 3a and Class 3b failure probabilities are used, the change is estimated as 1.15E-09/yr, which falls further within the very small change region of the acceptance guidelines in RG 1.174.

The change in dose risk for changing the Type A ILRT interval from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.80E-03 person-rem/yr using the EPRI guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr.

o The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology. The change in dose risk meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr.

The increase in the conditional containment failure frequency from the three in ten-year interval to one in fifteen years including corrosion effects using the EPRI guidance is 0.44%, which is below the acceptance criteria of 1.5% identified in the NRC SER on the issue (Reference 9).

o The increase in CCFP drops to about 0.05% using the EPRI Expert Elicitation methodology. This value meets both of the related acceptance criteria for change in CCFP of less than 1.5%.

To determine the potential impact from other hazard groups, an additional bounding assessment from the risk associated with the other relevant hazard groups for CPS (i.e.,

Seismic, Internal Fire, High Winds/Tornadoes, External Floods, Transportation) utilizing the latest information from various sources was performed. The total increase in LERF due to internal events and other hazard groups is 1.11E-07/yr, which is in Region II (i.e.,

"small" changes) of the RG 1.174 acceptance guidelines. The other acceptance criteria for change in population dose and change in CCFP are also still met when the other hazard groups are considered in the analysis.

Finally, a similar bounding analysis for the other hazard groups indicates that the total LERF from both internal events and the other hazard groups is 2.06E-06/yr, which is less than the RG 1.174 limit of 1E-05/yr given that the LERF is in Region II (i.e., "small" change in risk).

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

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

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Attachment 1 EVALUATION OF PROPOSED CHANGE Reducing the frequency of Type A tests (i.e., 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.

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 ILRTs 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.

Details of the CPS, Unit 1, risk assessment are contained in Attachment 4 of this submittal.

3.6 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.

3.6.1 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 other 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., C1R15) in April 2015.

During outage C1R15, 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 April 2015, there are 2,253.31 pounds of combined Degraded and Unqualified coatings.

Allowed is 20,000 lbs. per Engineering Evaluation EE-00-143, Rev. 0. Therefore, 2,253.31 lbs.

of combined Degraded and Unqualified coatings would not be of concern during a LOCA. The list was compiled from the field notes and a coatings document review.

3.6.2 Inservice Inspection Program (ISI)

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 lSI inspection intervals are displayed in Tables 3.6.2-1 and 3.6.2-2.

Table 3.6.2-1, CPS IWE Examination Schedule 2nd and 3rd Ten-Year Inspection Interval 2nd Interval 9/10/08 to 9/9/11 1st Period C1R12 Page 23 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Table 3.6.2-1, CPS IWE Examination Schedule 2nd and 3rd Ten-Year Inspection Interval 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 (1) 9/10/18 to 9/9/21 1st Period C1R19 C1R20 C1R21 3rd Interval (1) 9/10/21 to 9/9/25 2nd Period C1R22 C1R23 C1R24 C1R25 3rd Interval (1) 9/10/25 to 9/9/28 3rd Period C1R26 C1R27 C1R28 (1) The dates and outages for the 3rd ISI inspection interval are proposed as the 3rd interval inspection plan and schedule have not been developed at this time.

Table 3.6.2-2, CPS IWL Examination Schedule 2nd and 3rd Ten-Year Inspection Interval C1R12 - 2010 C1R15 - 2015 C1R20 - 2020 C1R25 - 2025 C1R30 - 2030 Page 24 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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, 2001 Edition through the 2003 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 2001 Edition through the 2003 Addenda and the ASME Section XI Code shall apply to components and items classified as ASME Code Class MC or ASME Code Class CC.

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 3.6.2-3, Class MC and/or CC Exempt Components Exam Category Item Number Description Applicability to CPS E-A E1.20 Vent System Not Applicable Accessible Surface Areas E1.30 Moisture Barriers Not Applicable L-B L2.10 Tendon Not Applicable L2.20 Wire or Strand Not Applicable L2.30 Anchorage Not Applicable Hardware and Surrounding Concrete L2.40 Corrosion Not Applicable Protection Medium L2.50 Free Water Not Applicable Page 25 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Augmented Inspections Table 3.6.2-4, 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 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 Owners Activity Report-1, 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.

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 owners 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.

3.6.3 Supplemental Inspection Requirements With the implementation of the proposed change, TS 5.5.13 will be revised by replacing the reference to RG 1.163 (Reference 1) with reference to NEI 94-01, Revision 3-A (Reference 2).

This will require 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 26 of 52

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.

For CPS, no additional inspections are required.

3.6.4 Primary Containment Leakage Rate Testing Program - Type B and Type C Testing Program CPS Types B and C testing program requires 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 6), 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 2015 for CPS has shown substantial margin between the actual As-Found (AF) and As-Left (AL) outage summations and the regulatory requirements as described below:

The As-Found minimum pathway leak rate average for CPS shows an average of 7.2% of La with a high of 9.77% La.

The As-Left maximum pathway leak rate average for CPS shows an average of 13.7% of La with a high of 18.39% La.

Table 3.6.4-1 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 3.6.4-1, CPS Type B and C LLRT Combined As-Found/As-Left Trend Summary RFO C1R11 C1R12 C1R13 C1R14 C1R15 2008 2010 2011 2013 2015 As-Found Min 11597.7 18425.3 31551.3 33798.1 35305.4 Path (sccm)

Percentage 3.21 5.1 8.73 9.35 9.77 of La As-Left Max Path 24924.6 36848.3 52689.2 66201.4 66465.1 (sccm)

Percentage 6.89 10.19 14.58 18.32 18.39 of La 3.6.5 Type B and Type C Local Leak Rate Testing Program Implementation Review Table 3.6.5-1 identifies the components that were on extended intervals and have not demonstrated acceptable performance during the previous two outages for CPS.

Table 3.6.5-1, CPS Type B and C LLRT Program Implementation Review C1R14 - 2013 Component As- Evaluation As-left Cause of Corrective Scheduled found Limit SCCM Failure Action Interval SCCM SCCM 1IA175 - IA 56000 500 1680 Seat Replaced 30 months Instrument Air (1) leakage Valve Isolation Check Valve to 1IA006 1MC057 Page 28 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE C1R15 - 2015 Component As- Evaluation As-left Cause of Corrective Scheduled found Limit SCCM Failure Action Interval SCCM SCCM 1RF021 - 22000 10000 695 Intermediate Refurbished 30 months Containment position actuator Building floor Drain indication (2)

Inboard Isolation (2)

Control Valve 1MC070 1VR006B - 24980 500 6500 Seat Replaced 30 months Continuous leakage valve.

Containment Purge (3) Accepted As-Inboard Isolation Left above Valve admin limit by 1MC113 evaluation (1) 1IA175 failed it's as-found LLRT with 56000 sccm. However, the as-found, Minimum Pathway, leakage for penetration 1MC057 was 704.25 sccm. As-left value of 1680 sccm was determined to be acceptable.

(2) 1RF021 indicated intermediate when attempting to close. When the test volume was pressurized to test pressure, air flow was felt on the outlet of the test vent valve 1RF030B. This intermediate indication was the most probable cause of the excessive leakage. Resolved intermediate position indication in 1RF021 and re-tested.

(3) 1VR006B failed its as-found LLRT with 24980 sccm. This is added to the Max-Path leakage. This is also a secondary containment bypass penetration so this leakage applies to both the 0.6La total and the 0.08La. The previous test results for this penetration was 6,904 sccm. Following maintenance, the As-Left Maximum Pathway leakage for penetration 1MC113 was tested as 6500 sccm. This value is 94.2% of the 1MC113 As-left Max Pathway leakage measured in C1R12 which had a value of 6900 sccm (6500 / 6900 = 94.2%). The leakage identified in IR 2501033 for 1MC113, was determined to be an improvement of past leakage for the penetration and secondary containment bypass as a whole. Because of this improvement Engineering accepted the as-found test results of 1MC113 and the impact on 0.08La (secondary containment bypass), and 0.6La (containment leakage). No further action was required.

The percentage of the total number of CPS Type B tested components that are on 120-month extended performance-based test intervals is 96%.

The percentage of the total number of CPS Type C tested components that are on 60-month extended performance-based test intervals is 90%.

3.7 Operating Experience During the conduct of the various examinations and tests conducted in support of the Containment related programs previously mentioned, issues that do not meet established Page 29 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE criteria or that provide indication of degradation, are identified, placed into the site's corrective action program, and corrective actions are planned and performed.

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

IN 1992-20, "Inadequate Local Leak Rate Testing" IN 2010-12, "Containment Liner Corrosion" IN 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" Each of these areas is discussed in detail in Sections 3.7.1 through 3.7.3, respectively.

3.7.1 IN-92-20, "Inadequate Local Leak Rate Testing" The issue discussed in IN-92-20, "Inadequate Local Leak Rate Testing," was based on events at four different plants, Quad Cities Nuclear Power Station, Dresden Nuclear Power Station, Perry Nuclear Plant and the Clinton Power Station. The common issue in the four events was the failure to adequately perform local leak rate testing on different penetration configurations leading to problems that were discovered during ILRT tests in the first three cases.

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

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

In the event at CPS, relief valve discharge lines that were assumed to terminate below the suppression pool minimum drawdown level were discovered to terminate at a level above that datum. These lines needed to be reconfigured and the valves should have been Type C LLRT tested. To correct this problem, CPS removed the vacuum breaker connections and the flanges and extended the pipes to ensure that a water seal would be maintained.

As for the testing of two-ply stainless steel bellows at CPS, a modification (i.e., FH-030) was installed on the Inclined Fuel Transfer System (IFTS) containment penetration in 1995 to eliminate the concern raised by IN-92-20 and to allow bellows testing to be performed using Type B test methods. The testing assembly provides a means of applying a static test pressure to the bellows to ensure containment integrity will be maintained in accordance with 10 CFR 50, Appendix J.

There are no mechanical bellows consistently exceeding the administrative limit of 500 sccm at CPS.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 3.7.2 IN 2010-12, "Containment Liner Corrosion" This IN was issued to alert plant operators to three events that occurred where the steel liner of the containment building was corroded and degraded. At Beaver Valley and Brunswick plants material had been found in the concrete which trapped moisture against the liner plate and corroded the steel. In one case it was material intentionally placed in the building and in the other case it was foreign material which had inadvertently been left in the form when the wall was poured. But the result in both cases was that the material trapped moisture against the steel liner plate leading to corrosion. In the third case, an insulating material placed between the concrete floor and the steel liner plate at Salem adsorbed moisture and led to corrosion of the liner plate.

The situation that occurred at Salem is not likely to take place at CPS. CPS does not have moisture barriers.

CPS should not experience the events that took place at Beaver Valley and Brunswick. EGC has implemented periodic examinations during refueling outages on metallic containment structures or liners in accordance with the Section XI, Subsection IWE. The applicable EGC visual examination procedure, ER-AA-335-018, requires the conditions described in the IN examples to be recorded. Conditions that may affect containment surface integrity are then required to be evaluated by engineering evaluation or repair/replacement prior to startup from refueling outages.

3.7.3 IN 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" The containment basemat metallic shell and liner plate seam welds of pressurized water reactors are embedded in a 3-to 4-feet thick concrete floor during construction and are typically covered by a leak-chase channel system that incorporates pressurizing test connections. This system allows for pressure testing of the seam welds for leak-tightness during construction and also in service, as required. A typical basemat shell or liner weld leak-chase channel system consists of steel channel sections that are fillet welded continuously over the entire bottom shell or liner seam welds and subdivided into zones, each zone with a test connection.

Each test connection consists of a small carbon or stainless steel tube (less than 1-inch diameter) that penetrates through the back of the channel and is seal-welded to the channel steel. The tube extends up through the concrete floor slab to a small steel access (junction) box embedded in the floor slab. The steel tube, which may be encased in a pipe, projects up through the bottom of the access box with a threaded coupling connection welded to the top of the tube, allowing for pressurization of the leak-chase channel. After the initial tests, steel threaded plugs or caps are installed in the test tap to seal the leak-chase volume. Gasketed cover plates or countersunk plugs are attached to the top of the access box flush with the containment floor. In some cases, the leak-chase channels with plugged test connections may extend vertically along the circumference of the cylindrical containment shell or liner to a certain height above the floor.

The CPS IWE/IWL Containment Inspection Program indicates the applicable requirements are ASME Section XI Code 2001 Edition through 2003 Addenda, and 10CFR50.55a requirements.

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Attachment 1 EVALUATION OF PROPOSED CHANGE A review of the program implementation document determined that CPS has weld leak chase design and test connections for containment liner plate seam welds. These are designed to test the enclosed seam welds for leakage. The design of the containment includes stainless steel liners. Therefore, the concern with corrosion is very unlikely and inspection is not required.

However, the seam welds are tested as part of the containment leakage test boundary per CPS 9861.01, "Integrated Leak Rate Test," by venting the leak chase channels by removal of leak chase channel plugs.

3.7.4 Results of Recent Inspections 3.7.4.1 Containment and Drywell Coatings - C1R15 - 2015 3.7.4.1.1 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 has 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. Overall the coatings at the concrete wall, liner plate, and existing steel are in good condition. In some areas, the coating on the concrete basement floor is in poor condition. A condition report, (AR 02498541) has been generated for the identified condition. Areas of mechanical damage to concrete substrate are present throughout the drywell floor. A large area (approximately 250 sq. ft.), is scheduled to be repaired refueling outage C1R17 (Spring 2017).

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.

3.7.4.1.2 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 liner plate on Elevation 737' is approximately 15 feet away. On this elevation, there are areas of Page 32 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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', one area of mechanical damage on the inner concrete wall exhibits spalling with rebar exposed. Overall for these elevations, the liner plate and inner wall are in good condition.

828' Elevation Most of the containment floor is covered with plastic for FME and to facilitate refuel services activities and as a result, it was not feasible to inspect. The dome was remotely observed from outside of the contaminated area. There were no concerns or issues about coatings degradation on this elevation.

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.

3.7.4.1.3 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.

3.7.4.2 IWE - C1R14 - 2013 3.7.4.2.1 Containment In-Service Inspections (CISI) on Containment liner 737' Elevation CISI was performed on containment liner on the 737' elevation during C1R14 for ASME Section XI IWE inspection requirements. The following observations document the C1R14 as-found reportable indications.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 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 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.

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

755' Elevation CISI were performed on containment liner on the 755' elevation during C1R14 for ASME Section XI IWE inspection requirements. The following observations document the C1R14 as-found reportable indications.

At 24 degrees, 776' elevation, 1" x 2" peeled, cracked and stained coatings 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 were performed on containment liner on the 781' elevation during C1R14 for ASME Section XI IWE inspection requirements. The following observations document the C1R14 as-found reportable indications.

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 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.

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.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 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 were performed on containment liner on the 800' elevation during C1R14 for ASME Section XI IWE inspection requirements. The following observations document the C1R14 as-found reportable indications.

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

At 85 degrees, 815' elevation, 8" x 8" area had 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 190 degrees, 823' elevation, stained coatings 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.

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

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

3.7.4.3 IWE/IWL - C1R15 - 2015 3.7.4.3.1 Inspection Scope In accordance with the 2nd 10-Year Inspection Interval, the scheduled examinations for ASME Section XI, IWE/IWL Containment were completed during outage C1R15. The scope of inspection as listed in the Second Inspection Period in the current Interval was completed.

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Attachment 1 EVALUATION OF PROPOSED CHANGE 3.7.4.3.2 Results This inspection was conducted for all accessible areas in accordance with ER-AA-335-018, "Visual Examination of ASME /WE Class MC and Metallic Liners of IWL Class CC Components." Thirty reports were created to organize the inspections. These reports are numbered C1R15-001 to C1R15-030, and the corresponding areas in each report are listed below. Inspector observations and all recordable indications are listed in this summary. In general, the inspections revealed these common observations of minor significance:

deteriorated caulking, as identified in reports C1R15-022, -024, and -029; flaking/peeling paint, coating damage, missing coating, light rust, light corrosion. There was only one area where the inspector noted the presence of medium to heavy rust but no apparent loss of material (Report No. C1R15-029). This location corresponds to Penetration 1K3E (OD). A picture of this observation is included in the report. Deteriorated coating repairs will be implemented as required. Deteriorated coatings were identified in reports C1R15-020, -025, -028, and -029. In cases where no penetration pictures were taken, the recorded conditions do not indicate loss in material and are thereby acceptable. This statement applies to reports C1R15-021, C1R15-026 and C1R15-027.

General Visual Examinations were conducted by VT-3 examiners to assess the general condition of the surface or component. Below is a listing of individual observations noted for each area and penetrations inspected:

Report No. C1R15-001 - Containment Steam Tunnel Concrete and liner inspections were satisfactory except for this recordable indication: (1) Flaking paint, light rust, coating damaged all around penetration 1MC-45 on the liner.

Report No. C1R15-002 - Containment Liner - 828' and Dome 0 - 90 deg. Satisfactory. No recordable indication.

Report No. C1R15-003 - Containment Liner - 828' and Dome 90 - 180 deg. Satisfactory. No recordable indication.

Report No. C1R15-004 - Containment Liner - 828' and Dome 180 - 270 deg. Satisfactory. No recordable indication.

Report No. C1R15-005 - Containment Liner - 828' and Dome 270 - 360 deg. Satisfactory. No recordable indication.

Report No. C1R15-006 - OD Concrete. 707' AB, 270 - 360 deg.

Concrete coatings were satisfactory except for two areas: (1) Az. 285° El. 707' had flaking and peeling paint. (2) Az. 293° El. 710' has peeling paint.

Report No. C1 R15-007- OD Concrete. 712' FB.90-180 deg. Satisfactory. No recordable indication.

Report No. C1R15-008 - OD Concrete. Reactor Water Cleanup Mezz. 750' AB. 40 - 90 deg.

Satisfactory. No recordable indication.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Report No. C1R15-009 - OD Concrete. FB 737', 90 - 180 deg. Satisfactory. No recordable indication.

Report No. C1 R15-010 - OD Concrete. AB 737'. 270- 90 deg. Satisfactory. No recordable indication.

Report No. C1R15-011 - OD Concrete. AB 762'. 270 - 90 deg. Concrete and coatings were satisfactory except for two areas: (1) Az. 295° El. 770' had flaking paint. (2) Az. 285° El. 770' has flaking paint.

Report No. C1 R15-012-OD Concrete. FB 755', 90 -270 deg. Satisfactory. No recordable indication.

Report No. C1R15-013 - OD Concrete. AB 781', and AB Steam Tunnel. Upper Aux. Steam Tunnel concrete is satisfactory. Typical pictures above the main steam system penetrations.

Recordable indications are: (1) missing surface coating and coating damage between feedwater penetrations and around penetration 1MC-14, apparent moisture damage. (2) Flaking paint on Az. 80°, El. 780' and Az. 85°, El. 785'.

Report No. C1 R15-014-OD Concrete. FB 781', 90 -270 deg. Satisfactory. No recordable indication.

Report No. C1 R15-015 - OD Concrete. Gas Control Boundary 800' and above AB & FB 0-360 deg. Satisfactory. No recordable indication.

Report No. C1R15-016 - Containment Equipment. Hatch Satisfactory. No recordable indication.

Report No. C1 R15-017 - Containment Personnel Lock- 741'. Satisfactory. No recordable indication.

Report No. C1R15-018 - Containment Personnel Lock - 832'. Satisfactory. No recordable indication.

Report No. C1R15-019 - Fuel Transfer Tube 755' FB and 770' Containment Building The outboard side was underwater and inaccessible. The inboard side had recordable indications: (1) Light corrosion noted on weld-o-lets. No material loss. No change from previous report. (2) Missing/incomplete coating on upper transfer tube area. No change noted from previous report.

Report No. C1R15-020 - Penetrations ID & OD Above 762', 0 - 90 deg. There were no recordable indications for all ID side inspections. The OD side inspection found recordable indications: (1) Flaking paint, moderate rust, no apparent material loss for penetrations 1MC-60, 1MC-61, 1MC-64, 1MC-65, 1MC-74, 1MC-89, 1MC210, and 1MC-211.

Report No. C1R15-021 - Penetrations ID & OD Below 737', 37 - 355 deg. There were no recordable indications for all ID side inspections. The OD side inspection found recordable indications: (1) Flaking paint, light rust, no apparent material loss for penetrations 1MC-12, and Page 37 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 1MC-13. This report only contains inspector comments on the recordable indications.

Inadvertently, the inspectors did not submit pictures for this report.

Report No. C1R15-022 - Penetrations ID & OD Above 737', 270 - 360 deg. There were no recordable indications for all ID side inspections. The OD side inspection found recordable indications: (1) Flaking and cracking paint, and light rust for penetrations 1MC-17, 1MC-20, and 1MC-29. Also, penetration 1MC-20 shows signs of caulking deterioration.

Report No. C1R15-023 - Penetrations ID & OD Above 737', 0 - 90 deg. There were no recordable indications for all ID side inspections. Inspector noted that one penetration, 1MC-77, was not accessible from the OD side, and that it was previously noted in the last inspection in C1R11. The OD side inspection found recordable indication: (1) Flaking paint and light rust for penetrations 1MC-17, 1MC-20, and 1MC-29.

Report No. C1R15-024 - Penetrations OD Above 737', 90 - 180 deg. Inspections of all penetrations were satisfactory, except for two: (1) Penetration 1MC-46 (OD) had deteriorated caulking, flaking paint and missing coating, (2) Penetration 1MC-47 (OD) had flaking paint and missing coating.

Report No. C1R15-025 - Penetrations OD Above 737', 180-360 deg. Inspections of all penetrations were satisfactory, except for three: (1) Penetration 1MC-35 (OD) had flaking paint, moderate rust, but no apparent material loss. (2) Penetration 1MC-76 (OD) had flaking paint and light rust. (3) Penetration 1MC-78 (OD) had flaking paint and light rust.

Report No. C1R15-026 - Penetrations OD Above 762', 120 - 277 deg. Inspections of all penetrations were satisfactory, except for three: (1) Penetration 1MC-102 (OD) had flaking paint, light rust, but no apparent loss of material, (2) Penetration 1MC-103 (OD) had flaking paint, light rust, but no apparent loss of material, (3) Penetration 1MC-104 (OD) had flaking paint, light rust, but no apparent loss of material. This report contains inspector comments and pictures for recordable indications in three locations. Inadvertently, the inspectors did not submit picture for one location.

Report No. C1R15-027 - Penetrations OD Above 781', 90 - 180 deg. (1 Penetration both ID &

OD). Inspections of all penetrations were satisfactory, except for four: (1) Penetration 1MC-50 (OD) had light rust but no apparent loss of material. (2) Penetration 1MC-51 (OD) had light rust but no apparent loss of material. (3) Penetration 1MC-155 (OD) had flaking paint and light rust but no apparent loss of material. (4) Penetration 1MC-170 (OD) had flaking paint, light rust but no apparent loss of material. This report contains inspector comments for recordable indications on four locations. Inadvertently, the inspectors only provided picture for one location.

Report No. C1R15-028 - Penetrations OD Above 762', 180 - 360 deg. Inspections of all penetrations were satisfactory, except for six. (1) Penetration 1MC-166 (OD) had flaking and peeling paint. (2) Penetration 1MC-168 (OD) had flaking paint, medium corrosion and rust, but no apparent loss of material. (3) Penetration 1MC-171 (OD) had flaking paint, corrosion, no apparent loss of material. (4) Penetration 1MC-203 (OD) had flaking and peeling paint, corrosion, but no apparent loss of material. (5) Penetration 1MC-204 (OD) had flaking paint, corrosion, but no apparent loss of material. (6) Penetration 1MC-204 (OD) had flaking and peeling paint.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Report No. C1R15-029 - Penetrations OD Above 762' & 781', 0 - 180 deg. Inspections of all penetrations were satisfactory, except for four: (1) Penetration 1K1E (OD) had discoloration and light rust. (2) Penetration 1K3N (OD) had flaking paint and light rust but no apparent material loss. (3) Penetration 1P1E-2 (OD) had rust but no apparent material loss. (4) Penetration 1K3E (OD) had coating degradation, medium to heavy rust, but no apparent loss of material.

Report No. C1R15-030 - Penetrations OD Above 762' & 781', 180 - 360 deg. Satisfactory. No recordable indication.

In summary, Containment ISI inspections completed in C1R15 identified minor surface conditions such as flaking paint, peeling paint, rust, light corrosion, coating damage, missing coating, and incomplete coating. However, none of the examinations found any loss of metal and none were of concern to the containment function.

3.8 NRC SE Limitations and Conditions 3.8.1 Limitations and Conditions Applicable to NEI 94-01, Revision 2-A The NRC staff found that the use of NEI TR 94-01, Revision 2, was acceptable for referencing by licensees proposing to amend their TSs to permanently extend the ILRT surveillance interval to 15 years, provided the following conditions as listed in Table 3.8.1-1 were satisfied:

Table 3.8.1-1: NEI 94-01, Revision 2-A, Limitations and Conditions Limitation/Condition (From Section 4.0 of SE) CPS Response For calculating the Type A leakage rate, the CPS will utilize the definition in NEI 94-01 licensee should use the definition in the NEI Revision 3-A, Section 5.0. This definition has TR 94-01, Revision 2, in lieu of that in remained unchanged from Revision 2-A to ANSI/ANS-56.8-2002. (Refer to SE Revision 3-A of NEI 94-01.

Section 3.1.1.1.)

The licensee submits a schedule of Reference Section 3.6.2 and Tables 3.6.2-1 containment inspections to be performed and 3.6.2-2.

prior to and between Type A tests. (Refer to SE Section 3.1.1.3.)

The licensee addresses the areas of the Reference Section 3.6.2 of this submittal.

containment structure potentially subjected to degradation. (Refer to SE Section 3.1.3.)

The licensee addresses any tests and There are no major modifications planned.

inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4.)

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Attachment 1 EVALUATION OF PROPOSED CHANGE Table 3.8.1-1: NEI 94-01, Revision 2-A, Limitations and Conditions Limitation/Condition (From Section 4.0 of SE) CPS Response The normal Type A test interval should be CPS will follow the requirements of NEI 94-01 less than 15 years. If a licensee has to utilize Revision 3-A, Section 9.1. This requirement the provision of Section 9.1 of NEI TR 94-01, has remained unchanged from Revision 2-A Revision 2, related to extending the ILRT to Revision 3-A of NEI 94-01.

interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an In accordance with the requirements of NEI unforeseen emergent condition. (Refer to SE 94-01 Revision 2-A, SER Section 3.1.1.2, Section 3.1.1.2.) CPS will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

For plants licensed under 10 CFR 52, Not applicable. CPS was not licensed under applications requesting a permanent 10 CFR 52.

extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No.

1009325, Revision 2, including the use of past containment ILRT data.

3.8.2 Limitations and Conditions Applicable to NEI 94-01, Revision 3-A The NRC staff found that the guidance in NEI TR 94-01, Revision 3, was acceptable for referencing by licensees in the implementation for the optional performance-based requirements of Option B to 10 CFR 50, Appendix J. However, the NRC staff identified two conditions on the use of NEI TR 94-01, Revision 3 (Reference NEI 94-01 Revision 3-A, NRC SE 4.0, Limitations and Conditions):

Topical Report Condition 1 NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI TR 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR MSIVs), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.

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Attachment 1 EVALUATION OF PROPOSED CHANGE Response to Condition 1 Condition 1 presents three separate issues that are required to be addressed. They are as follows:

ISSUE 1 - The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.

ISSUE 2 - In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.

ISSUE 3 - Use of the allowed 9-month extension for eligible Type C valves is only authorized for non-routine emergent conditions with exceptions as detailed in NEI 94-01, Revision 3-A, Section 10.1.

Response to Condition 1, ISSUE 1 The post-outage report shall include the margin between the Type B and Type C Minimum Pathway Leak Rate (MNPLR) summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.60 La.

Response to Condition 1, ISSUE 2 When the potential leakage understatement adjusted Type B and C MNPLR total is greater than the CPS leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the CPS leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.

Response to Condition 1, ISSUE 3 CPS will apply the 9-month allowable interval extension period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.

Topical Report Condition 2 The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending Page 41 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 refueling outages. Type C tests involve valves, which in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for.

Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.

When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total leakage, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Response to Condition 2 Condition 2 presents two separate issues that are required to be addressed. They are as follows:

ISSUE 1 - Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.

ISSUE 2 - When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Response to Condition 2, ISSUE 1 The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25%

in the LLRT periodicity. As such, CPS will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate, which will increase the As-Left leakage total for each Type C component currently on greater than a 60-month test interval up to the 75-month extended test interval. This will result in a combined conservative Page 42 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Type C total for all 75-month LLRTs being "carried forward" and will be included whenever the total leakage summation is required to be updated (either while on line or following an outage).

When the potential leakage understatement adjusted leak rate total for those Type C components being tested on greater than a 60-month test interval up to the 75-month extended test interval is summed with the non-adjusted total of those Type C components being tested at less than or equal to a 60-month test interval, and the total of the Type B tested components, if the MNPLR is greater than the CPS leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the CPS leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

Response to Condition 2, ISSUE 2 If the potential leakage understatement adjusted leak rate MNPLR is less than the CPS leakage summation limit of 0.50 La, then the acceptability of the greater than a 60-month test interval up to the 75-month LLRT extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

In addition to Condition 1, ISSUES 1 and 2, which deal with the MNPLR Type B and C summation margin, NEI 94-01, Revision 3-A, also has a margin related requirement as contained in Section 12.1, "Report Requirements."

A post-outage report shall be prepared presenting results of the previous cycles Type B and Type C tests, and Type A, Type B and Type C tests, if performed during that outage. The technical contents of the report are generally described in ANSI/ANS-56.8-2002 and shall be available on-site for NRC review. The report shall show that the applicable performance criteria are met, and serve as a record that continuing performance is acceptable. The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit. Adverse trends in the Type B and Type C leakage rate summation shall be identified in the report and a corrective action plan developed to restore the margin to an acceptable level.

At CPS, in the event an adverse trend in the aforementioned potential leakage understatement adjusted Type B and C summation is identified, then an analysis and determination of a corrective action plan shall be prepared to restore the trend and associated margin to an acceptable level. The corrective action plan shall focus on those components which have contributed the most to the adverse trend in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

At CPS an adverse trend is defined as three (3) consecutive increases in the final pre-mode change Type B and C MNPLR leakage summation values, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.

Page 43 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 3.9 Conclusion NEI 94-01, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, describe an NRC-accepted approach for implementing the performance-based requirements of 10 CFR 50, Appendix J, Option B. It incorporated the regulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals to 15 years and Type C test intervals to 75 months. NEI 94-01, Revision 3-A delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies. CPS is adopting the guidance of NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, for the CPS 10 CFR 50, Appendix J testing program plan.

Based on the previous ILRTs conducted at CPS, it may be concluded that the permanent extension of the containment ILRT interval from 10 to 15 years represents minimal risk to increased leakage. The risk is minimized by continued Type B and Type C testing performed in accordance with Option B of 10 CFR 50, Appendix J, Drywell Inspections and the overlapping inspection activities performed as part of the following CPS inspection programs:

• Inservice Inspection Program IWE/IWL

• Containment Inspections per TS SR 3.6.1.1.1

• Protective Coatings Program This experience is supplemented by risk analysis studies, including the CPS risk analysis provided in Attachment 4. The risk assessment concluded that increasing the ILRT interval on a permanent basis to a one-in-fifteen year frequency is not considered to be significant since it represents only a small change in the CPS risk profile.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. 10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR 50, "Leakage Rate Testing of Containment of Water Cooled Nuclear Power Plants." Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.

The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B and Type C testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of 10 CFR 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that Page 44 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type C test frequencies will not directly result in an increase in containment leakage.

EPRI TR-1009325, Revision 2, provided a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance.

NEI 94-01, Revision 3-A, Section 9.2.3.1 states that Type A ILRT intervals of up to 15 years are allowed by this guideline. The Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI Report 1018243 (Formerly TR-1009325, Revision 2) indicates that, in general, the risk impact associated with ILRT interval extensions for intervals up to 15 years is small. However, plant-specific confirmatory analyses are required.

The NRC staff reviewed NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2.

For NEI TR 94-01, Revision 2, the NRC staff determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163. The NRC staff finds that the Type A testing methodology as described in ANSI/ANS-56.8-2002, and the modified testing frequencies recommended by NEI TR 94-01, Revision 2, serves to ensure continued leakage integrity of the containment structure. Type B and Type C testing ensures that individual penetrations are essentially leak tight. In addition, aggregate Type B and Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.

For EPRI Report No. 1009325, Revision 2, a risk-informed methodology using plant-specific risk insights and industry ILRT performance data to revise ILRT surveillance frequencies, the NRC staff finds that the proposed methodology satisfies the key principles of risk-informed decision making applied to changes to TSs as delineated in RG 1.177 and RG 1.174. The NRC staff, therefore, found that this guidance was acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.2 of the Safety Evaluation (SE).

The NRC staff reviewed NEI TR 94-01, Revision 3, and determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, as modified by the conditions and limitations summarized in Section 4.0 of the associated Safety Evaluation. This guidance included provisions for extending Type C LLRT intervals up to 75 months. Type C testing ensures that individual containment isolation valves are essentially leak tight. In addition, aggregate Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The NRC staff, therefore, found that this guidance, as modified to include two limitations and conditions, was acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing. Any applicant may reference NEI TR 94-01, Revision 3, as modified by the associated SE and approved by the NRC, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, in a licensing action to satisfy the requirements of Option B to 10 CFR 50, Appendix J.

Page 45 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 4.2 Precedent This LAR is similar in nature to the following license amendments to extend the Type A Test Frequency to 15 years and the Type C test frequency to 75 months as previously authorized by the NRC:

Surry Power Station, Unit 1 (Reference 24)

Donald C. Cook Nuclear Plant, Unit 1 (Reference 25)

Beaver Valley Power Station, Unit Nos. 1 and 2 (Reference 26)

Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (Reference 27)

Peach Bottom Atomic Power Station, Units 2 and 3 (Reference 28)

Comanche Peak Nuclear Power Plant, Units 1 and 2 (Reference 40) 4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (EGC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

The proposed activity involves the extension of the Clinton Power Station (CPS), Unit 1, Type A containment test interval to 15 years, and the extension of the Type C test interval to 75 months. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test. The current Type C test interval of 60 months for selected components would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident.

The change in dose risk for changing the Type A Integrated Leak Rate Test (ILRT) interval from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.80E-03 person-rem/yr using the EPRI guidance with the base case corrosion included. This change meets both of the related acceptance criteria for change in population dose of less than 1.0 person-rem/yr or less than 1% person-rem/yr. The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology. The change in dose risk meets both of the related acceptance for change in population dose of less than 1.0 person-rem/yr or less Page 46 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE than 1% person-rem/yr. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

In addition, as documented in NUREG-1493, Types B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The CPS, Unit 1 Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) activity based, and, (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance.

Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with American Society of Mechanical Engineers (ASME) Section XI, and Technical Specifications (TS) requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for CPS. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how CPS is operated.

Therefore, the proposed change does not result in 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 amendment to the TS 5.5.13, "Primary Containment Leakage Rate Testing Program," involves the extension of the CPS, Unit 1 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plants ability to mitigate the consequences of an accident.

The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) nor does it alter the design, configuration, or change the manner in which the plant is operated or controlled beyond the standard functional capabilities of the equipment.

The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for CPS. This exception was for an activity that has already taken place; therefore, this deletion is solely an administrative action that does not result in any change in how CPS is operated.

Page 47 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The proposed amendment to TS 5.5.13 involves the extension of the CPS, Unit 1 Type A containment test interval to 15 years and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.

The proposed change involves the extension of the interval between Type A containment leak rate tests and Type C tests for CPS, Unit 1. The proposed surveillance interval extension is bounded by the 15-year ILRT interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Type B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section Xl, and TS serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing. The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained.

The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A and Type C test intervals.

The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for CPS, Unit 1. This exception was for an activity that has taken place; therefore, the deletion is solely an administrative action and does not change how CPS is operated and maintained. Thus, there is no reduction in any margin of safety.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed Page 48 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

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

September 1995

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

3. Regulatory Guide 1.174, Revision 2, "An Approach For Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis," May 2011

4. Regulatory Guide 1.200, Revision 2, "An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities," March 2009

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

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

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

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

9. 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 Page 49 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE 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

10. 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

11. 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

12. Clinton PRA Peer Review Report, October 2000

13. 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

14. Letter from D. Pickett (NRC) to M. Lyon (CPS), Clinton Power Station Unit 1 - Issuance of Amendment 106 Regarding Revision of Technical Specifications for the Drywell to Permit Bypass Testing On a 10-year Frequency (TAC NO. M94889), dated September 4, 1996

15. Letter from D. Pickett (NRC) to J. Skolds (AmerGen), Clinton Power Station Unit 1 -

Issuance of Amendment 160 Regarding The One-Time Technical Specification Change To Extend The Test Interval For The Next Appendix J Type A Test And The Next Drywell Bypass Leakage Rate Test From 10 To 15 Years (TAC NO. MB7675), dated January 8, 2004

16. Letter from K. Jabbour (NRC) to C. Crane (AmerGen), Clinton Power Station Unit 1 -

Issuance of Amendment 167 Regarding the Application of Alternative Source Term Methodology (TAC NO. MB8365), dated September 19, 2005

17. Letter from K. Jabbour (NRC) to C. Crane (AmerGen), Clinton Power Station Unit 1 -

Issuance of Amendment 173 Regarding the Revision Of Secondary Containment Bypass Leakage Surveillance Requirement (TAC NO. MC6488), dated March 21, 2006

18. Letter from J. Hopkins (NRC) to J. Sipek (CPS), Clinton Power Station, Unit 1 - Issuance of Amendment 121 Regarding the Deferral of The Next Scheduled Local Leak Rate Test for Valve 1MC-042 Until The Seventh Refueling Outage (TAC NO. MA3754), dated March 8, 1999

19. Letter from J. Hopkins (NRC) to O. Kingsley (Exelon), Clinton Power Station, Unit 1 -

Issuance of Amendment 145 Regarding the Replacement of Individual Main Steamline Leakage Limits With An Aggregate Leakage Limit, Revising Technical Specification Surveillance Requirement 3.6.1.3.9, dated March 26, 2002 Page 50 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE

20. EPRI Report 1018243, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325, dated October 2008

21. Letter from K. Jury (AmerGen) to NRC, Clinton Power Station, Unit 1 - Request for Amendment to Technical Specification 3.6.5.1, "Drywell" and 5.5.13, "Primary Containment Leakage rate Testing Program," dated November 16, 2006

22. Letter from P. Simpson (AmerGen) to NRC, Clinton Power Station, Unit 1 - Withdrawal of Request for Amendment to Technical Specification 3.6.5.1, "Drywell" and 5 .5.13, "Primary Containment Leakage Rate Testing Program," dated April 30, 2007

23. NEI 00-02, "Probabilistic Risk Assessment Peer Review Process Guidance," Rev. A3, dated March 2000

24. Letter to D. Heacock from S. Williams (NRC), Surry Power Station, Unit 1 - Issuance of Amendment Regarding the Containment Type A and Type C Leak Rate Tests, dated July 3, 2014 (ML14148A235)

25. Letter to L. Weber from A. Dietrich (NRC), Donald C. Cook Nuclear Plant, Unit 1 -

Issuance of Amendments Re: Containment Leakage Rate Testing Program, dated March 30, 2015 (ML15072A264)

26. Letter to E. Larson from T. Lamb (NRC), Beaver Valley Power Station, Unit Nos. 1 and 2

- Issuance of Amendment Re: License Amendment Request to Extend Containment Leakage Rate Test Frequency, dated April 8, 2015 (ML15078A058)

27. Letter to G. Gellrich from A. Chereskin (NRC), Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Issuance of Amendments Re: Extension of Containment Leakage Rate Testing Frequency, dated July 16, 2015 (ML15154A661)

28. Letter to B. Hanson from R. Ennis (NRC), Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Amendments Re: Extension of Type A and Type C Leak Rate Test Frequencies (TAC NOS. MF5172 AND MF5173), dated September 8, 2015 (ML15196A559)

29. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-S-2002, New York, New York, April 2002

30. ASME/American Nuclear Society, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009, March 2009

31. Clinton Power Station 2009 PRA Peer Review Report, April 2010.

32. 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, dated March 27, 2002 (Accession Number ML020920100)

Page 51 of 52

Attachment 1 EVALUATION OF PROPOSED CHANGE

33. U.S. Nuclear Regulatory Commission (NRC), Memorandum to Brian W. Sheron (Director Office of Nuclear Regulatory Research), From Patrick Hiland, (Chairman, Safety/Risk Assessment Panel for Generic Issue 199), "Safety/Risk Assessment Results for Generic Issue 199, 'Implications of Updated Probabilistic Seismic Hazard Estimates In Central and Eastern United States on Existing Plants'," dated September 2, 2010 (Accession Number ML11356A034)

34. CPS 2011 Self-Assessment of the PRA Against the ASME PRA Standard Requirements Notebook, CPS-PSA-016, Revision 1, December 2011

35. CPS Station Internal Events PRA, Model of Record 2006C

36. Letter from N. DiFrancesco (NRC) to M. Pacilio (Exelon), Clinton Power Station, Unit 1 -

Issuance of Amendment 192 Re: Technical Specification Change for The Relocation of Specific Surveillance Frequency Requirements Based On Technical Specification Task Force (TSTF)-425 (TAC NO. ME3332), dated February 15, 2011

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

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

39. BWROG PSA Peer Review Certification Implementation Guidelines, January 1997.

40. Letter from B. Singal (NRC) to R. Flores (Luminant), Comanche Peak Nuclear Power Plant, Units 1 and 2 - Issuance of Amendments Re: Technical Specification Change For Extension of the Integrated Leak Rate Test Frequency From 10 to 15 Years (CAC Nos.

MF5621 and MF5622), dated December 30, 2015 Page 52 of 52

ATTACHMENT 2 Mark-up of Technical Specifications Page TS 5.5.13 Page 5.0-16a

Programs and Manuals 5.5 5.5 Programs and Manuals (continued)

S.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 so, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, *Performance-Based Containment Leak-Test Program,* dated September 1995,as .

modified by the *following exceptions: (1) Bechtel Topical Report BN-TOP-1 is also an acceptable option for performance of Type A tests, and (2) NEI 94 1995, Section 9.2.3: The first Type A test performed after November 23, 1993 shall be performed no.

later than November 23, 2008.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 9.0 psig.

• The maximum allowable primary containment leakage rate L.,, at P8 ,

sha~l be 0.65\ of primary containment air weight*per day.

Leakage Rate acceptance c~iteria are:

• a. Primary containment leakage rate acceptance criterion is S 1.0 L.. During the first unit startup following testing in accord~ce with this program, the leak rate acceptance criteria are S 0.60 La for the Type B and Type C tests and S o.75 L. for Type A tests;

b. Air lock testing acceptance criteria are:

1) overall air lock leakage rate is s s scfh when teste~ at

~ Par

2) For each door, leakage rate is S 5 scfh when the gap.l between door seals is pressurized ~ P**

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.

NEI 94-01, "Industry Guideline for Implementing Performance-Based (continued)

Option of 10 CFR Part 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, I I

~

CLINTON S.0-16a .Amendment No. 16 0

ATTACHMENT 3 Mark-up of Technical Specification Bases Page (Provided for Information Only)

TS Bases 3.6.1.1 Page B 3.6-5

Primary Containment B 3.6.1.1 BASES SURVEILLANCE SR 3.6.1.1.1 (continued)

REQUIREMENTS (continued) With regard to leakage rate values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument indication uncertainties (Ref. 7).

SR 3.6.1.1.2 With respect to primary containment integrated leakage rate testing, the primary containment hydrogen recombiners (located outside the primary containment) are considered extensions of the primary containment boundary. This requires the smaller of the leakage from the PCIVs that isolate the primary containment hydrogen recombiner, or from the piping boundary outside containment, to be included in the ILRT results. The Frequency is required by the Primary Containment Leakage Rate Testing Program.

With regard to leakage rate values obtained pursuant to this SR, as read from plant indication instrumentation, the specified limit is considered to be a nominal value and therefore does not require compensation for instrument indication uncertainties (Ref. 7).

REFERENCES 1. USAR, Section 6.2.

2. USAR, Section 15.6.5.

3. 10 CFR 50, Appendix J, Option B.

4. USAR, Section 6.2.1. 3-A

5. NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J." 2002

6. ANSI/ANS-56.8-1994, "American National Standard for Containment System Leakage Testing Requirement."

7. Calculation IP-0-0056.

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

CLINTON B 3.6-5 Revision No. 4-6

ATTACHMENT 4 Risk Assessment for CPS Regarding the ILRT (Type A)

Permanent Extension Request

Risk Management Team Clinton Power Station (CPS)

Risk Assessment for CPS Regarding the ILRT (Type A) and DWBT Permanent Extension Request CL-LAR-07 Revision 0

Risk Management Team Risk Management Team Revisions:

REV. DESCRIPTION PREPARER/DATE REVIEWER/DATE APPROVER/DATE C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 METHEDOLOGY............................................................................................... 2-1 3.0 GROUND RULES.............................................................................................. 3-1 4.0 INPUTS ............................................................................................................. 4-1 4.1 General Resources Available ................................................................. 4-1 4.1.1 GGNS DWBT Method ........................................................... 4-11 4.1.2 RBS DWBT Method .............................................................. 4-12 4.1.3 CPS DWBT Method .............................................................. 4-12 4.2 Plant Specific Inputs ............................................................................. 4-19 4.3 CPS Population Dose Derivation .......................................................... 4-25 4.4 Impact of Extension on Detection of Component Failures That Lead to Leakage (Small and Large)................................................................... 4-43 4.5 Impact of Extension on Detection of Steel Corrosion that Leads to Leakage ................................................................................................ 4-46 4.6 Impact of DWBT Interval Extension of Release Categories .................. 4-53 4.6.1 DWBT Data Analysis ............................................................ 4-54 5.0 RESULTS .......................................................................................................... 5-1 5.1 Step 1 - Quantify the Base-Line Risk in Terms of Frequency per Reactor Year ........................................................................................................ 5-3 5.2 Step 2 - Develop Plant-Specific Person-REM Dose (Population Dose) per Reactor Year .................................................................................... 5-8 5.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval From 10-to-15 Years ...................................................................................... 5-11 5.4 Step 4 - Determine the Change in Risk in Terms of Large Early Release Frequency ............................................................................................. 5-14 5.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability ............................................................................................. 5-14 5.6 Summary of Internal Events Results ..................................................... 5-15 5.7 ContributionS from Other Hazard Groups ............................................. 5-17 i C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE OF CONTENTS (cont'd)

Section Page 6.0 SENSITIVITIES ................................................................................................. 6-1 6.1 Sensitivity to Corrosion Impact Assumptions .......................................... 6-1 6.2 EPRI Expert Elicitation Sensitivity ........................................................... 6-3 6.3 DWBT Data Sensitivity............................................................................ 6-6

7.0 CONCLUSION

S ................................................................................................ 7-1

8.0 REFERENCES

.................................................................................................. 8-1 APPENDIX A PRA TECHNICAL ADEQUACY ............................................................A-1 ii C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 permanent extension of the Clinton Power Station (CPS) containment Type A integrated leak rate test (ILRT) interval from ten years to fifteen years. 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 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 [3]. Additionally, consistent with other previous ILRT extension requests for BWR Mark III containments, the risk assessment also includes an assessment for extending the Drywell Bypass Test (DWBT) interval from ten years to fifteen years. 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.

1.2 BACKGROUND

Revisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing requirements from three-in-ten years to at least once per ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage was less than the normal containment leakage of 1.0La (allowable leakage).

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The basis for a 10-year test interval is provided in Section 11.0 of NEI 94-01 [1],

Revision 0, and was established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493 [6],

Performance-Based Containment Leak Test Program, provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project Report TR-104285 [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 percent per day to 5 percent per day leads to a barely perceptible increase in total population exposure, and increasing the leak rate to 50 percent per day increases the total population exposure by less than 1 percent. 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 NRC Regulatory Guide 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/DWBT interval extension risk assessment for CPS employs the EPRI 1018243 1-2 C0467100013-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval methodology, with the affected System, Structure, or Component (SSC) being the primary containment boundary.

1.3 ACCEPTANCE CRITERIA The acceptance guidelines in RG 1.174 [4] are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in core damage frequency (CDF) less than 1.0E-06 per reactor year and increases in large early release frequency (LERF) less than 1.0E-07 per reactor year. Note that CDF is not impacted by the proposed change for CPS. Therefore, since the Type A test does not impact CDF for CPS, 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval (when compared against the baseline interval of 3 tests per 10 years), 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 permanently extending the ILRT/DWBT surveillance interval to 15 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 and the DWBT are not expected to 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. 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 percent 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 would require that the increase in CCFP be less than or equal to 1.5 percentage points.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 2.0 METHEDOLOGY A simplified bounding analysis approach consistent with the EPRI methodology [3] is used for evaluating the permanent change in risk associated with increasing the test interval to fifteen 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/DWBT interval to fifteen 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 Conditional Containment Failure Probability (CCFP)

6. Evaluate the sensitivity of the results to assumptions in the steel corrosion analysis and to variations in the fractional contributions of large isolation failures (due to corrosion) to LERF.

Furthermore,

• 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 containment failure probability (CCFP) for defense-in-depth considerations, and change in LERF to demonstrate that the acceptance guidelines from RG 1.174 are met.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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].

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

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

• The CPS Level 1 and Level 2 internal events PRA models provide representative core damage frequency and release category frequency distributions to be utilized in this analysis.

• It is appropriate to use the CPS internal events PRA model as a gauge to effectively describe the risk change attributable to the ILRT/DWBT extension. It is reasonable to assume that the impact from the ILRT/DWBT extension (with respect to percent increases in population dose) will not substantially differ if other hazard groups were to be included in the calculations; however, other hazard groups (e.g., internal fires, seismic) have been accounted for in the analysis based on the available information for CPS [8, 9] as described in Section 5.7.

• Dose results for the containment failures modeled in the 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.

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

[31] and the Illinois Department of Public Health (IDPH) [30] are appropriate for this analysis.

• The representative containment leakage for Class 1 sequences is 1 La.

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

• The representative containment leakage for Class 3a is 10 La 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 utilized 35La for the Class 3b sequences.

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

• 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.

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

• An evaluation of the risk impact of the ILRT on shutdown risk is addressed using the generic results from EPRI TR-105189 [12].

• The methodology to evaluate the impact of concurrently extending the DWBT interval is performed consistent with previous one-time ILRT/DWBT extensions for BWR Mark III containment types [19, 20, 21],

including CPS, which have been approved by the NRC.

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

• Section 4.1 General Resources Available as input

• Section 4.2 Plant Specific Resources Required

• Section 4.3 CPS Population Dose Derivation

• Section 4.4 Details on the EPRI Methodology that is followed

• Section 4.5 Details of the Calvert Cliffs corrosion analysis method that is also used as a sensitivity for this assessment

• 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.

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

• NUREG/CR-3539 [13]

• NUREG/CR-4220 [14]

• NUREG-1273 [15]

• NUREG/CR-4330 [16]

• EPRI TR-105189 [12]

• NUREG-1493 [6]

• EPRI TR-104285 [2]

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

• Calvert Cliffs corrosion analysis [5]

• EPRI 1018243 [3]

• NRC Final Safety Evaluation Report [7]

• Prior Mark III ILRT/DWBT Extension Risk Assessments [19, 20, 21]

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval NUREG/CR-3539 [13]

Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leak rates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 [22] as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

The study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and to be included in the model.

NUREG/CR-4220 [14]

NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985. The study reviewed over two thousand LERs, ILRT reports and other related records to calculate the unavailability of containment due to leakage. It assessed the large containment leak probability to be in the range of 1E-3 to 1E-2, with 5E-3 identified as the point estimate based on 4 events in 740 reactor years and conservatively assuming a one-year duration for each event.

The study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident.

NUREG-1273 [15]

A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREG/CR-4220 database. This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected. In addition, this study noted that local leak rate tests can detect essentially all potential degradations of the containment isolation system.

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval NUREG/CR-4330 [16]

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

NUREG/CR-4330 is a study that examined the risk impacts associated with increasing the allowable containment leakage rates. The details of this report have no direct impact on the modeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals. However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies:

the effect of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of containment.

EPRI TR-105189 [12]

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

The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessment because this EPRI study provides insight regarding the impact of containment testing on shutdown risk. This study performed a quantitative evaluation (using the EPRI ORAM software) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk.

The result of the study concluded that a small but measurable safety benefit (i.e.,

shutdown CDF reduced by 1.0E-8/yr to 1.0E-7/yr) is realized from extending the test intervals from 3 per 10 years to 1 per 10 years.

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

NUREG-1493 is the NRCs cost-benefit analysis for proposed alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The NRC conclusions are consistent with other similar containment leakage risk studies:

• Reduction in ILRT frequency from 3 per 10 years to 1 per 20 years results in an imperceptible increase in risk.

• Given the insensitivity of risk to the containment leak rate and the small fraction of leak paths detected solely by Type A testing, increasing the interval between integrated leak rate tests is possible with minimal impact on public risk.

EPRI TR-104285 [2]

Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study), the EPRI TR-104285 study is a quantitative evaluation of the impact of extending Integrated Leak Rate Test (ILRT) and (Local Leak Rate Test) LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with NUREG-1150 [17] Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 [6] in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-104285 used a simplified Containment Event Tree to subdivide representative core damage sequences into eight categories of containment response to a core damage accident:

1. Containment intact and isolated

2. Containment isolation failures due to support system or active failures

3. Type A (ILRT) related containment isolation failures

4. Type B (LLRT) related containment isolation failures

5. Type C (LLRT) related containment isolation failures

6. Other penetration related containment isolation failures

7. Containment failure due to core damage accident phenomena

8. Containment bypass 4-4 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Consistent with the other containment leakage risk assessment studies, this study concluded:

These study results show that the proposed CLRT [containment leak rate tests] frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms...

Release Category Definitions The EPRI methodology [2,3] defines accident classes that may be used in the ILRT/DWBT extension evaluation. These containment failure classifications, reproduced in Table 4.1-1, are used in this analysis to determine the risk impact of extending the Containment Type A ILRT and DWBT intervals as described in Section 5 of this report.

NUREG-1150 [23] and NUREG/CR 4551 [17]

NUREG-1150 and the technical basis, NUREG/CR-4551, provide an ex-plant consequence analysis for a spectrum of accidents including a severe accident with the containment remaining intact (i.e., Tech Spec leakage). This ex-plant consequence analysis is calculated for the 50-mile radial area surrounding Grand Gulf, another plant with a Mark III containment. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the CPS Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent CPS. (The meteorology and site differences other than population are assumed not to play a significant role in this evaluation.)

Calvert Cliffs Steel Corrosion Analysis [5]

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting steel liner corrosion, and the corresponding change in risk. The methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms was factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner. CPS also has a concrete cylinder and dome and a concrete basemat, each with a steel liner. The drywell is a Class 1 seismic structure and features reinforced concrete walls and floor in a vertical right cylinder geometry. The ceiling is also reinforced concrete with a removable steel dome known as the drywell head. The floor is common with the primary containment basemat. Additional details are contained in Table 4.1-2. The function of the drywell is to maintain a pressure boundary that forces steam from a loss of coolant accident (LOCA) through the 102 horizontal vents in the drywell wall into the suppression pool.

The corrosion analysis for Calvert Cliffs is used for the CPS containment vessel with slight variations made to account for the design differences.

EPRI 1018243 [3]

EPRI 1018243 presents a risk impact assessment for extending integrated leak rate test (ILRT) surveillance intervals to 15 years and provides the results of an expert elicitation process to determine the relationship between pre-existing containment leakage probability and magnitude.

EPRI 1018243 complements the previous EPRI report 104285 [2]. The earlier report considered changes to local leak rate testing intervals as well as changes to ILRT testing intervals. The original risk impact assessment [2] considers the change in risk based on population dose, whereas the revision [3] considers dose as well as large early release frequency (LERF) and conditional containment failure probability (CCFP).

This report deals with changes to ILRT testing intervals and is intended to provide bases for supporting changes to industry and regulatory guidance on ILRT surveillance intervals.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The risk impact assessment using the Jeffreys Non-Informative Prior statistical method is further supplemented with a sensitivity case using expert elicitation performed to address conservatisms. The expert elicitation is used to determine the relationship between pre-existing containment leakage probability and magnitude. The results of the expert elicitation process from this report are used as a separate sensitivity investigation for the CPS analysis presented here in Section 6.2.

NRC Safety Evaluation Report [7]

This SER documents the NRC staffs evaluation and acceptance of NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, subject to the limitations and conditions identified in the SER and summarized in Section 4.0 of the SER. These limitations (associated with the ILRT Type A tests) were addressed in the Revision 2-A of NEI 94-01 which are also included in Revision 3-A of NEI 94-01 [1] and the final version of the updated EPRI report [3], which was used for this application. Additionally, the SER clearly defined the acceptance criteria to be used in future Type A ILRT extension risk assessments as delineated previously in the end of Section 1.3.

Previous ILRT/DWBT Extension Risk Assessments for Mark III Plants [19, 20, 21]

Reference is made to other extension requests for Mark III containments that considered extensions to the ILRT interval and the DWBT interval.

Consistent with other previous ILRT extension requests for BWR Mark III containments, the risk assessment also includes an assessment for extending the Drywell Bypass Test (DWBT) interval from ten years to fifteen years. 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 4-7 C467150095-12660-12/1/15

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

The DWBT verifies that pre-existing drywell bypass leakage does not exceed the maximum allowed leakage. For CPS, the DWBT acceptance criterion in the Tech Spec SR 3.6.5.1.3 [34] is <10% of the analyzed design limit. The design bypass limit is used to establish the timing of automatic initiation of the containment spray system following a LOCA. If a leakage path were to exist between the drywell and the containment, the leaking steam would produce pressurization of the containment. To mitigate the consequences of any steam which bypasses the suppression pool, a high containment pressure signal will automatically initiate the containment spray system any time after LOCA + 10 minutes. The allowable bypass leakage is defined as the amount of steam which could bypass the suppression pool without exceeding the design containment pressure. The acceptable leakage rates are 10% of the maximum allowable leakage rate.

The Design Basis and Test Leakage criteria are found in Section 6.2 of the USAR [18].

The DWBT thus affects the likelihood of suppression pool bypass in the Level 1 and Level 2 PRA analyses.

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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval It is assumed in this augmented 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 containment has a different arrangement compared to BWR Mark I/II containments or PWR containments. The difference is that the 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 further discussed below, this leads to at least nine combinations of drywell and containment leakage sizes (refer to Figure 4.1-1). Each combination will have an impact on radionuclide releases that corresponds approximately to one of the original containment failure categories.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 4-10 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The different combinations of the drywell and containment leakage sizes can result in different accident classes. To address this issue, the Grand Gulf Nuclear Station (GGNS) and the Clinton Power Station (CPS) DWBT methods have some slight differences, which are discussed in more detail in the following two sub-sections.

Following that, the approach used for the updated CPS DWBT method is presented.

The Mark III and CPS plant-specific data utilized for the DWBT portion of this risk assessment is then provided in Section 4.6.

4.1.1 GGNS DWBT Method In the GGNS assessment [20], the assignment of each of these combinations to an original containment failure category depends on the consideration of the availability of the containment spray system. If containment sprays are available, the combination of drywell and containment leakage is categorized based on the containment leakage category. If containment sprays are not available, the combination of drywell and containment leakage is assumed to result in containment failure (Class 7) except for the combinations with base drywell bypass leakage. The combinations with base drywell leakage (DWLb) are assumed to have the same categories as the base case ILRT evaluation. Table 4.1-3 summarizes the classification of combinations into the EPRI accident classes used in the GGNS assessment.

The probability for each combination in Table 4.1-3 is determined by multiplying the conditional probabilities for DWBT and ILRT category by each other. For those cases where containment spray is a factor the probability of the combination of DWBT and ILRT is multiplied by the probability that containment spray is available or is not available as applicable.

The other change in the methodology to address the DWBT is the need to increase the containment failure due to phenomenology class (Class 7) frequency for the extended test frequencies. This is done in a manner similar to the method applied to Class 3a and 3b. That is, the Class 1 frequency is also adjusted downward for the Class 7 frequency increase in order to maintain the same total CDF.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 4.1.2 RBS DWBT Method For the most part, the GGNS methodology for DWBT extension evaluation was previously used for RBS [21]. The main modifications to the GGNS methodology were as follows:

• RBS credited the containment unit coolers to mitigate the adverse effects of the increased drywell leakages instead of the containment spray credited in the GGNS evaluation. Containment spray has dual functions by reducing the containment pressure and scrubbing the fission products from the containment atmosphere while containment unit coolers were designed mainly to reduce containment pressure. However, the GGNS method conservatively does not credit the containment spray for scrubbing. Thus the effects of crediting containment unit coolers and containment spray are the same.

• The RBS base cases for DWBT extension evaluation used EPRI Class 1 frequency to calculate the Class 3a, Class 3b and additional Class 7 frequencies. The GGNS method base cases used the total CDF for the calculation, which was conservative since more Class 1 frequency would be re-categorized into Class 3a, 3b or Class 7 frequencies. Such a conservative approach was not appropriate for the RBS evaluation since the RBS Class 1 frequency only consisted of about 10% of the total CDF, and as such, the calculated Class 3a, 3b and additional Class 7 frequencies could exceed the Class 1 frequency if using the total CDF for calculations. Therefore, only the CDF portion that does not lead to a more severe release category in the Level 2 analysis is re-categorized to Class 3a, 3b, or 7. This exclusion of a portion of the CDF that is impacted by the DWBT extension is similar to the allowed exclusion of LERF contributors per the accepted EPRI methodology for ILRT extension assessments.

4.1.3 CPS DWBT Method The CPS DWBT methodology [19] is very similar to that used for GGNS except the assignment of the nine combinations of drywell and containment leakage sizes to an original containment failure category. Unlike the GGNS approach which conservatively assumed all combinations without containment spray available would contribute to accident class 7, CPS used a Clinton specific MAAP 4.0 model to determine the effects of the increased drywell leakages that could lead to higher containment pressure. The MAAP run results along with other considerations found that containment failure induced by containment pressurization aggravated by the drywell bypass leakage 4-12 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 is considered a negligible contributor to CDF.

CPS assigned the equivalent EPRI category and LERF characterization as shown in Table 4.1-4. Note, these assignments are consistent with GGNS and RBS assignments with the exception that leakage combinations leading to EPRI Class 7 assignments are not included and CPS combination CA1 is conservatively classified as 3a EPRI release, consistent with the previous CPS ILRT extension analysis [19] while GGNS assigned this combination to EPRI class 1.

No credit for the availability of containment spray was taken in the CPS analyses.

Similar to GGNS method, 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. As previously noted GGNS and RBS conservatively assumed all combinations without containment spray (GGNS) or Unit Coolers (RBS) available would contribute to accident class 7. The assumed normal leakage rate DWLb is sufficiently low, that a 100DWLb drywell release would be a negligible contributor to accident class

7. See section 4.6 for additional detail.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-1 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 La, 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-2 CPS CONTAINMENT AND DRYWELL STRUCTURAL DESIGN FEATURES DESIGN FEATURE CPS CONTAINMENT DESIGN CPS DRYWELL DESIGN Design Pressure 15 psig 30 psig Construction Reinforced concrete with carbon steel liner. Reinforced concrete with Stainless steel in suppression pool area. steel plate covering internal surfaces of drywell cylinder and top slab.

Stainless steel in suppression pool area.

Drywell vents below suppression pool surface connect drywell and containment.

Liner Thickness Nominally 1/4 to 1/2 Nominally 1/2 Concrete Wall 3-0 walls, 2-6 dome, 9-8 base mat 5-0 walls, 6-0 top slab, Thickness 9-8 base mat Equipment Hatch Dished welded steel hatch bolted onto flange on Same as containment.

hatch barrel. Flange connection is double gasketed.

Personnel Airlocks 2 Personnel airlocks each with interlocked 1 Personnel airlock (same double doors. Constructed of welded steel. as containment).

Doors have double gasketed flanges.

Piping Penetrations Three types: Same as containment.

Type 1 for high-energy lines, guard pipes enclose these lines to direct the energy into the 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.

Other Mechanical Inclined Fuel Transfer Tube consists of a 3/4 thick Drywell head (24.7 feet carbon steel rolled plate pipe sleeve of 40 inside diameter elliptical dome diameter with a 36 standard flange on the with double seal) is bolted containment side. onto drywell head flange.

Serves as boundary between drywell and upper containment pools during non-refueling periods.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-2 CPS CONTAINMENT AND DRYWELL STRUCTURAL DESIGN FEATURES DESIGN FEATURE CPS CONTAINMENT DESIGN CPS DRYWELL DESIGN Electrical Penetrations Dual header plate type electrical penetrations are Cables penetrating the used. drywell wall pass through penetrations that are filled with sealant.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-3 GGNS [20] DWBT AND ILRT LEAKAGE COMBINATION ACCIDENT CLASSES EPRI LEAKAGE DW BYPASS CONTAINMENT CLASSIFICATION COMBINATIONS LEAKAGE LEAKAGE ASSIGNMENT AA 1 DWLb 1 La 1 AB 1 DWLb 10 La 3a (2)

AC 1 DWLb 35 La 3b BA1 CS Available 10 DWLb 1 La 1 (1)

BA2 CS Not Available CF CF 7 BB1 CS Available 10 DWLb 10 La 3a BB2 CS Not Available CF CF 7 BC1 CS Available 10 DWLb 35 La 3b BC2 CS Not Available CF CF 7 (2)

CA1 CS Available 35 DWLb 1 La 1 CA2 CS Not Available CF CF 7 CB1 CS Available 35 DWLb 10 La 3a CB2 CS Not Available CF CF 7 CC1 CS Available 35 DWLb 35 La 3b CC2 CS Not Available CF CF 7 Notes to Table 4.1-3:

(1)

CF = Containment failure assumed to occur.

(2)

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.1-4 CPS [19] DWBT AND ILRT LEAKAGE COMBINATION ACCIDENT CLASSES EPRI LEAKAGE DW BYPASS CONTAINMENT CLASSIFICATION COMBINATIONS LEAKAGE LEAKAGE ASSIGNMENT AA 1 DWLb 1 La 1 (Non-LERF)

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

(1)

AC 1 DWLb 35 La 3b (LERF)

BA1 10 DWLb 1 La 1 (Non-LERF)

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

BC1 10 DWLb 35 La 3b (LERF)

(1)

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

CB1 35 DWLb 10 La 3b (LERF)

CC1 35 DWLb 35 La 3b (LERF)

Note to Table 4.1-4:

(1)

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

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

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

• Population within a 50-mile radius

• Reactor Power Level [18]

• Allowable Containment Leakage [18]

CPS Internal Events Core Damage Frequency The current CPS Internal Events 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 Report for the 2014 PRA Interim Update [24], the mean value of the internal events core damage frequency (CDF) is 2.13E-06/yr. (truncation limit 5E-13/yr.). Core Damage Frequency by Class is provided in Table 4.2-1.

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 4.2-2 summarizes the pertinent CPS results in terms of release category, taken from Table 3.4-4 of the CPS PRA Summary Notebook [24]. The total Large Early Release Frequency (LERF) which corresponds to the H/E release category in Table 4.2-2 was calculated to be 1.16E-7/yr (truncation limit 5E-14). The total release frequency is 1.30E-06/yr., with a total CDF of 2.23E-06/yr (using a 5E-14 truncation limit.). This corresponds to an OK release (i.e.,

intact containment limited to normal leakage) of 9.30E-7/yr.

Based on the Level 1 and LERF PRA model results described above, Table 4.2-3 lists the relevant EPRI release category frequencies pertinent for the ILRT/DWBT extension risk assessment, including the delineation of LERF and non-LERF frequencies for Class 7.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The non-LERF frequencies for Class 7 are determined by reviewing the top 114 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. The phenomena-induced containment failures (non-LERF) sequences contribute 9.16E-07/yr. as shown in the Table 4.2-3.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.2-1 2014 CLINTON LEVEL 1 CDF RESULTS(3)

(1)

CLASS DESCRIPTION CDF (/YR)  % OF CDF IA/IC Loss of Makeup at 4.9E-07 22.9%

High RPV Pressure (Transient Initiators)

IBE Early Station 1.4E-07 6.6%

Blackout (less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)

IBL Late Station 6.5E-07 30.4%

Blackout (greater than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)

ID Loss of Makeup at 1.0E-07 4.7%

Low RPV Pressure (Transient Initiators)

II Loss of Containment 4.2E-07 19.6%

Heat Removal IIIA Excessive LOCA 1.0E-09 <0.1%

IIIB Loss of Makeup at 3.7E-08 1.7%

High RPV Pressure (LOCA Initiators)

IIIC Loss of Makeup at 8.5E-08 4.0%

Low RPV Pressure (LOCA Initiators)

IIID Loss of Vapor 8.3E-09 0.4%

Suppression (LOCA Initiators)

(2)

IV Loss of Adequate 2.0E-07 9.6%

Reactivity Control (ATWS)

V Containment Bypass 1.6E-09 0.1%

Total - 2.1E-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)

Class IVL included in Class IV.

(3)

From Table 7.3-1 of The CPS PRA Level 2 Notebook [32].

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.2-2

SUMMARY

OF CONTAINMENT EVALUATION(3)

INPUT OUTPUT LEVEL 1 PSA CET EVALUATION RELEASE CORE DAMAGE CHARACTERIZE FREQUENCY FREQUENCY RELEASE RELEASE BIN (PER YEAR)

(1) 2.23E-6/year Little or No Release OK 9.30E-7

@ 5E-14 truncation (Intact)

(2)

LL & Late 2.00E-8 LL & I Low Public LL & E 1.16E-09 Risk Impact (2)

L & Late 9.88E-8 L&I L&E 2.17E-7 (2)

Moderate M & Late 2.02E-7 Release M&I M&E 2.00E-7 (2)

H & Late 4.45E-7 High Release H&I H&E 1.16E-7 4-22 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Notes to Table 4.2-2:

(1)

Level 1 CDF result when using a quantification truncation limit of 5E-14/yr. This result is only provided when comparing to the Level 2 results.

(2)

Sum of release frequencies in right column (not including Release Bin OK) is 1.30E-06/yr.

(3)

From Table 7.2-1 of The CPS PRA Level 2 Notebook [32].

(4)

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 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />sy, 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 />.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.2-3 RELEVANT LEVEL 2 RELEASE CATEGORY FREQUENCIES FOR CPS EPRI RELEASE CATEGORY FREQUENCY/YR SOURCE 1: No Containment Failure 9.30E-07 Table 4.2-2 CDF contribution of sequences (top 99%) that (1) included failure of the 2: Containment Isolation Failure 2.68E-07 containment isolation function (Event Tree Node IS=F)

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

(LERF) 1.14E-07 Containment Bypass (Class V)

Table 4.2-2 and sequence evaluation 2.23E-06 CDF -

7: Phenomena-induced containment failures - 9.30E-07 (non-LERF) OK (EPRI Class 1) -

- 1.16E-07 LERF (EPRI

- 2.68E-07 Class 7 LERF & 8) -

= 9.16E-07 Cont. Isolation (EPRI Class 2)

Class V 8: Containment Bypass 1.55E-09 (ISLOCA + BOC Sequences)

Total: 2.23E-06 Note to Table 4.2-3:

(1)

Not all Containment Isolation Failure sequences are found to be H/E (LERF) in the CPS PRA based on MAAP calculations. Never-the-less, 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 affects just the total dose and % change metrics.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 utilized. From the EPRI guidance it is noted that 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]. 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.

• 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.

• 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/DWBT interval extension submittal [19].

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

COLLAPSED APB NUMBER DESCRIPTION 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 4-27 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.3-2 GRAND GULF NUREG/CR-4551 [17] 50-MILE RADIUS POPULATION DOSE APB FRACTIONAL APB 50-MILE APB CONTRIBUTION RADIUS DOSE APB 50-MILE FREQUENCY TO 50-MILE RISK RADIUS DOSE APB APB (PER RADIUS TOTAL (PERSON- (PERSON-(1) (2) (3) (4) (5)

1. DEFINITION YEAR) DOSE RISK REM/YEAR) REM)

CD, VB, Early 6.46E-7 .268 0.139 2.15E+5 1 CF, Early SP Bypass, CS Not Available CD, VB, Early 2.00E-7 .056 0.029 1.45E+5 2 CF, Early SP Bypass, CS Available CD, VB, Early 2.86E-8 .011 5.7E-3 1.99E+5 3 CF, Late SP Bypass CD, VB, Early 8.92E-7 .267 0.139 1.56E+5 4 CF, 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 4-29 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The APBs described above can then be 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/DWBT frequency and are therefore (per the EPRI guidance) not included in the assignment process.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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) 4-31 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 substantial differences in reactor power level, population density, 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 utilizes census data to calculate population counts for user defined sector segments. The EPRI methodology does not specifically specify population projection to a future date. For this risk assessment the CPS population was projected to the year 2030 to represent the average population density over the next two 15 year extension intervals.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.3-5 SECPOP CODE POPULATION ESTIMATES SECPOP (1)(2) (1)(2)

RADIUS 2000 2010 2020 2030 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 (1)

The 2020 and 2030 estimates are made assuming the 5.79%

population increase experienced in the 50-mile radius region during the decade from July 2000 to July 2010 continues to occur each of the next two decades.

(2)

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 4.3-5b 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). See map and table below.

Illinois Counties Within 50 Miles of CPS Shown in Map Above 4-33 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.3-5B IDPH [30] POPULATION PROJECTION ESTIMATES AREA WITHIN 2020 2025 50 MILE 2010 POPULATION POPULATION COUNTY RADIUS POPULATION ESTIMATE ESTIMATE Champaign 99% 201,370 217,735 225,626 Christian 40% 34,804 33,152 32,345 Coles 20% 53,945 56,851 58,405 De Witt 100% 16,583 15,832 15,495 Douglas 75% 19,976 19,767 19,709 Ford 75% 14,074 13,450 13,244 Iroquois <5% 29,657 27,687 26,816 Livingston 40% 38,882 39,390 39,596 Logan 100% 30,272 30,380 30,441 Macon 100% 110,757 105,401 103,126 Mason 25% 14,627 12,841 12,074 McLean 100% 169,838 188,341 197,855 Menard 40% 12,708 12,867 12,913 Moultrie 90% 14,846 14,715 14,706 Piatt 100% 16,722 16,205 16,000 Sangamon 35% 197,822 203,501 207,194 Shelby 25% 22,339 21,496 21,118 Tazewell 75% 135,439 136,051 136,436 Vermillion <5% 81,588 77,965 76,441 Woodford 50% 38,664 40,350 41,360 (1)(2)

Total -- 1,255,013 -- 1,300,900 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%

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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:

• Grand Gulf Containment Volume [29] = 1.40E+6 ft3

• 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:

(0.73%

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 (1)

APB # REM) 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 4.2-3 for each relevant EPRI release category class from Table 4.1-1, and then 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 4-36 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 3 frequencies are derived later and the Class 4, 5, and 6 frequencies are not utilized in the EPRI methodology for the ILRT extension risk assessment).

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 class which differs 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 the top 114 sequences of the Clinton Level 2 model (which provide more than 99.0% of Level 2 release frequency) and to assign each sequence into one of the collapsed Accident Progression Bins (APBs) from NUREG/CR-4551. The CPS Level 2 model (i.e., containment event tree structure) contains a significantly larger amount of information about the accident sequences than what is used 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 utilized (or did not contribute) to the APB assignment performed here for the ILRT assessment. Additionally, it should be noted that 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 Accident Progression Bins from NUREG/CR-4551 as described in Section 4.2. The Class 7 Sequences are divided into two categories.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Class 7 LERF is defined to consist of LERF sequences (excluding the BOC, 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 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 1 CX - No Energetic (highest dose) is assigned.

Containment Failure DL - Drywell If the drywell is not isolated or Suppression Pool Scrubbing fails, it is Isolation assumed that an un-scrubbed release to containment occurs as SP - Suppression soon as the vessel is breached. If Containment Spray (CS) fails the Pool scrubbing 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 to Pool Failure Above 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 assigned to APB 5.

Isolation For Class IIE sequences with late GE declaration, containment SP - Suppression failure is assumed to occur in the early time frame. If the drywell is Pool scrubbing 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.

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

POPULATION DOSE RISK (50 RELEASE POPULATION MILES)

ACCIDENT CLASS FREQUENCY / DOSE (50 MILES) (PERSON-REM /

(APB NUMBER) YR PERSON-REM (1) YR) (2) 7 LERF (APB 1) 1.14E-07 5.48E+05 6.27E-02 7 non-LERF 7a (APB 1) 4.72E-08 5.48E+05 2.59E-02 7b (APB 3) 2.29E-09 5.08E+05 1.16E-03 7c (APB 4) 3.35E-08 3.97E+05 1.33E-02 7d (APB 5) 6.27E-07 3.21E+05 2.01E-01 7e (APB 6) 2.27E-07 3.34E+05 7.57E-02 7f (APB 8) 7.14E-09 1.38E+05 9.88E-04 (4)

Class 7 non-LERF Total 9.44E-07 3.37E+05 3.18E-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.

(3)

Total non-LERF release frequency is shown as 9.16E-07 in Table 4.3-9. Release frequency above used a summing of CAFTA sequences contributions using Fussell-Vesely contributions from a cutset report. Sequence tagging in the recovery led to a small percentage of cutsets being double counted (counted as contributing to two sequence endstates). URE CL2015-014 is tracking resolution. This issue has no impact to CDF/LERF values. Table 4.2-9 calculation used a different methodology to estimate 7 non-LERF contributions. Negligible impact to ILRT/DWBT results.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 9.30E-07 2.71E+03 2.52E-03 2: Containment Isolation Failure 2.68E-07 5.48E+05 1.47E-01 7: Phenomena-induced 1.14E-07 5.48E+05 6.27E-02 containment failures (LERF) 7: Phenomena-induced 9.16E-07 3.37E+05 3.09E-01 containment failures (non-LERF) 8: Containment Bypass 1.55E-09 5.48E+05 8.49E-04 Total: 2.23E-06 5.22E-01 4-42 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 test 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 latest 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 latest available EPRI data [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 delta 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 delta LERF. EPRI describes ways to demonstrate that, using plant-specific calculations, the delta 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 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 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), and the average time that a leak could exist without detection for a ten-year interval is 5 years (10 yr / 2). This interval change would lead to a non-detection probability that is a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRT testing, given a 10-year vs. a 3-yr (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 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval interval. Correspondingly, an extension of the ILRT interval to fifteen years can be estimated to lead to about a factor of 5.0 (7.5/1.5) increase in the non-detection probability of a leak.

CPS Past ILRT Results The surveillance frequency for Type A testing in NEI 94-01 [1] under option B criteria is at least once per ten years based on an acceptable performance history (i.e., two consecutive periodic Type A tests at least 24 months apart) where the calculated performance leakage rate was less than 1.0La, and in compliance with the performance factors in NEI 94-01, Section 11.3. Based on the successful completion of two (1) consecutive ILRTs at CPS , the current ILRT interval is once per ten years. Note that the probability of a pre-existing leakage due to extending the ILRT interval is based on the industry-wide historical results as noted in the EPRI guidance document [3].

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/DWBT interval to fifteen 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].

(1)

Per email from F. Sarantakos, Appendix J Program Engineer, the previous ILRTs were performed in 1993 and 2008 and both ILRTs met Tech Spec requirements (Leakage rates were below TS allowed leakage rate). DWBTs were also performed in 1993 and 2008 and also met TS requirements.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

5. Determine the impact on the Conditional Containment Failure Probability (CCFP).

6. Evaluate the sensitivity of the results to assumptions in the steel corrosion analysis and to variations in the fractional contributions of large isolation failures (due to corrosion) to LERF.

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 conditional containment failure probability (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.

4.5 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 would not be 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 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 4-47 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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:

• Differences between the containment basemat and the containment cylinder and dome

• The historical flaw likelihood due to concealed corrosion

• The impact of aging

• The corrosion leakage dependency on containment pressure

• The likelihood that visual inspections will be effective at detecting a flaw Assumptions

• Consistent with the Calvert Cliffs analysis, a half failure is assumed for the basemat concealed liner corrosion due to lack of identified failures (see Table 4.5-1, Step 1).

• 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 4-48 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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).

• 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). Sensitivity studies are included that address doubling this rate every two years and every ten years.

• 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. Sensitivity studies in Section 6 are included that increase and decrease the probabilities by an order of magnitude (See Table 4.5-1, Step 4).

• Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the basemat region is considered to be less likely than the containment walls. (See Table 4.5-1, Step 4.)

• 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 very 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. Sensitivity studies are included in Section 6 that evaluate total detection failure likelihoods as low as 5%

and as high as 15%, respectively.

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 2/(70

• 5.5) = 5.2E-3 0.5/(70

• 5.5) = 1.3E-3 with 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 avg 5-10 5.2E-3 avg 5-10 1.3E-3 failure rate doubles every five years (14.9% increase per 15 1.4E-2 15 3.5E-3 year). The average for 5th to 10th year is set to the historical 15 year average = 6.27E-3 15 year average = 1.57E-3 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 4.06% (1 to 10 years) 1.02% (1 to 10 years)

Uses age adjusted flaw 9.40% (1 to 15 years) 2.35% (1 to 15 years) likelihood (Step 2), assuming failure rate doubles every five (Note that the Calvert Cliffs (Note that the Calvert Cliffs years (consistent with Calvert analysis presents the delta analysis presents the delta Cliffs analysis - See Table 6 of between 3 and 15 years of between 3 and 15 years of Reference [5]). 8.7% to utilize in the estimation 2.2% to utilize in the of the delta-LERF value. For estimation of the delta-LERF this analysis, the values are value. For this analysis, calculated based on the 3, 10, however, values are and 15 year intervals.) calculated based on the 3, 10, and 15 year intervals.)

4 Likelihood of Breach in 1% 0.1%

Containment Given Steel Flaw The failure probability of the containment cylinder and dome is assumed to be 1%

(compared to 1.1% in the Calvert Cliffs analysis). The basemat failure probability is assumed to be a factor of ten less, 0.1% (compared to 0.11%

in the Calvert Cliffs analysis).

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.5-1 STEEL CORROSION BASE CASE CONTAINMENT CYLINDER STEP DESCRIPTION AND DOME CONTAINMENT BASEMAT 5 Visual Inspection Detection 10% 100%

Failure Likelihood 5% failure to identify visual Cannot be visually inspected.

Utilize assumptions consistent flaws plus 5% likelihood that with 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.

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) 0.00406% (at 10 years) 0.00102% (at 10 years)

=4.06%

• 1%

• 10% =1.02%

• 0.1%

• 100%

0.0094% (at 15 years) 0.00235% (at 15 years)

=9.40%

• 1%

• 10% =2.35%

• 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 10 years: 0.00406% + 0.00102% = 0.00508%

At 15 years: 0.0094% + 0.00235% = 0.0118%

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 4.6 IMPACT OF DWBT INTERVAL EXTENSION OF RELEASE CATEGORIES Similar to the prior CPS ILRT/DWBT interval extension risk assessment, Table 4.6-1 provides the release categories that are utilized in this assessment for the different combinations of drywell bypass leakage and containment leakages. These classifications are consistent with the CPS ILRT/DWBT risk assessment [19] except for using the multiplier of 100 (per the updated EPRI guidance) rather than a multiplier of 35.

TABLE 4.6-1 CPS DWBT AND ILRT LEAKAGE COMBINATION ACCIDENT CLASSES EPRI LEAKAGE DW BYPASS CONTAINMENT CLASSIFICATION COMBINATIONS LEAKAGE LEAKAGE ASSIGNMENT AA 1 DWLb 1 La 1 AB 1 DWLb 10 La 3a AC 1 DWLb 100 La 3b BA1 10 DWLb 1 La 1 BB1 10 DWLb 10 La 3a BC1 10 DWLb 100 La 3b CA1 100 DWLb 1 La 3a CB1 100 DWLb 10 La 3b CC1 100 DWLb 100 La 3b Note to Table 4.6-1:

(1)

CF = Containment failure assumed to occur.

Again, 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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.

Clinton Continuous Monitoring Capability Clinton has the ability to continuously monitor DW leakage. The latest drywell leakage test, performed in 2008, found drywell leakage to be 20.18 psi at 3 psi (2). Leakage during operation is less as differential pressure between the drywell and containment is

< 1 psid. However, small airline leaks cause the drywell to pressurize at a rate of approximately 0.03 psi/hr and operators must vent the drywell approximately once per (1)

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

(2)

Drywell Leakage provided by T. Hable, Clinton SRME to J. Steinmetz via email on 10/07/2015.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval day.(1) The following paragraphs are from the 2003 ILRT 15 year one time LAR submittal [19]. This description below continues to reflect the current operating practices and drywell performance:

Due to the demonstrated leaktight performance of the drywell, CPS is able to monitor the integrity of the drywell during normal plant operation. This is possible due to the normal operation of pneumatic controls and operators in the drywell that pressurize the drywell, plus the existence of small instrument air system leaks. These effects create a differential pressure between the drywell and primary containment that is monitored, and periodic operation action is required to vent the drywell.

For example, in 1994, the drywell was being pressurized at a rate of approximately 0.04 psi/hr. The drywell was being vented approximately once per day when pressure approached the upper TS limit of 1.0 psid.

Based on application of the ideal gas law and known data, such as the drywell pressurization rate and the drywell leakage measured during the fourth refueling outage (RF-4), the total amount of instrument air in-leakage was calculated to be between 21.5 and 22.5 scfm. The rate of drywell pressurization remained essentially constant since drywell closeout from RF-4. Pressurization rates following subsequent refuelings have also remained consistent with those observed following RF-4.

This steady drywell pressurization rate allows qualitative monitoring of the drywell leakage rate. An increase in this rate would be indication of an increase in the instrument air system leakage into the drywell since it is improbable that the drywell would become more leaktight. Conversely, a decrease in this rate would be evidence of a larger drywell leakage area.

The maximum drywell leakage rate that would still maintain a differential pressure between the drywell and wetwell must be less than the instrument air in-leakage rate (which after RF-4 was 23 scfm). The A/k for a 23 scfm leak at 0.2 psid is 0.0025 ft2 or 0.2% of the allowable leakage area.

Because of this large margin to the allowable drywell leakage rate, it has been concluded that as long as the drywell continues to pressurize, regardless of the rate, drywell integrity is always assured. This ability to qualitatively assess the integrity of the drywell during normal plant operation provides further support to extending the DWBT interval.

In order to provide added assurance that the drywell has not seriously degraded between the performances of DWBTs, a qualitative assessment of the drywell leak tightness is performed at least once per operating cycle.

The first assessment was performed prior to Operating Cycle 7. By (1)

The pressurization rate and frequency of venting are derived from graphs found in the 2009 to 2015 periodic drywell leakage assessments documented under Clinton AR 400694 to meet programmatic commitments found in NRC approval of License Amendment 160 [11] and License Request to extend Drywell testing to 10 years [10].

4-55 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval checking for gross leakage, this assessment will provide an indication of the ability of the drywell to perform its design function. As a check for gross leakage, the assessment may not identify drywell leakage that is masked by plant conditions, or identify leakage through systems that are not communicating with the drywell atmosphere at the time of the assessment.

For example, minor increases in drywell bypass leakage could be masked by a small leak in the instrument air system inside the drywell. The assessment is not detailed enough to account for such minor changes.

However, as demonstrated above, as long as the drywell continues to pressurize, regardless of the rate, drywell integrity is always assured.

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

[27] and updated with the latest (Feb-08) Clinton Power Station 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.

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 4-56 C467150095-12660-12/1/15

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

SITE TEST DATE (SCFM) 300 SCFM 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 then 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 6

• 1 DWLb and all test data is below 10 DWLb. 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.

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 4-57 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval These represent very 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).

FIGURE 4.6-1 MARK III DWBT RESULTS COMPARED TO 300 SCFM X Axis = Test # from Table 4.6-2 Y Axis = Test Leakage ÷ 300 SCFM (1)

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

• 10 = 36,540 scfm) 4-58 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Only 2 of the data points are measurably in the 6-10 range, and all are below the 10 Lb 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 one-time 15 year ILRT LAR [19]. 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 NEI/EPRI approach is reflective of drywell failures. The body of plant experience used considered the older BWR containment designs.

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

• 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.

• 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.

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 4.6-3 DW AND WW LEAKAGE PROBABILITIES DW DW WW WW LEAKAGE LEAKAGE PROBABILITY LEAKAGE LEAKAGE SIZE (LB) SIZE (LA) PROBABILITY (BASE)

(1) (1) 1Lb 0.9885 1La 0.9885 10Lb 0.0092 10La 0.0092 100Lb 0.0023 100La 0.0023 (1)

The probability of assumed normal drywell leakage (1La) is

[1-(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:

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Risk Impact Assessment of Extending the CPS ILRT/DWBT 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) 0.99 1 0.0115 3a 0.0023 3b A sensitivity case increasing the probability of a small (10DWLb) and large (100DWLb) leakage rates by a factor of 10 is used in the assessment as described in Section 6.3.

This sensitivity case is not considered representative of Clinton because Clinton has a daily check on the drywell integrity via the observed daily pressurization of the drywell.

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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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.

The assessment documented the following:

• 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.

• The pressure suppression capability of the containment is robust.

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

• 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 very high.

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The pressurization issue was addressed in the Safety Evaluation Report (SER) that was part of the NRC letter [11] approving the onetime 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, 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, the following conclusion is reached for the request for a permanent 15 interval risk assessment:

There is no change in CDF due to the small increases in drywell bypass leakage associated with the DWBT interval extension.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 5.0 RESULTS The application of the approach based on EPRI Guidance [3] has led 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 performed 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:

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

• 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).

• 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 5-1 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Guidance, this class is not specifically examined since it will not significantly influence the results of this analysis.

• 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.

• 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 risk impact of extending Type A test interval from 3 to 15 and 10 to 15 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 Conditional Containment Failure Probability (CCFP).

Step 6 Evaluate the sensitivity of the results to assumptions in the steel corrosion analysis and to variations in the fractional contributions of large isolation failures (due to corrosion) to LERF.

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

5-2 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 5.1 STEP 1 - QUANTIFY THE BASE-LINE RISK IN TERMS OF FREQUENCY PER REACTOR YEAR The CPS PRA Level 2 Model [25] 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.

For the assessment of the impact on the risk profile due to the ILRT/DWBT 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 1.55E-09/yr, as documented in Table 4.2-3.

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.

5-3 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 is 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 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.

5-4 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Class 2 Sequences This group consists of large containment isolation failures. The failure frequency for Class 2 sequences is 2.68E-07/yr, as documented in Table 4.2-3. Note that this frequency is not affected by the ILRT/DWBT 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 for relatively large flaws, consistent with the EPRI methodology [3].

The respective frequencies per year are determined as follows:

PROBClass_3a = probability of small pre-existing containment leakage

= 0.0092 (see Section 4.4)

PROBClass_3b = probability of large pre-existing containment leakage

= 0.0023 (see Section 4.4)

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 * [2.23E (1.14E-07 + 1.60E-09 + 4.20E-07 + 6.50E-07)]

= 1.20E-08/yr 5-5 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Class_3b = 0.0023 * [CDF - (EPRI Class 7 LERF + EPRI Class 8 + Class II +

Class IBL)]

= 0.0023 * [2.23E (1.14E-07 + 1.60E-09 + 4.20E-07 + 6.50E-07)]

= 2.42E-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 latest EPRI methodology [3]. The probability of Class 3a and Class 3b leakages are combined leakage probabilities from Table 4.6-4.

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 is9.16E-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)

= 2.23E (2.68E-07 (class 2) + 1.20E-08 (3a) + 2.42E-09 (3b) + 1.14E-07 (7 LERF) + 9.16E-07 (7-Non-LERF) + 1.55E-09 (Class 8))

= 9.16E-07/yr 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 9.16E-07 2 Large Isolation Failures (Failure to Close) 2.68E-07 3a Small Isolation Failures 1.20E-08 3b Large Isolation Failures 2.42E-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) 1.14E-07 7 non-LERF Failures Induced by Phenomena (non-LERF) 9.16E-07 8 Containment Bypass 1.55E-09 CDF All CET End states (including intact case) 2.23E-06 5-7 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 = 3.37E+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.

5-8 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 (1 La) 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 5.48E+05 (LERF) 7 non-LERF Failures Induced by Phenomena (non-3.37E+05 LERF) 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.

5-9 C467150095-12660-12/1/15

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

EPRI METHODOLOGY CORROSION ACCIDENT PERSON- CHANGE DUE TO CLASSES REM PERSON- PERSON- CORROSION (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 9.16E-07 2.49E-03 9.16E-07 2.49E-03 -2.52E-08 2 Large Isolation Failures 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 1.20E-08 3.25E-04 1.20E-08 3.25E-04 --

3b Large Isolation Failures 2.71E+05 2.42E-09 6.58E-04 2.43E-09 6.60E-04 2.52E-06 7 LERF Failures Induced by 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 --

CDF All CET end states 2.23E-06 0.523 2.23E-06 0.523 2.49E-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.

5-10 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 5.3 STEP 3 - EVALUATE RISK IMPACT OF EXTENDING TYPE A TEST INTERVAL FROM 10-TO-15 YEARS The next step is to evaluate the risk impact of extending the test interval from its current ten-year value to fifteen-years. To do this, an evaluation must first be made of the risk associated with the ten-year interval since the base case applies to a 3-year interval (i.e., a simplified representation of a 3-in-10 year interval).

Risk Impact Due to 10-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). However, as noted previously, the DWBT tests also impact the Class 7 sequences. Thus, the frequency of Class 3a, 3b, and Class 7 sequences are impacted by the ILRT/DWBT interval extension. The risk contribution is changed based on the EPRI guidance as described in Section 4.4 by a factor of 3.33 compared to the base case values. The results of the calculation for a 10-year interval are presented in Table 5.3-1.

Risk Impact Due to 15-Year Test Interval The risk contribution for a 15-year interval is calculated in a manner similar to the 10-year interval. The difference is in the increase in probability of not detecting a leak in Classes 3a and 3b for the ILRT Type A tests, and for Class 7 for the DWBT tests. For this case, the value used in the analysis is a factor of 5.0 compared to the 3-year interval value, as described in Section 4.4. The results for this calculation are presented in Table 5.3-2.

5-11 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 5.3-1 CPS ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR 1 IN 10 YEAR ILRT/DWBT FREQUENCY EPRI METHODOLOGY PLUS CHANGE DUE TO (2)

EPRI METHODOLOGY CORROSION CORROSION OR ACCIDENT PERSON- DWBT 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 8.82E-07 2.39E-03 8.82E-07 2.39E-03 -1.44E-07 2 Large Isolation Failures 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 3.98E-08 1.08E-03 3.98E-08 1.08E-03 --

3b Large Isolation Failures 2.71E+05 8.07E-09 2.19E-03 8.12E-09 2.20E-03 1.44E-05 7 LERF Failures Induced by 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 --

CDF All CET end states 2.23E-06 0.525 2.23E-06 0.525 1.43E-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.

5-12 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 5.3-2 CPS ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR 1 IN 15 YEAR ILRT/DWBT FREQUENCY EPRI METHODOLOGY PLUS CHANGE DUE TO (2)

EPRI METHODOLOGY CORROSION CORROSION OR ACCIDENT PERSON- DWBT 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 8.58E-07 2.33E-03 8.58E-07 2.33E-03 -3.35E-07 2 Large Isolation Failures 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 --

(Failure to Close) 3a Small Isolation Failures 2.71E+04 5.98E-08 1.62E-03 5.98E-08 1.62E-03 --

3b Large Isolation Failures 2.71E+05 1.21E-08 3.29E-03 1.22E-08 3.32E-03 3.35E-05 7 LERF Failures Induced by 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 --

Phenomena (LERF) 7 non-LERF Failures Induced by 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 Phenomena (non-LERF) 8 Containment Bypass 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 --

CDF All CET end states 2.23E-06 0.526 2.23E-06 0.526 3.32E-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.

5-13 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT 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 core damage frequency (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/DWBT 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 the original 3-in-10 year test interval assessment from Table 5.2-2, the Class 3b frequency is 2.43E-09/yr, which includes the corrosion effect of containment. Based on a ten-year test interval from Table 5.3-1, the Class 3b frequency is 8.12E-09/yr; and, based on a fifteen-year test interval from Table 5.3-2, it is 1.22E-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 3 to 15 years (including corrosion effects) is 9.81E-09/yr.

Similarly, the increase due to increasing the interval from 10 to 15 years (including corrosion effects) is 4.12E-09/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 year results to the original 3-in-10 year requirement.

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 conditional containment failure probability (CCFP). The change in CCFP is indicative of the effect of the ILRT/DWBT 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 5-14 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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).

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 CCFP CCFP15-3 CCFP15-10 3 IN 10 YRS 1 IN 10 YRS 1 IN 15 YRS 58.41% 58.66% 58.84% 0.44% 0.18%

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

The change in CCFP of approximately 0.5% as a result of extending the test interval to 15 years from the original 3-in-10 year requirement is judged to be relatively 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 3-in-10 years to 1-in-15 years all are below the acceptance criteria defined in Section 1.3, namely:

1. Change in LERF = 9.81E-9/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 3.80E-3 person-rem/yr (0.73%), which is less than 1.0 person-rem/year or 1% of the total population dose.

3. Change in CCFP is 0.44%, which is less than 1.5%.

5-15 C467150095-12660-12/1/15

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

BASE, 3 TO 10, AND 3 TO 15 YR EXTENSIONS (INCLUDING AGE ADJUSTED STEEL CORROSION LIKELIHOOD)

BASE CASE EXTEND TO EXTEND TO 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARS EPRI DOSE PERSON- PERSON- PERSON-CLASS PER-REM CDF (1/YR) REM/YR CDF (1/YR) REM/YR CDF (1/YR) REM/YR 1 2.71E+03 9.16E-07 2.49E-03 8.82E-07 2.39E-03 8.58E-07 2.33E-03 2 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 2.68E-07 1.47E-01 3a 2.71E+04 1.20E-08 3.25E-04 3.98E-08 1.08E-03 5.98E-08 1.62E-03 3b 2.71E+05 2.43E-09 6.60E-04 8.12E-09 2.20E-03 1.22E-08 3.32E-03 7 LERF 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 1.14E-07 6.27E-02 7 non-LERF 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 9.16E-07 3.09E-01 8 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 1.55E-09 8.49E-04 Total 2.23E-06 0.523 2.23E-06 0.525 2.23E-06 0.526 ILRT Dose Rate from 3a 9.85E-04 3.29E-03 4.95E-03 and 3b Delta From 3 yr --- 2.21E-03 3.80E-03 Total Dose From 10 yr --- --- 1.59E-03 Rate (1) 3b Frequency (LERF) 2.43E-09 8.12E-09 1.22E-08 Delta 3b From 3 yr --- 5.69E-09 9.81E-09 LERF From 10 yr --- --- 4.12E-09 CCFP % 58.41% 58.66% 58.84%

Delta From 3 yr --- 0.26% 0.44%

CCFP %

From 10 yr --- --- 0.18%

Note to Table 5.6-1:

1. The overall difference in total dose rate is less than the difference of only the 3a and 3b categories between two testing intervals. This is because the overall total dose rate includes contributions from other categories that do not change as a function of time, e.g., the EPRI Class 2 and 8 categories, and also due to the fact that the Class 1 person-rem/yr decreases when extending the IRLT/DWBT frequency.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 5.7 CONTRIBUTIONS FROM OTHER HAZARD GROUPS 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. Appendix A provides a technical adequacy assessment of the Seismic Core Damage Frequency (SCDF) and the Fire PRA. For this ILRT/DWBT risk assessment, contributions from other hazard groups are addressed using a CDF multiplier approach. This approach has been used in previous ILRT submittals [28, 39]

Internal Fire Risk [8]

CPS has a Fire PRA model that was updated in 2014. The total Fire PRA CDF reported is 6E-06/yr [8], which is a factor of 1.8 higher than the Fire CDF (i.e., 3.26E-06/yr) calculated in the CPS Individual Plant Examination of External Events (IPEEE)

[37]. The Fire PRA LERF reported is 9.21E-07/yr. In addition to modeling limitations, the Fire PRA may be subject to more modeling uncertainty than the internal events PRA evaluations. While the 2014 Fire PRA is based on the guidance of NUREG/CR-6850

[38] and is generally self-consistent within its calculational framework, the Fire PRA CDF results do not compare well with internal events PRAs because of the number of conservative assumptions that have been included in the Fire PRA process. Therefore, direct use of the Fire PRA results as a reflection of CDF may be inappropriate, and the actual fire CDF may be overestimated. In any event, the reported Fire CDF value from the 2014 Fire PRA Update is used as a bounding value for this calculation.

Seismic Risk [9]

A quantifiable seismic PRA model for Clinton has not yet been approved for general use in risk applications. However, a Clinton seismic risk analysis was performed as part of the CPS IPEEE [37]. Clinton performed a seismic margins assessment (SMA) following the guidance of NUREG-1407 [35] and EPRI NP-6041 [36]. The SMA is a deterministic evaluation process that does not calculate risk on a probabilistic basis. No core 5-17 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval damage frequency sequences were quantified as part of the IPEEE seismic risk evaluation.

The conclusions of the Clinton IPEEE seismic risk analysis are as follows:

No improvements to the plant were identified as a result of the Seismic Margins Assessment the plant was determined to be fully capable of attaining safe shutdown conditions after the Review Level Earthquake (RLE).

However, more recent information is available from the NRC. A Risk Assessment for NRC GI-199 Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States (CEUS) on Existing Plants, [9], Table D-1 lists the postulated core damage frequencies using the updated 2008 USGS Seismic Hazard Curves. For Clinton, two reference sources are provided for the seismic hazard (i.e.

NUREG/CR-0098 and UHS). A seismic CDF value of 1.7E-5/yr is selected for this ILRT/DWBT risk assessment because it is the higher identified value (i.e. Clinton (UHS)) in Reference [9]. The NRC study did not calculate LERF.

Other External Events In addition to internal fires and seismic events, the CPS IPEEE Submittal analyzed a variety of other external hazards:

• High Winds/Tornadoes

• External Flooding

• Transportation and Nearby Facility Accidents

• 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 5-18 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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/DWBT interval extension risk assessment.

Other Hazard Group Contributor Summary The method chosen to account for external events contributions is similar to that used in the other ILRT interval extension analyses [28, 39] in which a multiplier is applied to the internal events results. The contributions of the external events from various CPS analysis are summarized in Table 5.7-1.

TABLE 5.7-1 OTHER HAZARD GROUP CONTRIBUTOR

SUMMARY

OTHER HAZARD INITIATOR GROUP CDF (1/YR)

Seismic [9] 1.7E-05 Internal Fire [8] 6.0E-06 High Winds/Tornadoes Screened External Floods Screened Transportation and Nearby Facility Accidents Screened Total (for initiators with CDF available) 2.3E-05 Internal Events CDF 2.23E-06 (1)

External Events Multiplier 10.31 Note to Table 5.7-1:

(1)

The multiple for seismic alone is 7.62.

The EPRI Category 3b frequency for the 3-per-10 year, 1-per-10 year, and 1-per-15 year ILRT/DWBT intervals are shown in Table 5.6-1 as 2.43E-09/yr, 8.12E-09/yr, and 1.22E-08/yr, respectively. Using the other hazard group multiplier of 10.31 for CPS, the change in the LERF risk measure due to extending the ILRT/DWBT from 5-19 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 3-per-10 years to 1-per-15 years, including both internal events and other measurable hazard groups hazards risk, is estimated as shown in Table 5.7-2.

TABLE 5.7-2 CPS 3B (LERF) AS A FUNCTION OF ILRT/DWBT FREQUENCY FOR INTERNAL AND EXTERNAL EVENTS (INCLUDING AGE ADJUSTED STEEL CORROSION LIKELIHOOD) 3B 3B FREQUENCY 3B FREQUENCY (3-PER-10 FREQUENCY (1-PER-15 YEAR (1-PER-10 YEAR YEAR LERF (1)

ILRT/DWBT) ILRT/DWBT) ILRT/DWBT) INCREASE Internal Events Contribution 2.43E-09 8.12E-09 1.22E-08 9.81E-09 Other Hazard Group Contribution (Internal 2.51E-08 8.38E-08 1.26E-07 1.01E-07 Events CDF x 10.31)

Combined 2.75E-08 9.19E-08 1.38E-07 1.11E-07 Note to Table 5.7-2:

(1)

Associated with the change from the baseline 3-per-10 year frequency to the proposed 1-per-15 year frequency.

Thus, the total increase in LERF (measured from the baseline 3-per-10 year ILRT interval to the proposed 1-per-15 year frequency) due to the combined internal and external events contribution is estimated as 1.11E-07/yr, which includes the age adjusted steel corrosion likelihood.

The other metrics for the ILRT/DWBT interval extension risk assessment can be similarly derived using the multiplier approach. The results between the 3-in-10 year interval and the 15 year interval compared to the acceptance criteria are shown in Table 5.7-3. As can be seen, the impacts from including the other hazard group contributors are as follows:

1. Change in LERF = 1.11E-7/yr, which is slightly above the 1.0E-7/yr upper boundary for the very small risk increase as defined in RG 1.174, but at the bottom of the band for small risk increase.

2. Change in population dose rate is 4.30E-2 person-rem/yr (0.73%), which is less than 1.0 person-rem/year or 1% of the total population dose.

3. Change in CCFP is 0.44%, which is less than 1.5%.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Thus, the inclusion of external events does not change the conclusion of the risk assessment. That is, the acceptance criteria are all met such that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years has been demonstrated to be small. Note that a bounding analysis for the total LERF contribution follows Table 5.7-3 to demonstrate that the total LERF value for CPS is less than 1.0E-5/yr consistent with the requirements for a Small Change in risk of the RG 1.174 acceptance guidelines.

TABLE 5.7-3 COMPARISON TO ACCEPTANCE CRITERIA INCLUDING OTHER HAZARD GROUPS CONTRIBUTION FOR CPS LERF PERSON-REM/YR CCFP (1) (1)

CONTRIBUTOR CPS Internal Events 9.81E-9/yr 3.80E-03/yr (0.73%) 0.44%

CPS Other Hazard 1.01E-7/yr 3.92E-02/yr (0.73%) 0.44%

Groups CPS Total 1.11E-7/yr 4.30E-02/yr (0.73%) 0.44%

Acceptance Criteria <1.0E-6/yr <1.0 person-rem/yr or 1.5%

<1.0%

Notes to Table 5.7-3:

(1)

The EPRI Class (1, 2, 7, 8) release Person-Rem/yr are assumed to be the same percentage relative to base risk (0.73%) for internal and external events.

(2)

The Probability of DW and WW leakage due to the ILRT/DWBT extension is assumed the same for both Internal and External Events, therefore the percentage change for CCFP remains constant (0.44%).

The 1.11E-07/yr increase in LERF due to the combined hazard events from extending the CPS ILRT/DWBT frequency from 3-per-10 years to 1-per-15 years falls within Region II between 1E-7 to 1E-6 per reactor year (Small Change in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when the calculated increase in LERF due to the proposed plant change is in the Small Change range, the risk assessment must also reasonably show that the total LERF is less than 1E-5/yr. Similar bounding assumptions regarding the external event contributions that were made above are used for the total LERF estimate.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval From Table 4.2-2, the total LERF due to postulated internal event accidents is the sum of the LERF release categories, which is 1.16E-7/yr. For Fire, the total LERF is 9.21E-07/yr [8]. The base LERF due to seismic is assumed to be in the same proportion as the internal events contribution. The total LERF value for CPS is then shown in Table 5.7-4.

TABLE 5.7-4 IMPACT OF 15-YR ILRT EXTENSION ON LERF (3B)

FOR CPS Internal Events LERF 1.16E-07/yr Internal Fire LERF 9.21E-07/yr Other Hazard Group LERF 8.84E-07/yr (Internal Events LERF x 7.62)

Internal Events LERF due to ILRT (1) 1.22E-08/yr (Class 3b) at 15 years Other Hazard group LERF due to (1) 1.26E-07/yr ILRT at 15 years Total 2.06E-06/yr Note to Table 5.7-4:

(1)

Including age adjusted steel corrosion likelihood.

As can be seen, the estimated LERF for CPS using the Fire LERF and CDF based multiplier approach for seismic is 2.1E-06/yr, which is less than the RG 1.174 required value of 1E-5/yr.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 6.0 SENSITIVITIES 6.1 SENSITIVITY TO CORROSION IMPACT ASSUMPTIONS The results in Tables 5.2-2, 5.3-1, and 5.3-2 show that including corrosion effects calculated using the assumptions described in Section 4.5 does not significantly affect the results of the ILRT/DWBT extension risk assessment. Sensitivity cases were developed to gain an understanding of the sensitivity of the results to the key parameters in the corrosion risk analysis. The time for the flaw likelihood to double was adjusted from every five years to every two and every ten years. The failure probabilities for the cylinder, dome and basemat were increased and decreased by an order of magnitude. The total detection failure likelihood was adjusted from 10% to 15%

and 5%. The results are presented in Table 6.1-1. In every case, the impact from including the corrosion effects is minimal. Even the upper bound estimates with conservative assumptions for all of the key parameters yield increases in LERF due to corrosion of only 3.63E-09/yr. The results indicate that even with conservative assumptions, the conclusions from the base analysis would not change.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 6.1-1 STEEL CORROSION SENSITIVITY CASES INCREASE IN CLASS 3B FREQUENCY (LERF)

FOR ILRT/DWBT EXTENSION VISUAL FROM 3 IN 10 TO 1 IN 15 YEARS INSPECTION & (PER YEAR)

CONTAINMENT NON-VISUAL AGE BREACH FLAWS (STEP 3 IN THE (STEP 4 IN THE (STEP 5 IN THE CORROSION CORROSION CORROSION TOTAL INCREASE DUE TO ANALYSIS) ANALYSIS) ANALYSIS) INCREASE CORROSION Base Case Base Case Base Case (1.0% Cylinder- (10% Cylinder-Doubles every Dome, Dome, 9.81E-09 1.14E-10 5 yrs 0.1% Basemat) 100% Basemat)

Doubles every Base Base 9.95E-09 2.59E-10 2 yrs Doubles every Base Base 9.79E-09 9.56E-11 10 yrs 15% Cylinder-Base Base 9.85E-09 1.59E-10 Dome 5% Cylinder-Base Base 9.76E-09 6.88E-11 Dome 10% Cylinder-Base Dome, 1% Base 1.08E-08 1.14E-09 Basemat 0.01% Cylinder-Base Dome, 0.001% Base 9.70E-09 1.14E-11 Basemat Lower Bound 0.1% Cylinder- 5% Cylinder-Doubles every Dome, 0.001% Dome 100% 9.70E-09 5.74E-12 10 yrs Basemat Basemat Upper Bound 10% Cylinder- 15% Cylinder-Doubles every Dome, 1% Dome 1.33E-08 3.63E-09 2 yrs Basemat 100% Basemat 6-2 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 6.2 EPRI EXPERT ELICITATION SENSITIVITY An expert elicitation was performed to reduce excess conservatisms in the data associated with the probability of undetected leaks within containment [3]. Since the risk impact assessment of the extensions to the ILRT interval is sensitive to both the probability of the leakage as well as the magnitude, it was decided to perform the expert elicitation in a manner to solicit the probability of leakage as a function of leakage magnitude. In addition, the elicitation was performed for a range of failure modes which allowed experts to account for the range of failure mechanisms, the potential for undiscovered mechanisms, inaccessible areas of the containment as well as the potential for detection by alternate means. The expert elicitation process has the advantage of considering the available data for small leakage events, which have occurred in the data, and extrapolate those events and probabilities of occurrence to the potential for large magnitude leakage events.

The basic difference in the application of the ILRT interval methodology using the expert elicitation is a change in the probability of pre-existing leakage within containment. The base case methodology uses the Jeffreys non-informative prior for the large leak size and the expert elicitation sensitivity study uses the results from the expert elicitation. In addition, given the relationship between leakage magnitude and probability, larger leakage that is more representative of large early release frequency can be reflected.

For the purposes of this sensitivity, the same leakage magnitudes that are used in the base case methodology (i.e., 10La for small and 100La for large) are used here. Table 6.2-1 illustrates the magnitudes and probabilities of a pre-existing leak in containment associated with the base case and the expert elicitation statistical treatments. These values are used in the ILRT interval extension for the base methodology and in this sensitivity case. Details of the expert elicitation process, including the input to expert elicitation as well as the results of the expert elicitation, are available in the various appendices of EPRI TR-1018243 [3].

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 6.2-1 EPRI EXPERT ELICITATION RESULTS EXPERT ELICITATION MEAN PROBABILITY PERCENT LEAKAGE SIZE (La) BASE CASE OF OCCURRENCE [3] REDUCTION 10 9.2E-03 3.88E-03 58%

100 2.3E-03 2.47E-04 89%

The summary of results using the expert elicitation values for probability of containment leakage is provided in Table 6.2-2. As mentioned previously, probability values are those associated with the magnitude of the leakage used in the base case evaluation (10La for small and 100La for large). The expert elicitation process produces a relationship between probability and leakage magnitude in which it is possible to assess higher leakage magnitudes that are more reflective of large early releases; however, these evaluations are not performed in this particular study.

The net effect is that the reduction in the multipliers shown above has the same impact on the calculated increases in the LERF values. The increase in the overall value for LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is 1.15E-09/yr. Similarly, the increase due to increasing the interval from 10 to 15 years is 5.01E-10/yr. As such, if the expert elicitation mean probabilities of occurrence are used instead of the non-informative prior estimates, the change in LERF for CPS is much further within the range of a very small change in risk when compared to the current 1-in-10, or baseline 3-in-10 year requirement. The results of this sensitivity study are judged to be more indicative of the actual risk associated with the ILRT extension than the results from the assessment as dictated by the values from the EPRI methodology [3], and yet are still conservative given the assumption that all of the Class 3b contribution is considered to be LERF.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 6.2-2 CPS ILRT/DWBT CASES:

3 IN 10 (BASE CASE), 1 IN 10, AND 1 IN 15 YR INTERVALS (ILRT LEAKAGE BASED ON EPRI EXPERT ELICITATION PROBABILITIES)

BASE CASE EXTEND TO EXTEND TO 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARS EPRI DOSE PERSON- PERSON- PERSON-CLASS PER-REM CDF (1/YR) REM/YR CDF (1/YR) REM/YR CDF (1/YR) REM/YR 1 2.71E+03 9.23E-07 2.51E-03 9.08E-07 2.46E-03 8.96E-07 2.43E-03 2 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 2.68E-07 1.47E-01 3a 2.71E+04 6.44E-09 1.75E-04 2.15E-08 5.82E-04 3.22E-08 8.74E-04 3b 2.71E+05 2.67E-10 7.25E-05 9.12E-10 2.48E-04 1.41E-09 3.84E-04 7 LERF 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 1.14E-07 6.27E-02 7 non-LERF 3.37E+05 9.16E-07 3.09E-01 9.16E-07 3.09E-01 9.16E-07 3.09E-01 8 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 1.55E-09 8.49E-04 Total 2.23E-06 0.522 2.23E-06 0.522 2.23E-06 0.523 ILRT Dose Rate from 3a 2.47E-04 8.30E-04 1.26E-03 and 3b Delta From 3 yr --- 5.38E-04 9.37E-04 Total Dose From 10 yr --- --- 3.99E-04 Rate (1) 3b Frequency (LERF) 2.67E-10 9.12E-10 1.41E-09 Delta 3b From 3 yr --- 6.45E-10 1.15E-09 LERF From 10 yr --- --- 5.01E-10 CCFP % 58.31% 58.34% 58.36%

Delta From 3 yr --- 0.03% 0.05%

CCFP %

From 10 yr --- --- 0.02%

Note to Table 6.2-2:

(1)

The overall difference in total dose rate is less than the difference of only the 3a and 3b categories between two testing intervals. This is because the overall total dose rate includes contributions from other categories that do not change as a function of time, e.g., the EPRI Class 2 and 8 categories, and also due to the fact that the Class 1 person-rem/yr decreases when extending the IRLT/DWBT frequency.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 6.3 DWBT DATA SENSITIVITY An additional sensitivity is included related to the interpretation of the DWBT data used for the base case assessment. The probability of small and large drywell failures is increased by a factor of 10. The base case applied the containment failure probabilities to the drywell failure probabilities. The base case probabilities are considered conservative 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 NEI/EPRI approach is reflective of drywell failures. The body of plant experience used considered the older BWR containment designs.

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

• 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 4.1-1. As this comparison shows, the basic designs are much the same and therefore would be expected to have much the same leakage failure mechanisms.

• 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.

TABLE 6.3-1 DW LEAKAGE X10 PROBABILITY SENSITIVITY DW SENSITIVITY DW WW WW LEAKAGE DW LEAKAGE PROBABILITY LEAKAGE LEAKAGE SIZE (LB) LEAKAGE PROBABILITY SIZE (LA)

(BASE) PROBABILITY 10Lb 0.0092 0.092 10La 0.0092 100Lb 0.0023 0.023 100Lb 0.0023 Note: WW Leakage (10La and 100La remains constant) 6-6 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The summary of results using the revised values for probability of drywell bypass leakage is provided in Table 6.3-2. The results indicate increases to the population dose and to the CCFP values compared to the base risk assessment, but the results are all still within the acceptance criteria of less than 1.0 person-rem/yr or less than 1.0% person-rem/yr, and less than 1.5% change in CCFP.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval TABLE 6.3-2 CPS ILRT/DWBT CASES:

3 IN 10 (BASE CASE), 1 IN 10, AND 1 IN 15 YR INTERVALS (DWBT 10LB, 100LB LEAK PROB. INCREASED BY A FACTOR OF 10)

BASE CASE EXTEND TO EXTEND TO 3 IN 10 YEARS 1 IN 10 YEARS 1 IN 15 YEARS EPRI DOSE PERSON- PERSON- PERSON-CLASS PER-REM CDF (1/YR) REM/YR CDF (1/YR) REM/YR CDF (1/YR) REM/YR 1 2.71E+03 8.94E-07 2.43E-03 8.11E-07 2.20E-03 7.51E-07 2.04E-03 2 5.48E+05 2.68E-07 1.47E-01 2.68E-07 1.47E-01 2.68E-07 1.47E-01 3a 2.71E+04 3.31E-08 8.99E-04 1.10E-07 2.99E-03 1.66E-07 4.50E-03 3b 2.71E+05 2.63E-09 7.14E-04 8.78E-09 2.38E-03 1.32E-08 3.59E-03 7 LERF 5.48E+05 1.14E-07 6.27E-02 1.14E-07 6.27E-02 1.14E-07 6.27E-02 7 non-LERF 3.37E+05 9.16-07 3.09E-01 9.16E-07 3.09E-01 9.16E-07 3.09E-01 8 5.48E+05 1.55E-09 8.49E-04 1.55E-09 8.49E-04 1.55E-09 8.49E-04 Total 2.23E-06 0.523 2.23E-06 0.527 2.23E-06 0.530 ILRT Dose Rate from 3a 1.61E-03 5.38E-03 8.09E-03 and 3b Delta From 3 yr --- 3.54E-03 6.09E-03 Total Dose From 10 yr --- --- 2.55E-03 Rate (1) 3b Frequency (LERF) 2.63E-09 8.78E-09 1.32E-08 Delta 3b From 3 yr --- 6.15E-09 1.06E-08 LERF From 10 yr --- --- 4.45E-09 CCFP % 58.41% 58.69% 58.89%

Delta From 3 yr --- 0.28% 0.48%

CCFP %

From 10 yr --- --- 0.20%

Note to Table 6.3-2:

(1)

The overall difference in total dose rate is less than the difference of only the 3a, and 3b categories between two testing intervals. This is because the overall total dose rate includes contributions from other categories that do not change as a function of time, e.g., the EPRI Class 2 and 8 categories, and also due to the fact that the Class 1 person-rem/yr decreases when extending the IRLT/DWBT frequency.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

7.0 CONCLUSION

S Based on the results from Section 5 and the sensitivity calculations presented in Section 6, the following conclusions regarding the assessment of the plant risk are associated with permanently extending the Type A ILRT test frequency and the DWBT frequency to fifteen years:

• 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 10-6/yr and increases in LERF below 10-7/yr. Small changes in risk are defined as increases in CDF below 10-5/yr and increases in LERF below 10-6/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 and the DWBT interval for the base case with corrosion included is 9.81E-09/yr (see Table 6.1-1), which falls within the very small change region of the acceptance guidelines in Reg.

Guide 1.174.

If the EPRI Expert Elicitation methodology Class 3a and Class 3b failure probabilities are used, the change is estimated as 1.15E-09/yr (see Table 6.2-2), which falls further within the very small change region of the acceptance guidelines in Reg. Guide 1.174.

• The change in dose risk for changing the Type A ILRT interval and the DWBT interval from three-per-ten years to once-per-fifteen-years, measured as an increase to the total integrated dose risk for all accident sequences, is 3.80E-03 person-rem/yr using the EPRI guidance with the base case corrosion included (see Table 5.6-1). This change meets both of the related acceptance criteria identified in Section 1.3 for change in population dose of less than 1.0 person-rem/ year or less than 1% person-rem/yr.

The change in dose risk drops to 9.37E-04 person-rem/yr when using the EPRI Expert Elicitation methodology (see Table 6.2-2). The change in dose risk meets both of the related acceptance criteria identified in Section 1.3 for change in population dose of less than 1.0 person-rem/ year or less than 1% person-rem/yr.

• The increase in the conditional containment failure frequency from the three in ten year interval to one in fifteen years including corrosion effects using the EPRI guidance (see Table 5.6-1) is 0.44%, which is below the acceptance criteria of 1.5% identified in the NRC SER on the issue [7] as discussed in Section 1.3.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The increase in CCFP drops to about 0.05% using the EPRI Expert Elicitation methodology (see Table 6.2-2). This value meets both of the related acceptance criteria identified in Section 1.3 for change in CCFP of less than 1.5%.

• To determine the potential impact from other hazard groups, an additional bounding assessment from the risk associated with the other relevant hazard groups for CPS utilizing the latest information from various sources was performed. As shown in Table 5.7-3, the total increase in LERF due to internal events and other hazard groups is 1.11E-07/yr, which is in Region II of the Reg. Guide 1.174 acceptance guidelines. As also shown in Table 5.7-3, the other acceptance criteria for change in population dose and change in CCFP are also still met when the other hazard groups are considered in the analysis.

• Finally, as shown in Table 5.7-4, a similar bounding analysis for the other hazard groups indicates that the total LERF from both internal events and the other hazard groups is 2.06E-06/yr, which is less than the Reg. Guide 1.174 limit of 1E-05/yr given that the LERF is in Region II (small change in risk).

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

Previous Assessments The NRC in NUREG-1493 [6] has previously concluded the following:

• 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.

• 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.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval The findings for CPS confirm these general findings on a plant specific basis for the ILRT/DWBT 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/DWBT Interval

8.0 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 2, May 2011.

[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 2014 Fire PRA Summary and Quantification Notebook, CPS-PSA-021-06, Revision 1, December 2014.

[9] U.S. Nuclear Regulatory Commission (NRC), Memorandum to Brian W. Sheron, Director Office of Nuclear Regulatory Research, From Patrick Hiland, Chairman, Safety/Risk Assessment Panel for Generic Issue 199: Safety/Risk Assessment Results for Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates In Central and Eastern United States on Existing Plants, Accession Number ML11356A034, September 2, 2010.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval

[10] Letter U-602549 from Wilfred Connell (Vice President CPS) to the NRC, Subject Clinton Power Station Proposed Amendment of Facility Operating License No.

NPF-62 (LS-96-001), February 22, 1996.

[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.

[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. 16, January, 2014.

[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 8-2 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval 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 (2014 PRA Interim Update), CPS PSA-013, Revision 4, March 2014.

[25] CPS PRA Quantification Notebook, CPS PSA-014, Revision 6, March 2014.

[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] Letter from Thomas P. Kirwin (Entergy, Palisades Nuclear Plant) to U.S. Nuclear Regulatory Commission, License Amendment Request to Extend the Containment Type A Leak Rate Test Frequency to 15 Years, Accession Number ML110970616, April 6, 2011.

[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 2, March 2014.

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[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] CPS Technical Specification Section 3.6 of Amendment No. 158.

[35] U.S. Nuclear Regulatory Commission, Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, NUREG-1407, June 1991.

[36] Electric Power Research Institute, A Methodology for Assessment of Nuclear Power Plant Seismic Margin. EPRI NP-6041-LS, Rev. 1, August 1991.

[37] Clinton Power Station Individual Plant Examination For External Events, September 1995.

[38] EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, EPRI 1011989, NUREG/CR-6850, September 2005.

[39] Letter from James Barstow (Exelon Generation Company, LLC) 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" (Peach Bottom Nuclear Station), Accession Number ML14315A084, November 7, 2014.

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

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy APPENDIX A PRA TECHNICAL ADEQUACY C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A

A.1 OVERVIEW A technical Probabilistic Risk Assessment (PRA) analysis is presented in this report to help support an extension of the Clinton Unit 1 containment Type A test integrated leak rate test (ILRT) interval to fifteen years. The analysis follows the guidance provided in Regulatory Guide 1.200, Revision 2 [A-3], An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities. The guidance in RG-1.200 indicates that the following steps should be followed to perform this study:

1. Identify the parts of the PRA used to support the application SSCs, operational characteristics affected by the application and how these are implemented in the PRA model.

A definition of the acceptance criteria used for the application.

2. Identify the scope of risk contributors addressed by the PRA model If not full scope (i.e. internal and external), identify appropriate compensatory measures or provide bounding arguments to address the risk contributors not addressed by the model.

3. Summarize the risk assessment methodology used to assess the risk of the application Include how the PRA model was modified to appropriately model the risk impact of the change request.

4. Demonstrate the Technical Adequacy of the PRA Identify plant changes (design or operational practices) that have been incorporated at the site, but are not yet in the PRA model and justify why the change does not impact the PRA results used to support the application.

Document peer review findings and observations that are applicable to the parts of the PRA required for the application, and for those that have not yet been addressed justify why the significant contributors would not be impacted.

Document that the parts of the PRA used in the decision are consistent with applicable standards endorsed by the Regulatory Guide. Provide justification to show that where specific requirements in the standard are not met, it will not unduly impact the results.

Identify key assumptions and approximations relevant to the results used in the decision-making process.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy Items 1 through 3 are covered in the main body of this report. The purpose of this appendix is to address the requirements identified in item 4 above. Each of these items (plant changes not yet incorporated into the PRA model, relevant peer review findings, consistency with applicable PRA standards and the identification of key assumptions) are discussed in the following sections.

The risk assessment performed for the ILRT extension request is based on the current Level 1 and Level 2 PRA model. Note that for this application, the accepted methodology involves a bounding approach to estimate the change in the PRA risk metric of LERF from extending the ILRT interval. Rather than exercising the PRA model itself, it involves the establishment of separate evaluations that are linearly related to the plant CDF contribution. Consequently, a reasonable representation of the plant CDF that does not result in a LERF does not require that Capability Category II be met in every aspect of the modeling if the Category I treatment is conservative or otherwise does not significantly impact the results.

A discussion of the Exelon model update process, the peer reviews performed on the Clinton PRA model, the results of those peer reviews and the potential impact of peer review findings on the ILRT extension risk assessment are provided in Section A.2.

Section A.3 provides a qualitative assessment of External Hazards Fire and Seismic inputs. Section A.4 provides an assessment of key assumptions and approximations used in this assessment and Section A.5 briefly summarizes the results of the PRA technical adequacy assessment with respect to this application.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.2 PRA Model Evolution and Peer Review Summary A.2.1 Introduction The 2014A version of the CPS PRA model is the most recent evaluation of the Unit 1 risk profile at CPS for internal event challenges. The CPS PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the CPS PRA is based on the event tree/fault tree methodology, which is a well-known methodology in the industry.

Exelon Generation Company, LLC (Exelon) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the 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. The following information describes this approach as it applies to the CPS PRA.

PRA Maintenance and Update The Exelon risk management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plant. This process is defined in the Exelon Risk Management program, which consists of a governing procedure and subordinate implementation procedures. The PRA model update procedure delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating Exelon nuclear generation sites. The overall Exelon Risk Management program defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following activities are routinely performed:

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• Design changes and procedure changes are reviewed for their impact on the PRA model.

• Maintenance unavailabilities are captured, and their impact on CDF is trended.

• Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated approximately every four years.

In addition to these activities, Exelon risk management procedures provide the guidance for particular risk management maintenance activities. This guidance includes:

• Documentation of the PRA model, PRA products, and bases documents.

• The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.

• Guidelines for updating the full power, internal events PRA models for Exelon nuclear generation sites.

• Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10 CFR 50.65(a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately 4-year cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant.

The 2014A model was completed in March of 2014.

As indicated previously, RG 1.200 also requires that additional information be provided as part of the LAR submittal to demonstrate the technical adequacy of the PRA model used for the risk assessment. Each of these items (plant changes not yet incorporated into the PRA model, relevant peer review findings, and consistency with applicable PRA Standards) will be discussed in turn in this section.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.2.2 Plant Changes Not Yet Incorporated into the PRA Model A PRA updating requirements evaluation (URE) documented in the Clinton PRA model update tracking database is created for all issues that are identified that could impact the PRA model. The URE database includes the identification of those plant changes that could impact the PRA model.

A review of the open UREs indicates that there are no plant changes that have not yet been incorporated into the PRA model that would affect this application. UREs are evaluated for potential impact to applications and to the PRA base model results and are classified as High, Medium or Low priority. High priority items could significantly impact applications. Medium priority items are items that are assessed as potentially important to applications and Low priority items area items that are assessed as not important to applications and likely to have minimal or no numeric impact. There are no open High priority UREs and seventeen UREs identified as Medium priority. The remaining open UREs are low priority, having little or no impact to the PRA results.

Medium priority UREs were found not to impact this application. Low priority UREs were also reviewed and none were found that would impact this application.

A.2.3 Consistency with Applicable PRA Standards The Clinton FPIE PRA model has undergone several reviews including a BWROG Peer Review in 2000 [A-5, 6, and 7]. UREs were created and extensive changes were made to the PRA model in updates through 2006. As a result of the extensive changes made to the model a full Peer review was again performed in 2009 [A-12]. The results of the 2009 Peer review and the actions taken to address gaps identified, best represent the consistency of the model to current PRA standards.

The 2009 Peer Review was conducted using the 2009 ASME/ANS PRA Standard [A-1]

and the NRCs comments and clarifications contained in RG 1.200 [A-3]. The Peer Review was conducting using the CPS 2006C FPIE PRA model [A-4]. The general objective for the CPS PRA is to meet Capability Category II. The findings and A-5 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy observations (F&Os) that were identified in the Peer Review were between the 2006C CPS PRA and the requirements for Capability Category II. These F&Os (i.e., both findings and suggestions) were entered into the CPS Updating Requirements Evaluation (URE) database for tracking purposes. These UREs were used for scoping of the Clinton 2011 PRA update.

All Findings from the 2009 Peer Review were addressed as part of the 2011 PRA update. The 2009 CPS Peer Review observations were incorporated into the CPS URE database for tracking. All but one of the Suggestion observations have been addressed.

Following the 2009 Peer review, a self-assessment relative to the combined ASME/ANS PRA Standard [A-1] and the NRCs comments and clarifications contained in RG 1.200.

[A-3] was performed as part of the 2011 PRA update. The 2011 Self-Assessment used the 2009 Peer review results as input. Table A-2 lists gaps to Category II identified in the 2011 self-assessment and the status of those gaps following the 2011 and 2014 PRA updates. Included in the last column of Table A-2 is the URE number, a significance statement and the impact to this ILRT/DWBT for the gaps that have not been addressed. These gaps are judged to not have an impact on this application as justified in Table A-2.

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SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 1 Review initiating event precursors in identifying the IE-A7 Completed: URE CL2009-006 has been closed. This initiating events to be modeled. documentation aspect has been incorporated into the CPS PRA Initiating Event Notebook. This work included review of hundreds of events INPO SENs, SOERs, A rigorous explicit assessment of all the events in NUREG-SERs, and NRC SECY letters on precursors, as well as 1275 could be pursued (if determined that this is the true CPS specific experience records. No new initiating intent of SR IE-A7); however, such an effort is judged not event categories were identified.

to provide much benefit to the CPS IE analysis.

2 Loss of switchgear room cooling assumptions should be IE-C4 Deferred: Switchgear room cooling calculations were supported by room cooling calculation. not performed prior to completion of the 2014 PRA Update..This item is deferred and maintained here for future consideration.

Tracked under URE L2010-012. The current modeling assumptions regard for the need for room cooling in the long term are judged realistic. Performance of loss of cooling calculations is not expected to impact the model or results. No impact to the CPS ILRT/DWBT application.

3 Complete CPS URE 2001-055 (recommendation to AS-B3, Completed: URE CL2001-055 has been closed. The perform loss of room cooling calculations to support PRA SC-B2, documentation exists in the PRA Dependency Notebook success criteria assumptions). SC-C1, (CPS PSA-006). Room cooling assumptions were SC-C2, confirmed with the system manager during system SY-A19, manager interviews and are judged to be reasonable.

SY-B7 4 Complete CPS URE 2001-061 (concerning SY-A17, Completed: URE CL2001-061 has been closed. The recommendation to specifically perform loss of room SY-B8 documentation exists in the PRA Dependency Notebook cooling calculation for switchgear rooms). (CPS PSA-006). Room cooling assumptions were confirmed with the system manager during system manager interviews and are judged to be reasonable.

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SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 5 Complete URE 2001-144 (concerning recommendations to SY-B8 Completed: URE 2001-144 has been closed and the enhance containment isolation documentation). documentation enhanced.

6 To meet the requirements of SR HR-A1, the following HR-A1, Closed: The CPS 2011 updated involved re-assessing would be developed as supporting documentation for CPS: HR-A2, each PRA system for pre-initiator HEPs. The detailed

• A list of the PRA systems to consider for HR-A3, pre-IE identification process is described and test and maintenance actions HR-C2, documented in Appendix J of the HRA notebook (CPS-

• Rules for identifying and screening test HR-C3 PSA-004).

and maintenance actions from the PRA

• A list of procedures reviewed, the potential test and maintenance actions associated with the procedures, and the disposition of the action (screened or evaluated).

• Identify T&M activities that require realignment of the system outside its normal operational or stand by status.

However, performing this task is judged not to have significant impact on the PRA model and results.

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SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 7 Complete URE 2001-084 (concerning use of screening HR-B1, Closed: URE CL2001-084 was closed as part of the values for dominant pre-initiator HEPs). HR-B2 2011 PRA update. Pre-initiators were identified as part of a detailed system analysis described in Appendix J of the Clinton HRA notebook (PSA PSA-004). The pre-Leading candidates for risk significant pre-initiators were initiator HEPs that were identified in App. J were identified through review of the BNL study on sensitivity of calculated using the EPRI HRA Calculator and plant PRA to HEPs, review of other PRAs, and review of plant-specific procedures were available.

specific operating experience. Non-risk significant pre-initiators were screened out. The approach taken is one that reflects the best use of resources by excluding pre-initiators for which data is unavailable and overall contribution is insignificant. To meet the requirements of SR HR-B1, the following would be developed as supporting documentation for CPS:

• A list of the PRA systems to consider for pre-initiator actions.

• Rules for identifying and screening pre-initiator actions from the PRA.

• A list of procedures reviewed, the potential pre-initiator actions associated with the procedures, and the disposition of the action (screened or evaluated).

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SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 8 Complete URE 2001-084. Although this will not HR-D1, Closed: URE CL2001-084 was closed as part of the significantly impact the HRA results, future PRA updates HR-D2, 2011 PRA update. Pre-initiators were identified as part should include an assessment of the quality of plant written HR-D3, of a detailed system analysis described in Appendix J of procedures and administrative controls as well as human- HR-D4 the Clinton HRA notebook (PSA PSA-004). The pre-machine interface for both pre-initiator and post-initiator initiator HEPs that were identified in App. J were human actions. calculated using the EPRI HRA Calculator and plant specific procedures were available.

Alternative:

Possible upgrade to the pre-initiator HRA to include specific quantifications for each pre-initiator HEP would be strict compliance with the standard. This is not considered necessary for most applications. It is recommended that CPS await further ASME clarification on this item before proceeding. This can be confirmed for each application in lieu of performing the quantifications.

9 Failure data development using surveillance test data DA-C10 Deferred: Current industry PRA efforts and PRA peer should fulfill the requirements of DA-C10, and should be reviews are having difficulty understanding the full intent documented appropriately. Review surveillance test of this SR. Future updates of the CPS PRA will procedures and identify all failure modes that are fully consider enhancement to the documentation and tested by the procedures. Include data for the failure investigation of the plant failure data implied by this SR.

modes that are fully tested. The results of unplanned The impact on the overall CDF or LERF values is demands on equipment should also be accounted for. judged to be non-significant.

Tracked by URE CL2009-008, Primarily a documentation/process issue. Impact on overall CDF and LERF expected to be negligible. No impact to the CPS ILRT/DWBT application.

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SUMMARY

OF CLINTON 2006 PRA SELF-ASSESSMENT IDENTIFIED ENHANCEMENTS (Status After Completion of Clinton 2014 PRA Update Provided)

APPLICABL STATUS AFTER COMPLETION OF 2014 UPDATE

1. CL06C SELF-ASSESSMENT RECOMMENDATION E SRS AND IMPACT TO ILRT APPLICATION 10 As needed in maintenance unavailability determination, DA-C13 Completed: All the risk significant maintenance events perform interviews of maintenance staff for equipment with in the CL11a PRA, with the exception of two, are based incomplete or limited maintenance information and on current plant-specific data and are judged document appropriately. reasonable. Two risk significant maintenance events, 1ADASDIV2SRVSM-- and 1APTR-RATSVC-M--, are based on plant-specific data from the previous PRA revision and are judged reasonable. Interviews not warranted.

11 Identify significant basic events that contribute to the QU-D5a Closed: The support system initiating event fault trees significant initiating events whose frequencies are have been incorporated into the single-top CPS 2011 quantified using fault tree methods. model. Basic event importance measures are Otherwise, importance measures calculated and assessed calculated for all SSIE basic events that appear in the to ensure results make logical sense. cutsets.

12 Strict reading of SR QU-F2 would indicate that the QU-F2 Closed: Item (b) has been incorporated into the following enhancements to the documentation of the CPS Quantification Notebook of the CPS 2011 update.

PRA would need to be made to comply with the Standard:

a) Provide a list of human actions and equipment Deferred: Items (a) and (c) are documentation failures (significant basic events) that cause enhancements for the base PRA and are maintained for accidents to be non-dominant. consideration for future updates. Non-significant b) Bases for the elimination of mutually exclusive documentation item.

events from the model need to be added.

c) Include cutsets segregated by accident sequence Tracked by URE CL2009-011. Considered in the documentation. This is available but may documentation enhancements. No impact to the CPS not be needed in the formal documentation. This ILRT/DWBT application.

should await further ASME/ANS clarification before extensive resources are committed.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.3 EXTERNAL HAZARDS Although EPRI report 1018243 [A-10] recommends a quantitative assessment of the contribution of external events (for example, fire and seismic) where a model of sufficient quality exists, it also recognizes that the external events assessment can be taken from existing, previously submitted and approved analyses or another alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. Based on this, currently available information for external events models was referenced, and a multiplier was applied to the internal events results based on the available external events information. This is further discussed in Section 5.7 of the risk assessment. The fire and seismic PRA Technical Adequacy are discussed in additional detail.

A.3.1 FIRE PRA TECHNICAL ADEQUACY The Clinton Power Station (CPS) Fire Probabilistic Risk Assessment (FPRA) is an update/upgrade of the original fire risk assessment performed as part of the plants Individual Plant Examination of External Events (IPEEE) and as previously updated.

The 2014 FPRA update generally uses the fire scenarios developed for the 2008 FPRA update and incorporates them into the 2014 Full Power Internal Events (FPIE) model.

This FPRA is an interim implementation of NUREG/CR-6850 [A-8]. That is, not all tasks identified in NUREG/CR-6850 are completely addressed or implemented in this update due to the limited scope of the current incremental update and due to the changing state-of-the-art of the industry at the time of the 2014 FPRA development. The 2014 FPRA has therefore not received a Peer Review.

The methodologies employed in the FPRA analysis are consistent with those provided in NUREG/CR-6850 [A-8] and Supplement 1 [A-9]; however, no commitment to compliance with the NUREG is made or implied. At this time, there is no commitment for A-12 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy CPS to transition to 50.49(c), Performance based / Risk Informed Fire Protection, or NFPA 805.

A.3.2 Fire PRA Limitations This 2014 Fire PRA Update is an interim implementation of NUREG/CR-6850 [A-8].

That is, not all tasks identified in NUREG/CR-6850 are completely addressed or implemented in this update due to limited scope of the current incremental update or due to the changing state-of-the-art of industry at the time of the 2014 CPS FPRA development. Limitations and other precautions regarding the development of the 2014 FPRA, in terms of the tasks identified in NUREG/CR-6850 [A-8], are as follows:

• Confirmatory Walkdowns (Tasks 1-3, 6) - Confirmatory ignition source and scenario walkdowns for the 2014 PRA update were not conducted due to the limited scope of the project. The results and insights from the 2008 FPRA walkdowns were retained for the 2014 update. These are judged adequate for the ILRT/DWBT risk evaluation.

• Instrumentation Review (Task 2) - The requirements of NUREG/CR-6850 regarding the explicit identification and modeling of instrumentation required to support PRA credited operator actions is not fully addressed.

The instrumentation review from the FPIE PRA is retained for the FPRA update. Available instrumentation is generally redundant and diverse such that it is judged unlikely that more detailed treatment would significantly impact the FPRA, especially for use in the ILRT/DWBT risk evaluation.

• Balance of Plant (Task 2) - The BOP (PCS) is not fully treated. BOP support system failure is conservatively assumed in locations where related components and cables may exist. This represents a potential conservatism in the model, and therefore potential conservativism in the Fire CDF used in the ILRT/DWBT risk evaluation.

• Limited Analysis Iterations (Tasks 9-12) - The process of conducting a FPRA is iterative, identifying conservative assumptions and high risk compartments and performing analyses to refine the assumptions and reduce those compartment risks. The ability to conduct iterations is limited based on resources. The scenarios included in the 2014 CPS FPRA may benefit from additional refinement. This reflects a potential conservativism in the Fire CDF used in the ILRT/DWBT risk evaluation.

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• Multi-Compartment Review (Task 11) - This subtask reviews the fire analysis compartment boundaries to ensure they are sufficiently robust to prevent the spread of fire between FPRA Physical Analysis Units (PAUs) or that such propagations are adequately addressed by the developed scenarios (i.e., multi-compartment fire scenarios). Based on the 2008 FPRA walkdowns, the compartment boundaries abilities to contain the effects of fires were deterministically considered as part of scenario development. Probabilistic failure potential (e.g., failure of fire door due to fire impacts), while not currently included in the FPRA in terms of multi-compartment fire scenarios is judged to be a very small contributor to the overall Fire CDF that is far outweighed by other FPRA conservatisms (e.g., limited credit for balance of plant due to limited cable data). The design and plant layout of CPS makes fire propagation to multiple compartments unlikely compared to the fire risk in individual compartments. Therefore, the FPRA is judged adequate for use in the ILRT/DWBT risk evaluation.

• Seismic Fire Interactions (Task 13) - This task reviews previous assessments to identify any specific interaction between suppression system and credited components or adverse impact of fire protection system interactions that should be accounted for in the FPRA. The seismic fire interactions assessment is typically qualitative in nature and therefore would not impact the ILRT/DWBT results.

• Uncertainty and Sensitivity Analysis (Task 15) - This task explores the impacts of possible variation of input parameters used in the development of the model and the inputs to the analysis on the FPRA results. This task is considered qualitatively for the 2014 FPRA update. These analyses do not impact the ILRT/DWBT risk evaluation.

Given the above, the 2014 Clinton FPRA model is judged to provide a meaningful representation of Fire CDF and LERF contribution, and is appropriate for use in risk-informed decision-making, to the extent that these limitations are recognized and addressed in each application, as appropriate (as performed above for the ILRT/DWBT assessment).

A.3.3 Fire PRA Technical Adequacy for Clinton ILRT/DWBT The conservative methodology described in Section 5.7 that uses a multiplier to obtain ILRT/DWBT impact is considered in the evaluation of Fire PRA Technical Adequacy. A review of limitations reported in the Fire PRA Summary and Quantification Notebook A-14 C467150095-12660-12/1/15

Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy

[A-2] is also considered. The Fire PRA technical adequacy is judged to be adequate to support the conclusions found in Section 7.0 of the main body of this report.

A.3.4 Seismic Core Damage Frequency Estimate Technical Adequacy for Clinton ILRT/DWBT As described in NRC memorandum, Safety/Risk Assessment Results for Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants [A-12]:

In accordance with Management Directive (MD) 6.4, Generic Issues Program, a Safety/Risk Assessment panel was established to:

• Determine, on a generic basis, if the risk associated with Generic Issue (GI) 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States (CEUS) on Existing Plants, warrants further investigation for potential imposition as a cost-justified backfit.

• Provide a recommendation regarding the next step (i.e., should the issue continue to the Regulatory Assessment Stage for identification and evaluation of potential generic, cost justified backfits, be dropped due to low risk, or have other actions taken outside the Generic Issues Program

[GIP]).

The panel reviewed available information, including IPEEE information. Also, as noted in Reference [A-12]:

Approximate SCDF estimates were developed using a method which includes integrating the mean seismic hazard curve and the mean plant-level fragility curve for each NPP. This method, developed by Kennedy (1997), is discussed in Section 10.8.9 of AMSE/ANS RA-Sa-2009 and has previously been used by the staff in the resolution of GI-194, Implications of Updated Probabilistic Seismic Hazard Estimates, and during reviews of various risk-informed license amendments. This approach was discussed with EPRI under an NRC-EPRI seismic research memorandum of understanding. EPRI agreed that this is a reasonable approach for evaluating GI-199.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy The calculated Clinton seismic CDF was derived using two inputs: The NUREG-0098 Review Level Earthquake (RLE) spectrum and a site-specific uniform hazard spectrum (UHS). The following eight SCDF estimates were developed from each set of seismic hazard curves (NUREG-0098 and UHS) [A-12]:

1. SCDFpga - integration of the pga-based seismic hazard and plant-level fragility curves.

2. SCDF10 - integration of the 10-Hz seismic hazard and plant-level fragility curves.

3. SCDF5 - integration of the 5-Hz seismic hazard and plant-level fragility curves.

4. SCDF1 - integration of the 1-Hz seismic hazard and plant-level fragility curves.

5. SCDFmax - maximum of the SCDFpga, SCDF10, SCDF5, and SCDF1 estimates.

6. SCDFavg - simple average of the SCDFpga, SCDF10, SCDF5, and SCDF1 estimates.

7. SCDFIPEEE - weighted average of the SCDFpga, SCDF10, SCDF5, and SCDF1 estimates, where the weights were obtained from Appendix A of NUREG-1407 (SCDFpga was weighted by one-seventh and the other SCDF estimates were weighted by two-sevenths).

8. SCDFwl - SCDF estimate based on the weakest link model described in Appendix A Clinton Seismic CDF results for the eight categories and two hazard spectrums (16 SCDF values) ranged from 1.1E-06 to 1.7E-05. To conservatively estimate the Clinton Seismic CDF for the ILRT/DWBT Other Hazards impact evaluation (Section 5.7 of the Main Body), the highest SCDF of 1.7E-05 was chosen. This appears to be the most recent and conservative SCDF representing Clinton Stations seismic risk. The Seismic CDF technical adequacy is judged to be adequate to support the conclusions found in Section 7.0 of the main body of this report.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.4 IDENTIFICATION OF KEY ASSUMPTIONS The methodology employed in this risk assessment followed the EPRI guidance [A-10]

as previously approved by the NRC. The analysis included the incorporation of several sensitivity studies and factored in the potential impacts from external events in a bounding fashion. None of the sensitivity studies or bounding analyses indicated any source of uncertainty or modeling assumption that would have resulted in exceeding the acceptance guidelines. Since the accepted process utilizes a bounding analysis approach which is mostly driven by CDF contribution that does not already lead to LERF, there are no identified key assumptions or sources of uncertainty for this application (i.e. those which would change the conclusions from the risk assessment results presented here).

A.5

SUMMARY

A PRA technical adequacy evaluation was performed consistent with the requirements of RG-1.200, Revision 2. This evaluation combined with the details of the results of this analysis demonstrate with reasonable assurance that the proposed extension to the ILRT/DWBT interval for CPS Unit 1 to fifteen years satisfies the risk acceptance guidelines in RG 1.174.

The Fire PRA results and Seismic CDF inputs were used to bound external event impacts of the proposed extension. The technical adequacy of the Fire PRA Model and the Seismic CDF input were qualitatively assessed and found to be adequate to support the conclusions found in Section 7.0 of this document.

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Risk Impact Assessment of Extending the CPS ILRT/DWBT Interval Appendix A - PRA Technical Adequacy A.6 REFERENCES

[A-1] ASME/American Nuclear Society, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009, March 2009.

[A-2] CPS 2014 Fire PRA Summary and Quantification Notebook, CPS-PSA-021-06, Revision 1, December 2014.

[A-3] NRC Regulatory Guide, RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Rev. 2, March 2009.

[A-4] CPS Station Internal Events PRA, Model of Record 2006C.

[A-5] NEI 00-02, Probabilistic Risk Assessment Peer Review Process Guidance, Rev. A3, March 2000.

[A-6] BWROG PSA Peer Review Certification Implementation Guidelines, January 1997.

[A-7] Clinton PRA Peer Review Report, October, 2000.

[A-8] EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, EPRI 1011989 - NUREG/CR-6850, September 2005.

[A-9] Fire Probabilistic Risk Assessment Methods Enhancements: Supplement 1 to NUREG/CR-6850 and EPRI 1011989, EPRI and NRC, EPRI 1019259, December 2009.

[A-10] Electric Power Research Institute, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325. EPRI TR-1018243, October 2008.

[A-11] Clinton Power Station 2009 PRA Peer Review Report, April, 2010.

[A-12] U.S. Nuclear Regulatory Commission (NRC), Memorandum to Brian W. Sheron, Director Office of Nuclear Regulatory Research, From Patrick Hiland, Chairman, Safety/Risk Assessment Panel for Generic Issue 199: Safety/Risk Assessment Results for Generic Issue 199, Implications of Updated Probabilistic Seismic Hazard Estimates In Central and Eastern United States on Existing Plants, Accession Number ML11356A034, September 2, 2010.

[A-13] CPS 2011 Self-Assessment of the PRA Against the ASME PRA Standard Requirements Notebook, CPS-PSA-016, Revision 1, December 2011 A-18 C467150095-12660-12/1/15