ML14265A219

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

ML14265A219
Person / Time
Site:	Calvert Cliffs
Issue date:	09/18/2014
From:	George Gellrich Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML14265A219 (142)

Text

' N George Gellrich Exeon Generation Site Vice President Calvert Cliffs Nuclear Power Plant 1650 Calvert Cliffs Parkway Lusby, MD 20657 410 495 5200 Office 717 497 3463 Mobile www.exeloncorp.com george.gellrich@exeloncorp.com 10 CFR 50, Appendix J September 18, 2014 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Renewed Facility Operating License Nos. DPR-53 and DPR-69 NRC Docket Nos. 50-317 and 50-318

Subject:

License Amendment Request: Revise Technical Specification Section 5.5.16 for Permanent Extension of Type A and C Leak Rate Test Frequencies Pursuant to 10 CFR 50.90, Calvert Cliffs Nuclear Power Plant, LLC (Calvert Cliffs) requests an amendment to Renewed Operating Licenses No. DPR-53 and DPR-69 for Calvert Cliffs Units No. 1 and 2. The proposed amendment revises Calvert Cliffs Technical Specification 5.5.16, "Containment Leakage Rate Testing Program" to allow for permanent extensions of the Type A Integrated Leak Rate Testing and Type C Leak Rate Testing frequencies.

The proposed amendment and significant hazards discussion are provided in Attachment (1). The marked up page of the affected Technical Specification is provided in Attachment (2).

The proposed amendment is risk-informed and follows the guidance in Regulatory Guide 1.174, Revision 2. Calvert Cliffs performed a plant-specific evaluation to assess the risk impact of the proposed amendment. A copy of the risk assessment is provided in Attachment (3).

A list of commitments associated with this proposed amendment is provided in Attachment (4).

Calvert Cliffs requests approval of this proposed amendment by July 1, 2015 with an implementation period of 75 days. Approval by this time will allow Calvert Cliffs to avoid performing final preparations that would otherwise be necessary to conduct an integrated leakage rate test during the Unit 1 refueling outage that is scheduled to begin in March 2016.

Document Control Desk September 18, 2014 Page 2 Should you have questions regarding this matter, please contact Mr. Douglas E. Lauver at (410) 495-5219.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 18, 2014.

Respectfully, GHG/KLG/bjd George H. Gellrich Site Vice President Attachments: (1) Evaluation of the Proposed Change (2) Marked up Technical Specifications Page (3) Evaluation of Risk Significance of Permanent ILRT Extension (4) Regulatory Commitment cc: NRC Project Manager, Calvert Cliffs NRC Resident Inspector, Calvert Cliffs NRC Regional Administrator, Region I S. Gray, MD-DNR

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Calvert Cliffs Nuclear Power Plant September 18, 2014

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating License Numbers DPR-53 and DPR-69 for Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Units 1 and 2. The proposed change would revise the Operating Licenses by amending Technical Specification (TS)

Section 5.5.16, Containment Leakage Rate Testing Program. The proposed change to the Technical Specification contained herein would revise Calvert Cliffs TS 5.5.16, by replacing the reference to Regulatory Guide (RG) 1.163 (Reference 1) with a reference to Nuclear Energy Institute (NEI) Topical Report NEI 94-01 Revision 3-A (Reference 2) [Nuclear Regulatory Commission (NRC)-approved version specified in the 10 CFR Part 50, Appendix J Program Plan] as the implementation document used by Calvert Cliffs to implement the Units 1 and 2 performance-based leakage testing program in accordance with Option B of 10 CFR Part 50, Appendix J. The proposed change would also delete the listing of one time exceptions previously granted to Integrated Leak Rate Test (ILRT) test frequencies and exceptions from post modification ILRTs when Calvert Cliffs replaced Steam Generators. These exceptions are historical in that the one-time extensions of ILRT test frequencies have already been used and replacement of both Unit 1 and 2 Steam Generators have already been completed. Additional information on these exceptions is provided in Section 3.2.3 of this document.

The proposed change would allow an increase in the ILRT test interval from its current 10 year frequency to a maximum of 15 years and the extension of the containment isolation valves leakage test (Type C tests) from its current 60 month frequency to 75 months in accordance with NEI 94-01 Revision 3-A.

Calvert Cliffs has transitioned through three parent owners in its history, initially Baltimore Gas & Electric, followed by Constellation Energy Nuclear Group. Effective April 1, 2014, the parent owner of Calvert Cliffs is Exelon Generation Company, LLC (Exelon). As such, when referring to past technical matters involving Calvert Cliffs as discussed in this License Amendment Request, for simplification purposes Exelon will be referred to as the Owner.

2.0 DETAILED DESCRIPTION Calvert Cliffs TS 5.5.16, "Containment Leakage Rate Testing Program" currently states, in part:

"A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in RG 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, including errata, as modified by the following exceptions:

a. Nuclear Energy Institute (NEI) 94 1995, Section 9.2.3: The first Unit 1 Type A test performed after the June 15, 1992 Type A test shall be performed no later than June 14, 2007. The first Unit 2 Type A test performed after the May 2, 2001 Type A test shall be performed no later than May 1, 2016."

b. Unit 1 is excepted from post modification integrated leakage rate testing requirements associated with steam generator replacement.

c. Unit 2 is excepted from post modification integrated leakage rate testing requirements associated with steam generator replacement.

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE The proposed change to Calvert Cliffs TS 5.5.16, "Containment Leakage Rate Testing Program" will remove exceptions (a), (b), and (c) and replace the reference to RG 1.163 with a reference to NEI Topical Report NEI 94-01 Revision 3-A. The proposed change will revise TS 5.5.16 to state, in part:

"A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012."

A markup of TS 5.5.16 is provided in Attachment (2).

This proposed change is requested to extend the performance of the next Unit 1 ILRT from the 2016 refueling outage to a subsequent refueling outage (no later than May 3, 2021) when it can be performed in a refueling outage that involves fewer conflicts with other planned activities and without extending the refueling outage duration. This proposed amendment would also extend the performance of the next Unit 2 ILRT to be performed no later than March 17, 2028.

Attachment (3) contains the plant specific risk assessment conducted to support this proposed change. This risk assessment followed the guidelines of NRC RG 1.174 (Reference 3) and NRC RG 1.200, Revision 2 (Reference 4). The risk assessment concluded that the increase in risk as a result of this proposed change is small and is well within established guidelines.

3.0 TECHNICAL EVALUATION

3.1 Containment Structure Description The basic design criteria of the Containment Structure are that the structure shall have a low strain elastic response such that its behavior will be predictable under all design loadings and that the integrity of the liner plate be maintained under all loading conditions.

Each containment structure consists of a post-tensioned reinforced concrete cylinder and dome connected to and supported by a reinforced concrete foundation slab. The interior surface of the structure is lined with a Y1/4" thick welded steel plate to assure a high degree of leak tightness.

The containment structure has personnel and equipment access openings as well as numerous mechanical and electrical systems that penetrate the containment structure wall through welded steel penetrations. The penetrations and access openings were designed, fabricated, inspected, and installed in accordance with the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel (B&PV) Code,Section III, Class B.

The containment structure, in conjunction with Engineering Safeguards Features, is designed to withstand the internal pressure and coincident temperature resulting from the energy released in the event of the loss-of-coolant accident (LOCA) associated with rated full power operation. The current design conditions for the structure are an internal pressure of 50 psig, a coincident concrete surface temperature of 276 0F and a leak rate of 0.16% by weight per day at design temperature and pressure.

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 3.1.1 Containment Liner The containment liner is a 1/4" thick welded steel plate that is attached to the inside face of the containment structure dome, cylindrical wall, and foundation slab. It forms a leak-tight barrier against the release of radioactive material outside the containment structure. The 1/4"-thick liner plate is attached to the concrete by means of an angle grid system stitch welded to the liner plate and embedded in the concrete. The frequent anchoring is designed to prevent significant distortion of the liner plate during accident conditions and to ensure that the liner maintains its leak-tight integrity. The liner plate is protected from corrosion on the inside with 3 mils of inorganic zinc primer topped with 6 mils of an organic epoxy up to Elevation 75'0", and 3 mils of an inorganic topcoat above that elevation. There is no paint on the side that comes in contact with the concrete.

A finished concrete floor covers the portion of the liner on the containment foundation slab. A leak chase system allows the containment liner welds located under the concrete floor to be leak tested during the ILRT of the containment.

3.1.2 Electrical Penetrations Two types of electrical penetration assemblies are used; canister and unitized header. All electrical penetration assemblies were fabricated and tested in accordance with the ASME, B&PV Code,Section III, Nuclear Vessel Code. The canister type is inserted in a nozzle of suitable diameter integral with the containment structure and field welded on the inside end.

The unitized header type is welded to the nozzle on the outside end.

3.1.3 Piping Penetrations Single barrier piping penetrations are provided for all piping passing through the containment walls. The closure of the pipe to the liner plate is accomplished with a pipe cap welded to the pipe and to the liner plate reinforcement. In the case of piping that carries hot fluid, the pipe is insulated and cooling is provided to restrict the concrete maximum temperature to 150 0 F.

The anchorage of penetration closure connecting pipes to the containment wall were designed as Seismic Category I structures to resist all forces and moments caused by a postulated pipe rupture. The design conditions include the maximum pipe reactions and pipe rupture forces.

The penetration assembly, consisting of pipe cap and the assembly welds and welds to the liner plate, utilizes full penetration welds. The assembly is anchored into the wall concrete and designed to accommodate all forces and moments due to pipe rupture and thermal expansion.

3.1.4 Containment Penetration Bellows Assemblies Expansion bellows are not utilized in the design of the mechanical penetrations at Calvert Cliffs. There are bellows used on the fuel transfer tube penetration to accommodate relative movement between the refueling canal liner and the containment building penetration.

However, those bellows do not form part of the containment building vessel or pressure boundary. They are unaffected by this proposed amendment.

3.1.5 Refueling Tube Penetration A refueling tube penetration is provided for fuel movement between the refueling pool in the containment structure and the spent fuel pool in the Auxiliary Building. The penetration 3

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE consists of a 36" stainless steel pipe installed inside a 42" pipe sleeve. The inner pipe acts as the refueling tube and is fitted with a gate valve in the spent fuel pool and an encapsulating pipe sleeve, which is welded to the refueling pool liner and sealed off from the Containment with a testable double 0-ring blind flange in the refueling pool. This arrangement prevents leakage through the refueling tube in the event of a LOCA. The 42" pipe sleeve is welded to the containment liner.

Bellows expansion joints are provided on the transfer tube to compensate for any differential movement between the tube and the building structures. The bellows do not form any part of the containment boundary so they are unaffected by this proposed change.

3.1.6 Moisture Barrier A layer of compressible material covers both sides of the containment liner on the containment wall where the finished concrete floor joins the wall. This cork layer, covered with a waterproof seal, serves as an expansion joint to accommodate any relative movement between the containment wall, floor, and liner.

3.1.7 Containment Tendons There are four types of Containment tendons: Dome, Hoop, Vertical (Original -

Undisturbed), and Vertical (Replaced or Restressed). The total active population of tendons for Unit 1 is 871 tendons. Five additional locations are considered "abandoned" since the tendon wires were never installed during original construction and not inspected.

The total active population of tendons for Unit 2 is 876 tendons.

Each tendon consists of approximately 90 1/4" diameter wires with button-headed BBRV-type anchorages. The tendons are housed in spiral wrapped, corrugated, thin-wall, carbon steel sheathing.

After fabrication, each tendon was shop dipped in a petroleum corrosion protection material. After installation, the tendon sheathing was filled with corrosion preventive grease. The ends of all tendons were covered with pressure-tight, grease filled caps for corrosion protection. All the vertical tendons for each unit have received new corrosion preventive grease between 1997 and the end of 2002. In addition some original vertical tendons for each unit were re-stressed or replaced with new tendons between 2001 and 2002.

In the concept of a post-tensioned containment structure, the internal pressure load is balanced by the application of an opposing external force on the structure. Sufficient post-tensioning was used on the cylinder and dome to more than balance the internal pressure so that a margin of external pressure exists beyond that required to resist the design pressure. Nominal, bonded reinforcing steel was also provided to distribute strains due to shrinkage and temperature. Additional bonded reinforcing steel was used at penetrations and discontinuities to resist local moments and shears.

The internal pressure loads on the foundation slab are resisted by both the external bearing pressure due to dead load and the strength of the reinforced concrete slab.

Thus, post-tensioning was not required to exert an external pressure for this portion of the structure.

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 3.2 Justification for the Technical Specification Change 3.2.1 Chronology of Testing Requirements of 10 CFR Part 50, Appendix J The testing requirements of 10 CFR Part 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 Part 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 Part 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach.

Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR Part 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 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 "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 Part 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC Safety Evaluation Report (SER) on NEI 94-01. The NRC SER was included in the front matter of this NEI report. Nuclear Energy Institute 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 (September 1995). It delineates a performance-based 5

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 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, was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J and includes provisions for extending Type A ILRT intervals to up to 15 years. Nuclear Energy Institute 94-01 has been endorsed by RG 1.163 and NRC SERs of June 25, 2008 (Reference 9) and June 8, 2012 (Reference 10) as an acceptable methodology for complying with the provisions of Option B to 10 CFR Part 50. The regulatory positions stated in RG 1.163 as modified by NRC SERs of June 25, 2008 and June 8, 2012 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 of 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 licensee's allowable administrative limits. Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment airlocks) and up to a maximum of 75 months for Type C tests. If a licensee considers extended test intervals of greater than 60 months for Type B or Type C tested components, the review should include the additional considerations of as-found tests, schedule and review as described in NEI 94-01, Revision 3-A, Section 11.3.2.

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

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

NEI 94-01, Revision 3-A, Section 10.1 concerning the use of grace 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 6

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 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 SER Section 4.0:

"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."

3.2.2 Current Calvert Cliffs ILRT Requirements Title 10 CFR Part 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 March 13, 1996 the NRC-approved License Amendment No. 212 for Calvert Cliffs Unit 1 and Amendment 189 for Unit 2 authorizing the implementation of 10 CFR Part 50, Appendix J, Option B for Type A tests. On January 11, 1997 the NRC-approved License Amendment No. 219 for Calvert Cliffs Unit 1 and Amendment 196 for Unit 2 authorizing the implementation of 10 CFR Part 50, Appendix J, Option B for Type B and Type C tests. Current TS 5.5.16 requires that a program be established to comply with the containment leakage rate testing requirements of 10 CFR 50.54(o) and 10 CFR Part 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. Regulatory Guide 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.

Regulatory Guide 1.163, Section C. 1 states that licensees intending to comply with 10 CFR Part 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 American National Standards Institute (ANSI)/American Nuclear Society (ANS) 56.8-1994. Nuclear Energy Institute 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 1.0 La (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 Calvert Cliffs' containment structures. NUREG-1493 concluded in Section 10.1.2 that 7

ATTACHMENT (1)

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

3.2.3 Calvert Cliffs 10 CFR Part 50, Appendix J Option B Licensing History March 13, 1996 The Commission issued on March 13, 1996 Amendment No. 212 to Facility Operating License No. DPR-53 and Amendment No. 189 to Facility Operating License No. DPR-69 for Calvert Cliffs Units 1 and 2, respectively (Reference 11). The amendment revised TSs to reflect the approval for the use of 10 CFR Part 50, Appendix J, Option B, for Calvert Cliffs Units 1 and 2, containment leakage rate test program for Type A tests only.

February 11, 1997 The Commission issued on February 11, 1997 Amendment No. 219 to Facility Operating License No. DPR-53 and Amendment No. 196 to Facility Operating License No. DPR-69 for Calvert Cliffs Units 1 and 2 (Reference 12).

The amendments adopted Option B of 10 CFR Part 50, Appendix J, approving Type B and Type C containment leakage testing to be performed on a performance-based testing schedule.

May 1, 2002 The Commission issued on May 1, 2002 Amendment No. 252 to Renewed Facility Operating License No. DPR-53 for the Calvert Cliffs Unit 1 (Reference 13).

The amendment allowed a one-time five-year extension, for a total of 15 years, for the performance of the next Unit 1 ILRT following the June 15, 1992 Type A test. This test was to be performed no later than June 14, 2007 (ILRT was conducted on May 3, 2003). The amendment also exempted Calvert Cliffs Unit 1 from the requirement to perform a post-modification containment ILRT associated with steam generator replacement. The Calvert Cliffs Unit 1 steam generators were replaced during the 2002 refueling outage.

June 27, 2002 The Commission issued on June 27, 2002 Amendment No. 230 to Renewed Facility Operating License No. DPR-69 for Calvert Cliffs Unit 2 (Reference 14).

The amendment revised TS 5.5.16 to eliminate the requirement to perform post-modification containment integrated leakage rate testing following replacement of the Unit 2 steam generators. The Calvert Cliffs Unit 2 steam generators were replaced during the 2003 refueling outage.

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE August 29, 2007 The Commission issued on August 29, 2007 Amendment No. 281 to Renewed Facility Operating License No. DPR-53 and Amendment No. 258 to Renewed Facility Operating License No. DPR-69 for Calvert Cliffs Units 1 and 2 (Reference 15).

The amendments revised the accident source term in the design basis radiological consequence analyses. This resulted in a change in Calvert Cliffs design basis containment leak rate, La, from a value of 0.2% of containment air weight per day (% per day), at containment peak pressure, to a value of 0.16% per day as expressed in TS 5.5.16.

March 22, 2011 The Commission issued on March 22, 2011 Amendment No. 274 to Renewed Facility Operating License No. DPR-69 for the Calvert Cliffs Unit 2 (Reference 16). The amendment revised TS 5.5.16, "Containment Leakage Rate Testing Program," to allow a one-time five-year extension for Unit 2s ILRT interval from 10 to 15 years. This would require the licensee to perform its next ILRT no later than May 1, 2016 (ILRT was performed on March 17, 2013).

July 31, 2013 The Commission issued on July 31, 2013 Amendment No. 303 to Renewed Facility Operating License No. DPR-53 and Amendment No. 281 to Renewed Facility Operating License No.

DPR-69 for Calvert Cliffs Units 1 and 2 (Reference 17).

The amendments revised TS 5.5.16 by increasing the peak calculated containment internal pressure (Pa) from 49.4 pounds per square inch gauge (psig) to 49.7 psig for the design basis LOCA. In support of the revised Pa, the amendment also revised TS 3.6.4 by decreasing the upper bound internal containment pressure limit from 1.8 psig to 1.0 psig.

3.2.4 Integrated Leakage Rate Testing History (ILRT)

As noted previously, Calvert Cliffs TS 5.5.16 currently requires Type A, B, and C testing in accordance with RG 1.163, which endorses the methodology for complying with Option B. The performance leakage rates are calculated in accordance with NEI 94-01, Section 9.1.1 for Type A testing. The performance leakage rate includes the Type A Upper Confidence Limit at 95% plus the as-left minimum pathway leakage rate for all Types B and C pathways not in service, isolated, or not lined up in their test position. Tables 3.2-1 and 3.2-2 list the Type A ILRT past results for Units 1 and 2, respectively.

Table 3.2-1, Unit I Type A ILRT History Leakage Rate (1) As Left Type C Minimum (Containment air Path Contribution Test Date (weight %/day) (weight %/day) 12/01/1973 0.0466 0.00 03/06/1978 0.136 0.028 06/22/1982 0.0514 0.0254 05/20/1985 0.032 0.002 05/27/1988 0.036 0.006 The results of the last two Type A ILRTs for Calvert Cliffs Unit 1:

07/05/1992 0.1564(3) 0.07421 05/03/2006 0.09515 0.00168 9

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE Table 3.2-2, Unit 2 Type A ILRT History Leakage Rate (1) As Left Type C Minimum (Containment air Path Contribution Test Date (weight %/day) (weight %/day) 03/14/1976 0.019 0.0001 11/15/1979 0.052 0.00084 12/22/1982 0.025 0.0015 11/24/1985 0.142(2) 0.081 01/16/1991 0.061 0.001 The results of the last two Type A ILRTs for Calvert Cliffs Unit 2:

05/02/2001 0.0738 0.001 03/17/2013 0.0802 0.0025 (1) On August 29, 2007 Calvert Cliffs design basis containment leak rate, La, was changed from a value of 0.2 wt%/day at containment peak pressure, to a value of 0.16 wt%/day as expressed in TS 5.5.16.

(2) Local Leakage Rate Test, repair, and adjustments of containment isolation valves was performed prior to the 11/24/1985 ILRT. The minimum pathway leakage improvement due to repairs and adjustments was 140,591.24 sccm or 0.081 wt%/day. The as found containment leakage rates, the sum of the minimum pathway leakage improvement and ILRT upper 95% confidence level of 0.061 wt%/day satisfied the as found acceptance criterion that the sum must be less than La = 0.2 wt%/day.

(3) Repairs and adjustments were made to various penetrations during the outage associated with the 07/05/1992 ILRT. These repairs resulted in an improvement to the overall performance of the containment totaling 123,236 sccm, or 0.07421 wt%/day. The adjusted ILRT leakage rate was determined by adding the minimum pathway leakage improvements to the "as-left" test results of 0.0824 wt%/day equaling 0.1564 wt%/day, which is below the technical specification's maximum allowable limit of 0.2 wt%/day.

The results of the last two Type A ILRTs for both Calvert Cliffs Units 1 and 2 are less than the current maximum allowable containment leakage rate of 0.16 wt%/day at the test pressure of 50 psig. As a result, since both tests for both units were successful, the current ILRT interval frequency for Calvert Cliffs Units I and 2 are ten years.

3.3 Plant Specific Confirmatory Analysis 3.3.1 Methodology An evaluation has been performed to assess the risk impact of extending the Calvert Cliffs Units 1 and 2 ILRT intervals from the current 10 years to 15 years. The purpose of this analysis is to provide a risk assessment of permanently extending the currently allowed Containment Type A ILRT out to 15 years. The risk assessment follows the guidelines from:

" NEI 94-01, Revision 3-A, the methodology used in EPRI TR-104285,

" NEI "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals" from November 2001, 10

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE

" NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in RG 1.200 as applied to ILRT interval extensions,

• Risk insights in support of a request for a plant's licensing basis as outlined in RG 1.174,

• Methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion induced leakage of steel liners going undetected during the extended test interval,

" Methodology used in EPRI Report No. 1009325, Revision 2-A (Reference 18), the methodology improvements in EPRI Report No. 1018243.

In the SER issued by NRC letter dated June 25, 2008, the NRC concluded that the methodology in EPRI TR-1 009325, Revision 2, is 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 3.3-1 addresses each of the four limitations and conditions for the use of EPRI 1009325, Revision 2.

Table 3.3-1, EPRI Report No. 1009325 Revision 2 Limitations and Conditions LimitationlCondition (From Section 4.2 of SE) Calvert Cliffs Response

1. The licensee submits documentation Calvert Cliffs PRA quality is addressed in indicating that the technical adequacy of Section 3.3.2 and Attachment (3), "Calvert their PRA is consistent with the Cliffs Nuclear Power Plant: Evaluation of requirements of RG 1.200 relevant to the Risk Significance of Permanent ILRT ILRT extension Extension" Attachment 1, "PRA Quality Statement for Permanent 15-Year ILRT Extension"

2. The licensee submits documentation EPRI Report No. 1009325, Revision 2-A, indicating that the estimated risk increase incorporates these population dose and associated with permanently extending the CCFP acceptance guidelines, and these ILRT surveillance interval to 15 years is guidelines have been used for the Calvert small, and consistent with the clarification Cliffs plant specific assessments.

provided in Section 3.2.4.5 of this SE.

Specifically, a small increase in population The increase in population dose is dose should be defined as an increase in 0.20 person-rem/year for Unit 1 and population dose of less than or equal to 0.11 person-rem/year for Unit 2.

either 1.0 person-rem per year or 1 % of the total population dose, whichever is restrictive.

In addition, a small increase in CCFP The increase in CCFP is 0.558% for Unit 1 should be defined as a value marginally and 0.490% for Unit 2.

greater than that accepted in a previous one-time ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point.

11

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE Limitation/Condition (From Section 4.2 of SE) Calvert Cliffs Response

3. The methodology in EPRI Report No. EPRI Report No. 1009325, Revision 2-A, 1009325, Revision 2, is acceptable except incorporated the use of 100 La as the average for the calculation of the increase in leak rate for the pre-existing containment expected population dose (per year of large leakage rate accident case (accident reactor operation). In order to make the case 3b), and this value has been used in the methodology acceptable, the average leak Calvert Cliffs plant specific risk assessment.

rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

4. A licensee amendment request (LAR) is For Calvert Cliffs, containment over-pressure required in instances where containment is NOT relied upon for emergency core over-pressure is relied upon for cooling system (ECCS) performance.

emergency core cooling system (ECCS) performance 3.3.2 PRA Quality Statement for Permanent 15-Year ILRT Extension The Calvert Cliffs Internal Events and Wind Model, Calvert-CAFTA-TREE-6.2a, was used for this analysis. An independent PRA peer review was conducted under the auspices of the Pressurized Water Reactor Owners Group in June of 2010, and was performed against the guidance of RG 1.200, Revision 2, and requirements of ASME/ANS RA-Sa-2009. The scope of the review was a full-scope review of the Calvert Cliffs Nuclear Plant (Calvert Cliffs) at-power, internal initiator PRA.

Findings (generally, documentation issues or model concerns that have been evaluated as not significant using a sensitivity study) have been captured in the PRA Configuration Risk Management Program (CRMP) database. On an on-going basis, other potential PRA model and documentation changes are captured and prioritized in the CRMP database.

The Calvert Cliffs Internal Events model was also updated to support the Calvert Cliffs Fire PRA.

The Calvert Cliffs Internal Events model was peer reviewed in June 2010. All findings, which had significant impact on this analysis, have been addressed. This assessment is provided in Attachment (3) as Table 1. The ILRT application was determined to be an application requiring a Capability Category II PRA model per the RG 1.200 criteria, Revision 2. This is based on the requirement for numerical results for CDF and LERF to determine the risk impact of the requested change and the fact that this change is risk-informed, not risk-based. Table 1 includes discussion of all findings from the industry-peer review along with the assessment and evaluation of the finding that shows that they have either been addressed or have no material impact on the ILRT interval extension request.

The peer review found that 97% of the supporting requirements (SRs) evaluated Met Capability Category II or better. There were 3 SRs that were noted as "not met" and eight that were noted as Category I. As noted in the peer review report, the majority of the findings were documentation related. Of the 11 SRs, which did not meet Category II or better, seven were related to conservatisms or documentation in LERF and two were related to internal floods.

12

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE There were 39 findings. All findings, which could be relevant to the ILRT extension evaluation, were updated in the internal events model used to quantify the Level 2 release states. Thus, with the exception of minor documentation concerns, the internal events model meets Capability Category II or causes conservative results for all SRs relevant to the ILRT extension evaluation results. No significant changes have been implemented in the internal events PRA. As there are no new methods applied, no follow on or focused peer reviews were required.

The Calvert Cliffs Fire PRA peer review was performed January 16-20, 2012 using the NEI 07-12 Fire PRA peer review process, the ASME PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Rev. 2. The purpose of this review was to establish the technical adequacy of the Fire PRA for the spectrum of potential risk-informed plant licensing applications for which the Fire PRA may be used. The 2012 Calvert Fire PRA peer review was a full-scope review of all of the technical elements of the Calvert Cliffs at-power FPRA (2012 model of record) against all technical elements in Section 4 of the ASME/ANS Combined PRA Standard, including the referenced internal events SRs. The peer review noted a number of facts and observations (F&Os). The findings and their dispositions are provided in Attachment (3) as Table 2. All findings are being provided and have been dispositioned. All F&Os that were defined as suggestions have been dispositioned and will be available for NRC review. The Fire PRA is adequate to support the ILRT extension.

The Calvert Cliffs seismic PRA model is relatively conservative and, other than the high magnitude acceleration event, is not a dominant contributor. The Calvert Cliffs high winds PRA model is very conservative in the tornado area in that all tornados are grouped into the most conservative event. PRA risk for tornadoes and high winds are based upon IPEEE values.

Calvert Cliffs has maintained and updated a high wind PRA model in order to perform risk assessment of tornado missile impacts and hurricane force winds. Although this model has not been peer reviewed in compliance with the ASME/ANS RA-Sa-2009 standard, the model is based upon accepted methodology and utilizes the ASME/ANS RA-Sa-2009 compliant internal events model. High winds updates are not expected to cause a significant increase in CDF or LERF. A more detailed assessment would be expected to cause a decrease in CDF.

3.3.3 Summary of Plant-Specific Risk Assessment Results The findings of the Calvert Cliffs Unit 1 and 2 Risk Assessment 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 small. The Calvert Cliffs plant-specific results for extending ILRT interval from the current 10 years to 15 years are summarized below:

Based on the results from Attachment (3), Sections 5.2, "Analysis" and 5.3, "Sensitivities" the following conclusions regarding the assessment of the plant risk associated with extending the Type A ILRT test frequency to 15 years are as follows:

RG 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.OE-06/year and increases in LERF less than 1.OE-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 4.54E-8/year for Unit 1 and 2.46E-8/year for Unit 2 using the EPRI guidance. As such, the estimated change in LERF is determined to be "very small" for both units using the acceptance guidelines of RG 1.174.

13

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE

" The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.20 person-rem/year for Unit 1 and 0.11 person-rem/year for Unit 2.

EPRI Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of

• 1.0 person-rem per year, or < 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals.

The results of this calculation meet these criteria for both units. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

" The increase in the conditional containment failure (CCFP) from the 3 in 10 year interval to 1 in 15 year interval is 0.558% for Unit 1 and 0.490% for Unit 2. EPRI Report No. 1009325, Revision 2-A states that increases in CCFP of < 1.5% is very small. Therefore, this increase is judged to be very small.

Calvert Cliffs is transitioning to NFPA 805 licensing basis for fire protection and submitted a License Amendment Request (LAR) on September 24, 2013 (Reference 19). This transition included performing a Fire PRA and committing to modifications to reduce the fire-induced core damage and large early release frequencies to those reported in the NFPA 805 LAR.

Compensatory actions have been implemented to reduce the fire risk until the modifications are implemented. The Unit 1 ILRT is scheduled for 2016, which is prior to the scheduled implementation of all the modifications by 2018. It is anticipated that many, but not all, of the NFPA 805 modifications will be completed by the end of Unit 1 2016 refueling outage. Risk mitigation strategies will be in place for any open modifications. These strategies may be actions to reduce fire initiating event probabilities, actions to improve suppression probability, and/or actions to recover or protect systems that mitigate core damage and large early release accident sequences. The next Unit 2 ILRT is scheduled for 2023, so the NFPA 805 modifications will be implemented prior to the extension.

• An assessment of the impact of external events was performed using fire risk analysis from the Fire PRA. The total LERF value for Unit 1 is 6.OOE-6/yr for Unit 1 and 7.38E-6/yr for Unit 2. Since the total LERF for both units is less than 1.OE-5, it is acceptable for the ALERF to be between 1.OE-7 and 1.OE-6.

" An assessment of the impact of external events was also performed using fire risk analysis from the Individual Plant Examination of External Events (IPEEE) rather than the Fire PRA model. The total LERF value for Unit 1 is 8.24E-6/yr for Unit 1 and 3.90E-6/yr for Unit 2.

Since the total LERF for both units is less than 1.OE-5, it is acceptable for the ALERF to be between 1.OE-7 and 1.OE-6.

Therefore, increasing the ILRT interval to 15 years is considered to be insignificant since it represents a very small change to the Calvert Cliffs Unit 1 and 2, risk profile.

3.3.4 Previous Assessments The NRC in NUREG-1493 has previously concluded that:

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

14

ATTACHMENT (1)

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

The findings for Calvert Cliffs confirm these general findings on a plant-specific basis considering the severe accidents evaluated for Calvert Cliffs, the Calvert Cliffs containment failure modes, and the local population surrounding Calvert Cliffs.

Details of the Calvert Cliffs Unit 1 and 2 risk assessment are contained in Attachment (3) of this submittal.

3.4 Non-Risk Based Assessment Consistent with the defense-in-depth philosophy discussed in RG 1.174, Calvert Cliffs has assessed other non-risk based considerations relevant to the proposed amendment. Calvert Cliffs 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.

3.4.1 Safety-Related and Controlled Protective Coatings Inspection Program The requirements of 10 CFR Part 50, Appendix B are implemented through specification of appropriate technical and quality requirements for the Service Level 1 coatings program which includes ongoing maintenance activities. Calvert Cliffs has implemented controls for the procurement, application, and maintenance of Service Level I protective coatings used inside the Containment in a manner that is consistent with the licensing basis and regulatory requirement- applicable to Calvert Cliffs.

Calvert Cliffs conducts condition assessments of Service Level I coatings inside Containment as part of the safety-related and controlled protective coatings program. Inspections of coatings systems are scheduled every outage on a pre-established basis to verify containment liner coating thickness and condition.

This program also satisfies the License Renewal Application commitment contained in Table 16-2 of Calvert Cliffs Updated Final Safety Analysis Report (UFSAR), as managing general corrosion of the containment wall and dome liner plates.

3.4.2 Containment Inservice Inspection Program The purpose of the Calvert Cliffs Containment Inservice Inspection (CISI) program is to periodically perform destructive and nondestructive examination of ASME Class MC and CC components in order to identify the presence of any service-related degradation.

The CISI program is established in accordance with 10 CFR 50.55a. This program has been developed to comply with ASME Section Xl 2004 Edition, except where specific written alternatives from Code requirements have been requested by Calvert Cliffs and granted by the NRC and implements the requirements of the following:

• UFSAR 5.1, Containment Structure 15

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE

" UFSAR 5.5.2.2, Surveillance of Structural Integrity

" TS 5.5.6, Concrete Containment Tendon Surveillance Program

" TS 5.5.16, Containment Leakage Rate Testing Program

" Technical Requirements Manual, Section 15.6.1, Containment Structural Integrity The program defines the Class MC and CC components and the Code-required examinations for each ASME Section XI examination category, and the augmented inspection scope, as applicable.

The components subject to the requirements of this CISI program are those which make up the containment structure, its leak tight barrier (including integral attachments) and those which contribute to its structural integrity, specifically, Class MC pressure-retaining components, and their integral attachments and Class CC post tensioned concrete containments.

The administrative procedures and inspection schedule described in the CISI program, combined with applicable Calvert Cliffs and approved vendor procedures, constitute the CISI portion of the Calvert Cliffs Ten-year Inservice Inspection (ISI) program. The Second Interval CISI Program Plan dated September 2009 is currently in effect as of the date of this amendment request. The Third Interval of the CISI Program Plan has not been developed at this time so all dates associated with the Third Interval are postulated.

IWE(Class MC) Inspection Interval and Periods The second ten-year CISI interval for both Units for the performance of containment ISI (IWE) complies with IWE-2412 Inspection Program B and began on September 9, 2009 and will end on September 9, 2018. This interval was shortened as a result of extending the first ten-year CISI interval by one year. Each interval is then further divided into three periods.

IWL (Class CC) Inspection Periods (Concrete)

The second ten-year containment interval for the performance of containment ISI (IWL) for both Units complies with IWL-2400 and is effective for IWL inspections conducted between September 9, 2009 and September 9, 2018.

Concrete examinations shall be conducted every five years (+/- one year) as described in IWL 2410(a) and (c). For the purposes of the containment ISI program, an IWL inspection period is five years, with two periods per inspection interval.

Concrete surface areas affected by a repair/replacement activity shall be examined at one year

(+/- three months) following completion of repair/replacement activity. If plant operating conditions are such that examination of portions of the concrete cannot be completed within this time interval, examination of those portions may be deferred until the next regularly-scheduled plant outage.

This Second Ten-Year Interval for the performance of Containment ISI (IWL) for both Units complies with IWL-2400 and is effective for IWL inspections conducted between September 9, 2009 and September 9, 2018. The Third Ten-Year Interval will be effective between September 9, 2018 and September 9, 2028.

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE IWL (Class CC) Inspection Periods (Tendons)

For multiple-unit plant sites, such as Calvert Cliffs, the tendon examination frequency may be extended to ten years per unit provided the containment structures utilize the same pre-stressing system, are essentially identical in design, had their original structural integrity test performed within two years of one another, and experience similar environmental exposure.

The examinations required by IWL-2500 for unbonded post-tensioning systems can then alternate between the two units every five years, as allowed by IWL-2421 (sites with multiple units).

For Calvert Cliffs Unit 1, all examinations required by IWL-2500 (Items L2.10 thru L2.50) shall be performed at 1, 3, and 10 years and every 10 years thereafter (20, 30, 40, 50, 60 years).

Only the visual examination of Tendon Anchorage Area and analysis of the Corrosion Protection Medium need be performed at 5 and 15 years and every 10 years thereafter (25, 35, 45, 55 years). The 35-year examination was performed in 2012.

For Calvert Cliffs Unit 2, all examinations required by IWL-2500 (Items L2.10 thru L2.50) shall be performed at 1, 5, and 15 years and every 10 years thereafter (25, 35, 45, 55 years). Only the visual examinations Tendon Anchorage Area and analysis of the Corrosion Protection Medium need be performed at 3 and 10 years and every 10 years thereafter (20, 30, 40, 50, 60 years). The 35-year examination was performed in 2013.

Tables 3.4-1 thru 3.4-3 below describe the second-ten year ISI IWE/IWL interval and the subsequent interval for both units and encompasses the first extended interval ILRT testing:

Table 3.4-1, Calvert Cliffs IWL (Concrete) Examination Periods and Schedule Unit 1 and 2 Period Specified Date Tolerance 10 year 8/14/2011 +/-1 year 15 year 8/14/2016 +/- 1 year 20 year 8/14/2021 +/- 1 year 25 year 8/14/2026 +/- 1 year 30 year 8/14/2031 +/- 1 year Table 3.4-2, Calvert Cliffs IWL (Tendons) Examination Periods and Schedule Unit 1 Unit 2 Period Specified Date Tolerance Period Specified Date Tolerance 35 Year 9/9/2011 +/- 1 Year 35 Year 9/9/2012 +/- 1 Year 40 Year 9/9/2016 +/- 1 Year 40 Year 9/9/2017 +/- 1 Year 45 Year 9/9/2021 +/- 1 Year 45 Year 9/9/2022 +/- 1 Year 50 Year 9/9/2026 +/- 1 Year 50 Year 9/9/2027 +/- 1 Year 55 Year 9/9/2031 +/- 1 Year 55 Year 9/9/2032 +/- 1 Year 17

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE Table 3.4-3, Calvert Cliffs Unit I and 2 IWEIIWL Interval, Periods, and Outages Unit 1 Unit 2 Period Period Refuel Period Period Refuel Inspection Start End Refuel Outage Start End Refuel Outage Periods Dates Dates Outage Year Dates Dates Outage Year 1 July 1, June 30, RFO-20 2010 July 1, June 30, RFO-19 2011 2009 2012 RFO-21 2012 2009 2012 2 July 1, June 30, RFO-22 2014 July 1, June 30, RFO-20 2013 2012 2016 RFO-23 2016 2012 2016 RFO-21 2015 3 July 1, June 30, RFO-24 2018 July 1, June 30, RFO-22 2017 2016 2019 2016 2019 RFO-23 2019 1 July 1, June 30, RFO-25 2020 July 1, June 30, RFO-24 2021 2019 2022 RFO-26 2022 2019 2022 2 July 1, June 30, RFO-27 2024 July 1, June 30, RFO-25 2023 2022 2026 RFO-28 2026 2022 2026 RFO-26 2025 3 July 1, June 30, RFO-29 2028 July 1, June 30, RFO-27 2027 2026 2029 2026 2029 RFO-28 2029 Adoption of Code Cases All Code Cases adopted for ASME Section Xl activities for use during the second ten-year containment ISI interval are listed below. The use of Code Cases is in accordance with ASME Section Xl, IWA-2440, 10 CFR 50.55a, and RG 1.147 (Reference 20). As permitted by ASME Section Xl and RG 1.14 7 or 10 CFR 50.55a, ASME Section Xl Code Cases may be adopted and used as described below:

Code Cases Adopted from RG 1.147 N-532-4 Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission N-624 Successive Inspections N-686 Alternative Requirements for Visual Examinations, VT-1, VT-2, and VT-3 N-739 Alternative Qualification Requirements for Personnel Performing Class CC Concrete and Post-Tensioning System Visual Examinations. This fulfills NRC concerns stated in 10 CFR 50.55a(b)(2)(ix)(F) regarding "owner-defined" personnel qualifications Relief Requests Table 3.4-4 contains an index of Requests for Alternatives and Requests for Relief written in accordance with 10 CFR 50.55a(a)(3) and (g)(5). The applicable submittal and NRC SER correspondence numbers are also included in Table 3.4-4 for each request for alternative and request for relief. Note that only Requests for Alternatives or Requests for Relief applicable to the requirements for Class MC and CC components are addressed in Table 3.4-4.

Table 3.4-4, Second Ten-year CISI Interval Relief Requests Relief Code Case Relief Request Licensee NRC SER Request Number Description Correspondence Correspondence ISI-04-02 N-753 Vision Tests ML090020097 ML093220090 Relief Request ISI-04-02, "Alternative Requirements to the Visual Acuity Demonstration Requirements of IWA-2321 (a)," proposes to use the ASME B&PV Code (ASME Code) Case 18

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE N-753, "Vision Tests" in lieu of the annual visual acuity requirements for Calvert Cliffs.

Specifically, the licensee is requesting the use of Code Case N-753 in lieu of the requirements of the 2004 Edition of the ASME Code,Section XI, paragraph IWA-2321 (a), "Visual Tests," for the near-distance acuity testing requirements.

Code Case N-753 provides an alternative to the visual acuity demonstration requirements of IWA-2321 (a) that will allow the testing to be administered and documented by an Optometrist, Ophthalmologist, or other health care professional who administers vision tests.

The NRC staff has reviewed Relief Request ISI-04-02 and concluded that the licensee's proposed alternative to use ASME Code Case N-753 in lieu of ASME Code, Section Xl, paragraph IWA-2321 (a) will provide an acceptable level of safety and quality. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), the proposed alternative, Code Case N-753, is authorized for the fourth 10-year ISI interval at Calvert Cliffs or until Code Case N-753 is approved for general use by reference in RG 1.147, "Inservice Inspection Code Case Acceptability." After that time, if the licensee wishes to continue to use Code Case N-753, the licensee must follow all conditions and limitations place on the use of the code case, if any, that are specified in RG 1.147.

Component Exemptions, IWE and IWL The basis for the selection of components at Calvert Cliffs which are determined to be within the scope of the required examinations was done in accordance with the requirements of IWE-1 220 and IWL-1220 respectively.

Calvert Cliffs does have areas that are considered inaccessible which are therefore exempt from inspection and are described below:

" IWE - The containment liner covering the containment foundation slab is inaccessible. This area is covered with the finished concrete floor and moisture barrier and accounts for approximately 15% of the containment liner surface area.

" IWL - Portions of the concrete surface that are covered by the liner, foundation material or backfill, or are otherwise obstructed by adjacent structures, components, parts or appurtenances are inaccessible. The entire inside concrete surface of the Calvert Cliffs containment buildings area covered in steel, which makes them inaccessible for examination.

Examination Methods & Personnel Qualifications The examination methods used to perform Code examinations for the nonexempt Class MC and CC components are in accordance with 10 CFR 50.55a requirements and the applicable ASME Codes.

Personnel performing IWE examinations shall be qualified in accordance with Exelon's written practice, or approved vendor written practice for certification and qualification of nondestructive examination personnel.

Personnel performing IWL examinations shall be qualified in accordance with written procedures prepared as required by IWL-2300, as modified by applicable Code Cases.

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE Inaccessible Areas For Class MC applications, Calvert Cliffs 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, Calvert Cliffs shall provide the following in the Owners Activity Report-I, 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.

Calvert Cliffs has not needed to implement any new technologies to perform inspections of any inaccessible areas at this time. However, Exelon 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 applicability to Calvert Cliffs. 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.

Containment Surfaces Requiring Augmented Examination Examination Category E-C of the ASME Code Section XI, 2004 Edition requires examination of Class MC "Metallic Containment" pressure-retaining components and their integral attachments, as well as, the metallic shell and penetration liners of Class CC "Concrete Containment" pressure-retaining components and their integral attachments that are likely to experience accelerated degradation and aging. Such components would include areas subject to accelerated corrosion or pitting, excessive wear, or other degradation mechanisms.

Calvert Cliffs examines 100% of the augmented areas on both units' Containment Structures each inspection period. The Code Item, Summary Number and Description related to the augmented components are identified in Table 3.4-5.

Table 3.4-5, Augmented Scope - Containment Interior Visible Surfaces Unit 1 Unit 2 Item Summary Item Summary No. No. Description No. No. Description E4.11 A27601 Liner bulging - liner 176 E4.11 B118221 Liner indentation - liner 221 E4.11 A28901 Liner bulging - liner 189 E4.11 B130254 Bulging - liner 254 E4.11 A29001 Liner bulging - liner 190 E4.11 B45153 Bulging - liner 153 E4.11 A29201 Liner bulging - liner 192 E4.11 B45154 Bulging - liner 154 E4.11 A30501 Liner bulging - liner 205 E4.11 B50000 Moisture barrier E4.12 A27602 Liner bulging - liner 176 E4.11 B69179 Liner bulging - liner 179 E4.12 A28901 Liner bulging - liner 189 E4.11 B69187 Liner bulging - liner 187 E4.12 A29002 Liner bulging - liner 190 E4.11 B69190 Liner bulging - liner 190 E4.12 A29202 Liner bulging - liner 192 E4.11 B69195 Liner bulging - liner 195 E4.12 A30502 Liner bulging - liner 205 E4.11 B69198 Liner bulging - liner 198 Examination of each item number is required each period until the areas remain essentially unchanged for one period.

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ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 3.4.3 Supplemental Inspection Requirements With the implementation of the proposed change, TS 5.5.16 will be revised by replacing the reference to RG 1.163 with reference to NEI 94-01, Revision 3-A. 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 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" In addition to the inspections performed in accordance with the CISI Program, Procedures Surveillance Test Procedure (STP)-M-665-1 and STP-M-665-2 "Containment Visual Inspection" are utilized to perform visual inspection of the normally accessible internal and exterior surfaces of the primary containment to identify evidence of structural deterioration, which could affect either structural integrity or leak tightness. The performance of STP-M-665-1 and STP-M-665-2 satisfy TS Surveillance Requirement 3.6.1.1 and TS 5.5.16 for the visual inspection of the interior and exterior surfaces of the containments. Personnel performing STP-M-665-1 and STP-M-665-2 are qualified as an ISI Engineer or certified as an Nondestructive Examination Visual Level II Examiner.

STP-M-665-1 and STP-M-665-2 are surveillance tests and are scheduled for performance as follows:

• Containment Liner - scheduled for inspection during each refueling outage in accordance with License Renewal Commitment - LRA Section 3.3.A, AMBD-0053 Rev 0001 and prior to each Type A test.

" Containment Concrete - scheduled for inspection every 36 +/- 14 months and prior to every Type A test.

Performance of these tests are also listed in Table 16-2 of Calvert Cliffs UFSAR as part of our License Renewal activities and are credited as managing general corrosion of the containment wall and dome liner plates. The current scheduling of STP-M-665-1 and STP-M-665-2 will also satisfy the inspection requirements of NEI 94-01 Revision 3-A.

3.4.4 Containment Leakage Rate Testing Program - Type B and Type C Testing Program Calvert Cliffs Types B and C testing program requires testing of electrical penetrations, airlocks, hatches, flanges, and containment isolation valves in accordance with 10 CFR Part 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. Per TS 5.5.16, the allowable maximum pathway total Types B and C leakage is 0.6 La where La equals approximately 276,800 sccm.

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EVALUATION OF THE PROPOSED CHANGE As discussed in NUREG-1493, 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 the Spring of 2008 through the Spring of 2014 for Calvert Cliffs Unit 1 and from the Spring of 2003 through the Spring of 2013 has shown an exceptional amount of 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 Calvert Cliffs Unit 1 shows an average of 6.1% of 0.6 La with a high of 8.5% or 0.051 La.

" The As-Left maximum pathway leak rate average for Calvert Cliffs Unit 1 shows an average of 6.9% of 0.6 La with a high of 9.9% or 0.060 La.

" The As-Found minimum pathway leak rate average for Calvert Cliffs Unit 2 shows an average of 8.2% of 0.6 La with a high of 12.4% or 0.074 La.

" The As-Left maximum pathway leak rate average for Calvert Cliffs Unit 2 shows an average of 7.2% of 0.6 La with a high of 9.4% or 0.057 La.

Tables 3.4-6 and 3.477 provide local leak rate test (LLRT) data trend summaries for Calvert Cliffs since the performance of the Unit 1 2006 ILRT and the Unit 2 2001 ILRT.

This summary shows that there has been no As-Found failure that resulted in exceeding the TS 5.5.16 limit of 0.6 La (166,080 sccm) and demonstrates a history of successful tests. The As-Found (AF) minimum pathway summations represent the high quality of maintenance of Type B and Type C tested components while the As-Left (AL) maximum pathway summations represent the effective management of the Containment Leakage Rate Testing Program by the program owner.

Table 3.4-6, Unit 1 Type B and C LLRT Combined As-Found/As-Left Trend Summary RFO 2008 2010 2012 2014 AF Min Path (sccm) 11546.9 9949.8 4862.99(1) 14197.19 Fraction of La 0.042 0.036 0.017 0.051 AL Max Path (sccm) 9677.8 5737.49 17506.39 13234.94 Fraction of La 0.035 0.021 0.063 0.048 AL Min Path (sccm) 8003.7 4358.99 16510.19(1) 11885.74 Fraction of La 0.029 0.016 0.060 0.043 Table 3.4-7, Unit 2 Type B and C LLRT Combined As-Found/As-Left Trend Summary RFO 2003 2005 2007 2009 2011 2013 AF Mi Path (sccm) 10535.95(2) 14380.4 10689.8 15772.6 9660.7 20638.99 Fraction of La 0.030 0.042 0.031 0.057 0.035 0.074 AL Max Path (sccm) 12347.1 3848.9 13936.6 15668.4 11723.7 13810.99 Fraction of La 0.036 0.011 0.050 0.057 0.042 0.050 AL Min Path (sccm) 11091.6(2) 2784.9 9070.2 10727.50 9250.7 11797.69 Fraction of La 0.032 0.008 0.026 0.039 0.033 0.043 22

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EVALUATION OF THE PROPOSED CHANGE (1) The predominant contributors to this mismatch are seen in five Penetrations (20A, 23, 41, 68 and 69), with details regarding these as follows:

Penetration 20A (0-N2-344, 1CV612, 1CV622, 1CV632, 1CV642), was AF tested within administrative limits. Following maintenance of 1CV632, subsequent AL testing resulted in a leakage rate of 1,110 sccm. This exceeded the administrative leakage limit of 800 sccm, but was not in excess of the maximum limit of 2,000 sccm. The AL leakage was accepted-as-is. AF/AL testing performed in 2014 measured a leakage rate of 1005 sccm, again exceeding the admin limit. CR-2014-001906 was initiated.

Penetration 23 (1CV4260) was AF tested within administrative limits. Following maintenance of 1CV4260, subsequent AL testing resulted in a leakage rate of 2,600 sccm.

This exceeded the administrative leakage limit of 296 sccm, but was not in excess of the maximum limit of 10,000 sccm. The AL leakage was accepted-as-is. AF/AL testing performed in 2014 measured a leakage rate of 1,321 sccm, again exceeding the admin limit. CR-2014-001688 was initiated.

Penetration 41 (1MOV651, 1MOV652) was AF tested within administrative limits. Following maintenance of 1MOV652, subsequent AL testing resulted in a leakage rate of 2,718 sccm.

This exceeded the administrative leakage limit of 1,770 sccm, but was not in excess of the maximum limit of 40,000 sccm. The AL leakage was accepted-as-is. AF testing performed in 2014 measured a leakage rate of 1,907 sccm, again exceeding the admin limit.

CR-2014-002264 was initiated. Subsequent AL testing performed following maintenance resulted in a leakage rate of 572 sccm.

Penetration 68 (Personnel Airlock) was AF/AL tested within administrative limits with a measured leakage rate of 14,790 sccm in 2010. Subsequent AF/AL testing performed in 2012 resulted in a leakage rate of 4,808.2 sccm. This was below the administrative limit of 8,000 sccm. AF/AL testing performed in 2014 measured a leakage rate of 3,698.59 sccm.

Penetration 68 has remained below the administrative limit of 8,000 sccm but was the largest contributor to the identified mismatch between AF and AL minimum pathway leakage in 2012.

Penetration 69 (Emergency Airlock) was AF/AL tested within administrative limits with a measured leakage rate of 878.4 sccm in 2010. Subsequent AF/AL testing performed in 2012 resulted in a leakage rate of 1,849.29 sccm. This was below the administrative limit of 8,000 sccm. AF/AL testing performed in 2014 measured a leakage rate of 2,404.08 sccm.

Penetration 69 has remained below the administrative limit of 8,000 sccm.

(2) The predominant contributors to this mismatch are seen in two Penetrations (2B and 21 SG), with details regarding these as follows:

Penetration 2B (2CVC435) - Unit 2 check valve CVC435 was AF tested within administrative leakage limits during the 2003 Refueling Outage (RFO). Subsequently, the valve was replaced during the 2003 RFO with an equivalent approved Enertech nozzle type check valve under a Calvert Cliffs work order. Subsequent AL testing of the new check valve resulted in a leakage rate of 1,488 sccm. This exceeded the administrative leakage limit of 296 sccm, but was not in excess of the maximum limit of 10,000 sccm.

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EVALUATION OF THE PROPOSED CHANGE Penetration 21 SG (21 SG South Manway) - 21 SG South Manway AF leakage rate was measured at 161.8 sccm, which was below the 2,500 sccm administrative leakage limit.

During the 2003 RFO, Calvert Cliffs replaced Unit 2s steam generators. The manway covers were reinstalled on the newly Steam Generator and this manway penetration was AL tested SAT below the administrative leakage limit of 2,500 sccm at 1,701 sccm.

Subsequently, manways have been removed and reinstalled during the 2005, 2007, 2009 RFOs for eddy current and visual inspections when necessary, and continued to remain within administrative leakage limits when tested each outage from 2005 through 2013. The 2003 RFO AF versus AL leakage rate differences can be directly attributed to removal and reinstallation of the manway on the replacement steam generator.

3.5 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 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 Calvert Cliffs Containments there are three issues of degradation that have been identified and corrected. The three areas of note involve:

" Moisture barrier seal degradation/Liner corrosion

• Containment concrete surface degradation

" Vertical tendon corrosion Each of these areas is discussed in detail in Sections 3.5.1 through 3.5.3, respectively.

3.5.1 Containment Liner and Moisture Barrier Seal Inspections Inspections on the containment liner are conducted in accordance with Examination Category E-A of the ASME Code Section Xl, 2004 Edition. These inspections are performed such that 100% of the accessible portion of the liner is inspected during each inspection period. As previously mentioned the portion of the liner that covers the containment foundation slab is considered inaccessible. Since this inaccessible area cannot be inspected, Calvert Cliffs must therefore evaluate its acceptability whenever conditions exist in the accessible areas that could indicate the presence of, or result in, degradation to the inaccessible area.

The moisture barrier seal is examined so that 100% of the seal is visually examined during each inspection period.

Inspection Results In 1994 Calvert Cliffs discovered significant age related degradation of the Unit 1 moisture barrier seal. As part of the corrective actions, a decision was made to subsequently replace the Unit 2 moisture barrier seal. In 1999 during the replacement of the Unit 2 moisture barrier seal, areas of pitting and general corrosion were discovered on the metal containment liner that exceeded 10% of the nominal wall thickness of W." The liner area of concern was the wall to floor transition under the moisture barrier seal, between the wearing floor slab and the containment liner wall.

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EVALUATION OF THE PROPOSED CHANGE An evaluation was subsequently performed which determined that if the degradation was not stopped, but instead continued at its current rate, the pitted areas would degrade further and pose a concern in the future. While it was impossible to determine when the pitting began, it was reasonable to assume that the degradation would be stopped or significantly slowed by the replacement of the moisture barrier seal.

As part of the evaluation, consideration was given as to whether additional areas needed to be examined. A determination was made that no additional examinations of other areas were necessary.

The evaluation also concluded that it was acceptable, in accordance with ASME Section XI, 1992 Addendum, Subsection IWE, Article IWE-3122.4, to return the liner to service without repair of the degraded area since the area of degradation is non-structural in nature and has no effect on the structural integrity of Containment.

The replacement of the moisture barrier seal involved use of a new seal material (high density silicone elastomer (HDSE)) that provides an effective seal against water, smoke, gas, and pressure. Along with the installation of the new HDSE sealant, a modification to the design of the seal was done. The original base sealant was applied to a shallow depth at the top of the compressible material in the joints and made flush with the nominal base slab. The new HDSE sealant was installed in such a manner to form a small curb above the joint, which would shed water in addition to providing a seal. Also, to improve the seal, the HDSE was placed a minimum of 3" into the joint by removing some of the compressible material. A polyethylene backer rod was then placed in the joint between the HDSE and the compressible material to separate them.

The replacement of the Unit 1 and 2 moisture barrier seals have been completed.

The most recent inspections of the containment liner and moisture barrier seal indicate that the replacement of the moisture barrier seal has arrested the corrosion and pitting throughout the affected area and has prevented any new areas of corrosion and pitting from occurring. As a result the liner continues to be acceptable to perform its safety function (i.e., act as a leak tight membrane). The Unit 2 moisture barrier continues to be subject to Augmented Inspections due to the identification of a crack and subsequent repair of the seal during the 2013 refueling outage.

3.5.2 Containment Concrete Concrete Inspections The reinforced concrete portions of Containment are inspected in accordance with Examination Category L-A of the ASME Code Section XI, 2004 Edition. The concrete containment structure is divided into 129 areas on Unit 1 and 115 areas on Unit 2. Calvert Cliffs conducts a 100%

visual examination of each unit every five years.

Inspection Results During the 2005 and 2007 inspections examiners identified new grease leaks, efflorescence, and other stains. All these items were entered into the corrective action program for resolution.

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EVALUATION OF THE PROPOSED CHANGE The examiners also identified issues on the Unit 1 and 2 Containments that had been identified in a previous inspection but had not been fully addressed. These items are:

• Containment structure dome area is suffering from the effects of weathering due to freeze-thaw cycles. This issue, if not addressed, would eventually pose a threat to containment integrity as water soaking into the concrete would attack the reinforcing steel. The occurrence of freeze-thaw cycles accelerates this process by breaking up the concrete surface. The proposed corrective action for this issue is to remove any loosened concrete, clean stains from around areas of major grease leaks, and apply a sealer to minimize moisture penetration. Completion of these actions for Unit 1 and 2 Containments are scheduled for July 1, 2015.

• Concrete was found to be delaminating around the sloped surface above the equipment hatch. Delamination opens the surface to water entry and could cause pieces of concrete to fall off. The proposed corrective action for this issue involves the removal of loose concrete and the application of an epoxy-bonding compound to which low slump 5000 psi concrete will be applied. Completion of these actions for Unit 1 and 2 Containments are scheduled for July 1, 2015.

An evaluation of these two issues determined the concrete in those areas is still capable of maintaining its structural integrity in the event of a design basis LOCA and that it will continue to perform this function beyond the completion date for the repairs.

3.5.3 Containment Tendons Containment Tendon Inspections The containment tendons are inspected in accordance with Examination Category L-B of the ASME Code Section XI, 2004 Edition. Table 3.5-1 below shows the tendon population distribution for Calvert Cliffs.

Table 3.5-1, Calvert Cliffs Tendon Population Distribution Vertical Vertical (Replaced or Unit Dome Hoop (Original) Restressed) Total 1 204 465 123 77 871 1 0 3 2 0 5 (Abandoned) 2 204 468 125 79 876 The ASME required tendon lift-off test is conducted on a minimum of 25 tendons once every ten-years. Per the ASME Code, a sample of each of the four types of active tendons must be examined. The sample selections by type are as follows:

• Dome: 5 (1 Common and 4 Random)

• Hoop: 10 (1 Common and 9 Random)

" Vertical (Original-Undisturbed): 6 (1 Common and 5 Random)

" Vertical (Replaced or Restressed): 4 (0 Common, 4 Random) 26

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EVALUATION OF THE PROPOSED CHANGE A common tendon is a tendon that is tested each time the test is performed. A random tendon is a tendon selected for this test and is not selected again in subsequent tests. The replaced or re-stressed tendons do not have a common tendon because these tendons were recently replaced as a result of the tendon issues discussed below.

The ASME required wire removal tensile test is conducted once every ten years on a minimum of one tendon from each of the four tendon types.

The ASME required visual examinations of the tendon anchorage area, the free water analysis and the analysis of the corrosion protection medium (grease analysis) are performed on a minimum of 25 tendons every five years. Each of these three tests is performed on tendons that are selected so as to have the same distribution between the four tendon types as for the tendon lift-off tests.

Inspection Results In 1997, during the performance of the 20-year (time from first tendon inspection) tendon surveillance on Unit 1, conditions that did not meet the acceptance standards were found on some of the Containment tendons. Conditions that did not meet the acceptance standards were found in all three containment tendon populations, i.e., hoop, dome, and vertical tendons. The abnormal conditions found on the hoop and dome tendons were considered minor enough that the acceptability of the concrete containment was not affected. However the conditions found on the vertical tendon population were more significant. Several of the vertical tendons selected for the surveillance were found to contain broken and corroded wires at their top ends, just below the stressing washer. The discovery of broken wires in these tendons initiated an expansion of the Unit 1 vertical tendon inspection scope to perform visual inspections and lift-off testing on all Unit 1 vertical tendons. Subsequently, broken and corroded wires were found throughout the Unit 1 vertical tendon population at the top ends of the tendons. Following completion of the Unit 1 surveillance, the 20-year surveillance of the Unit 2 tendons was conducted. Although Unit 2 was only required to perform visual inspections, it was decided to also perform lift-off testing of all the vertical tendons in order to facilitate inspection of the tendon wires in the region of concern [below the upper (top) stressing washer]. Abnormal conditions very similar to Unit 1 were found on the Unit 2 vertical tendons. A non-conformance report was written for every abnormally degraded condition that did not meet the acceptance criteria.

Corrective Actions to Address Vertical Tendon Corrosion As a result of the corrosion and broken wires discovered on some vertical tendons during the 1997 surveillance on the Unit 1 and 2 Containments, an evaluation was conducted. The evaluation concluded that the tendon wire failures and corrosion problems resulted from a combination of water and moist air intrusion into the vertical tendon end caps (grease cans),

and inadequate initial grease coverage of wires in the area just under the top stressing washer.

To address the issues identified in the evaluation, short-term and long-term corrective actions were taken. The short-term actions included spraying hot grease under the top stressing washer, reorienting the stressing shims so as to leave a gap between the shims to allow a vent path to help eliminate voids, re-greasing non-corroded vertical tendons, and resealing around the original tendon can all-thread penetrations with caulking. Additional inspections were performed in 1999 and 2000 to verify the assumptions that were considered in the evaluation and to provide additional data to help develop the long-term corrective action plan.

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EVALUATION OF THE PROPOSED CHANGE The goal of the long-term corrective action plan was to ensure that the Containments meet their design basis requirements until plant end-of-life. As one part of the long-term corrective action plan, all the vertical tendons were re-greased using a new corrosion inhibiting grease (Visconorust 2090-P4). The non-corroded vertical tendons were re-greased in 2000, and the tendons with less severe corrosion that were not replaced were re-greased during 2001. The remaining vertical tendon population (46 tendons per Unit) were replaced in 2001 and 2002, and had new grease put in place at that time. At the end of these corrective actions, all of the vertical tendons had a redesigned pressure-tight, grease-filled cap installed at the upper-bearing plate to prevent water intrusion. The bottom grease cap for every vertical tendon was also replaced with a new redesigned pressure-tight grease cap. The redesigned grease cap has a flange that is attached by studs and nuts to the tendon bearing plates by utilizing existing taps in the plates.

Enhanced Vertical Tendon Inspections To further confirm the effectiveness of the short- and long-term corrective actions, an enhanced inspection program was initiated that consisted of a two-tiered approach. The first tier involved the performance of the required, ASME Section XI Code inspections at their normal periodicity.

The second tier involved enhanced visual inspections of a selected sample size of vertical tendons that would be in addition to tendons inspected as part of the ASME required inspection.

The visual inspections included inspection of the anchorhead and buttonhead region to determine if any wire breaks have occurred in the area under the vertical tendon top-stressing washers. The first enhanced inspections were performed in 2005 and the second enhanced inspections were conducted in 2007. No new issues were identified as a result of these inspections.

Based on the satisfactory performance of the enhanced inspections, an assessment was conducted which determined that continuance of the enhanced inspections was not necessary.

The assessment determined that the Code required inspections are sufficient to adequately determine whether tendon performance remains acceptable.

Latest ASME Code Inspection Results The evaluation of the in-service inspection results for the 35th year, conducted in 2012 for Unit 1 and 2013 for Unit 2 containment structures, concluded that no abnormal degradation of the post-tensioning systems have been experienced.

3.6 License Renewal Aging Management The containment structures are in scope for license renewal based on 10 CFR 54.4(a).

Updated Final Safety Analysis Report, Chapter 16 lists the plausible age-related degradation mechanisms of the containment components. These age-related degradation mechanisms are managed through the conduct of various surveillance tests, in-service inspections, preventive maintenance activities, and maintenance procedures. These documents will continue to be modified as necessary to ensure they continue to provide reasonable assurance that the aging effects will be adequately managed throughout the operating life of the units.

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EVALUATION OF THE PROPOSED CHANGE 3.7 NRC SER Limitations and Conditions 3.7.1 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 3.7-1 were satisfied:

Table 3.7-1, NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition (From Section 4.0 of SE) Calvert Cliffs Response For calculating the Type A leakage rate, the Calvert Cliffs will utilize the definition in NEI licensee should use the definition in the NEI 94-01 Revision 3-A, Section 5.0. This TR94-01, Revision 2, in lieu of that in definition has remained unchanged from ANSI/ANS-56.8-2002. (Refer to SE Revision 2-A to Revision 3-A of NEI 94-01.

Section 3.1.1.1).

The licensee submits a schedule of Reference Sections 3.4.2 and 3.4.3 of this containment inspections to be performed submittal.

prior to and between Type A tests. (Refer to In addition to the scheduled Containment 151 SE Section 3.1.1.3). inspections, general visual observations of the accessible interior and exterior surfaces of the containment structure shall continue to be performed in accordance with STP-M-665-1 and STP-M-665-2 "Containment Visual Inspection."

Containment Liner - scheduled for inspection during each refueling outage in accordance with License Renewal Commitment and prior to each Type A test.

Containment Concrete - scheduled for inspection every 36 +/- 14 months and prior to every Type A test.

These are scheduled surveillance tests and are performed to meet the requirements of TS 5.5.16 and will ensure that the inspection requirements of NEI 94-01 Revision 3-A, Sections 9.2.1 and 9.2.3.2 continue to be satisfied.

The licensee addresses the areas of the Reference Sections 3.4.2 and 3.4.3 of this containment structure potentially subjected to submittal.

degradation. (Refer to SE Section 3.1.3). Procedures STP-M-665-1 and STP-M-665-2 "Containment Visual Inspection" are utilized to perform visual inspection of the normally accessible internal and exterior surfaces of the primary containment to identify evidence of structural deterioration, which could affect 29

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EVALUATION OF THE PROPOSED CHANGE Limitation/Condition (From Section 4.0 of SE) Calvert Cliffs Response either structural integrity or leak tightness.

These are scheduled surveillance test and are performed during each refueling outage.

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

inspections performed following major The Calvert Cliffs Unit 1 Steam Generators modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4).

The Calvert Cliffs Unit 2 Steam Generators were replaced in 2003.

The normal Type A test interval should be Calvert Cliffs will follow the requirements of less than 15 years. If a licensee has to utilize NEI 94-01 Revision 3-A, Section 9.1. This the provision of Section 9.1 of NEI TR 94-01, requirement has remained unchanged from Revision 2, related to extending the ILRT Revision 2-A to Revision 3-A of NEI 94-01.

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

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

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

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

3.7.2 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 Part 50, Appendix J. However, the NRC staff identified two conditions on the use of NEI TR 94-01, Revision 3 (Reference NEI 94-01 Revision 3-A, NRC SER 4.0, Limitations and Conditions):

Topical Report Condition I 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 30

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EVALUATION OF THE PROPOSED CHANGE 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.

Response to Condition 1 Condition 1 presents three (3) 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.

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 & C MNPLR total is greater than the Calvert Cliffs administrative 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 Calvert Cliffs administrative 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 Calvert Cliffs will apply the 9-month grace 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 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 31

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EVALUATION OF THE PROPOSED CHANGE 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 & 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 Condition 2 presents two (2) 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 & 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, Calvert Cliffs will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the As-Left leakage total for each Type C component currently on the 75 month extended test interval. This will result in a combined conservative Type C total for all 75 month LLRT 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 a 75 month extended interval is summed with the non-adjusted total of those Type C components being tested at less than the 75 month interval and the total of the Type B tested components, if the Minimum pathway leakage rate is greater than the Calvert Cliffs administrative leakage summation limit of 0.50 La, but less than the 32

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 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 Calvert Cliffs administrative 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 Minimum pathway leakage rate is less than the Calvert Cliffs administrative leakage summation limit of 0.50 La, then the acceptability of the 75 month LLRT extension for all affected Type C components has been adequately demonstrated and that the calculated local leak rate total represents the actual leakage potential of the penetrations.

In addition to Condition 1, Parts 1, 2 which deal with the MNPLR Type B & 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 cycle's 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 Calvert Cliffs in the event an adverse trend in the aforementioned potential leakage understatement adjusted Type B & 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 Calvert Cliffs an adverse trend is defined as three (3) consecutive increases in the final pre-RCS Mode change Type B & C MNPLR leakage summation value, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.

3.8 NRC Information Notice 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner The NRC issued Information Notice (IN) 2014-07 (Reference 21) to inform addressees of issues identified by the NRC staff concerning degradation of floor weld leak-chase channel systems of steel containment shell and concrete containment metallic liner that could affect leak-tightness and aging management of containment structures. The NRC expects that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems.

33

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE At Calvert Cliffs, a test pipe was provided for each continuous segment of the bottom liner plate weld chase test channels (equivalent to containment weld channels). The tops of the pipes are located above the cover slab and are sealed with caps. These pipes were initially used to test the leak tightness of the bottom liner. During the performance of an ILRT, the caps for the liner plate weld chase test channels are removed for the test and are replaced upon test completion.

Calvert Cliffs has performed a preliminary review of this IN and determined that this notice is not applicable to Calvert Cliffs. A more thorough analysis of this issue is in progress to determine if this issue is applicable to Calvert Cliffs and an update to the CISI Program is required.

3.9 Conclusion NEI 94-01, Revision 3-A, describes an NRC-accepted approach for implementing the performance-based requirements of 10 CFR Part 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. Calvert Cliffs is adopting the guidance of NEI 94-01, Revision 3-A for the Calvert Cliffs 1 and 2, 10 CFR Part 50, Appendix J testing program plan.

Based on the previous ILRT tests conducted at Calvert Cliffs I and 2, it may be concluded that 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 Part 50, Appendix J and the overlapping inspection activities performed as part of the following Calvert Cliffs Unit 1 and 2 inspection programs:

" Containment Inservice Inspection Program (IWE/IWL)

• Containment Inspections per TS 3.6.1.1

" Containment Coatings Inspection and Assessment Program This experience is supplemented by risk analysis studies, including the Calvert Cliffs 1 and 2 risk analysis provided in Attachment 3. The findings of the risk assessment confirm the general findings of previous studies, on a plant-specific basis, that extending the ILRT interval from ten to 15 years results in a very small change to the Calvert Cliffs Unit 1 and 2 risk profiles.

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 Part 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.

34

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 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 Part 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type C test frequency 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 Part 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 Band Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. 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 SER.

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 Part 50, Appendix J, as modified by the conditions and limitations summarized in Section 4.0 of the associated SE. This guidance included provisions for extending Type C LLRT intervals up to 75 months. Type C testing ensures that individual 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 35

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE modified by the associated SER and approved by the NRC, in a licensing action to satisfy the requirements of Option B to 10 CFR Part 50, Appendix J.

Based on the considerations above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will continue to be conducted in accordance with the site licensing basis, and (3) the approval of the proposed change will not be inimical to the common defense and security or to the health and safety of the public.

4.2 Precedent This request is similar in nature to the following license amendments to extend the Type A Test Frequency to 15 years, as previously authorized by the NRC:

• Nine Mile Point Nuclear Station Unit 2 (Reference 22)

• Arkansas Nuclear One, Unit 2 (Reference 23)

" Palisades Nuclear Plant (Reference 24)

" Virgil C. Summer Nuclear Station, Unit 1 (Reference 25) 4.3 Significant Hazards Consideration A change is proposed to the Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Units 1 and 2, Technical Specification (TS) 5.5.16, "Containment Leakage Rate Testing Program." The proposed change to the TS would revise Calvert Cliffs TS 5.5.16, by replacing the reference to Regulatory Guide (RG) 1.163 with a reference to Nuclear Energy Institute (NEI) topical report NEI 94-01 Revision 3-A (NRC-approved version specified in the 10 CFR Part 50, Appendix J Program Plan) as the implementation document used by Calvert Cliffs to implement the Units 1 and 2 performance-based leakage testing program in accordance with Option B of 10 CFR Part 50, Appendix J and incorporate the permanent 15 Year Integrated Leak Rate Test (ILRT) intervals and 75 Month Type C Test intervals in accordance with NEI 94-01 Revision 3-A. The proposed change also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for both Units 1 and 2 and exceptions from conducting post modification ILRT following replacement of the Units 1 and 2 Steam Generators.

Calvert Cliffs 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 change involve a significant increase in the probabilityor consequences of an accidentpreviously evaluated?

Response: No.

The proposed amendment to the TS involves the extension of the Calvert Cliffs Unit 1 and 2 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 36

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 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. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

As documented in NUREG-1493, Type 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 Calvert Cliffs Unit 1 and 2 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 ASME Section Xl, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extension does not significantly increase the consequences of an accident previously evaluated.

The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for both Units 1 and 2 and exceptions from conducting post modification ILRT following replacement of the Units 1 and 2 Steam Generators. These exceptions were for things that have already taken place so their deletion is solely an administrative action that has no effect on any component and no impact on how the units are 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 involves the extension of the Calvert Cliffs Unit 1 and 2 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 plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

37

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE The proposed amendment also deletes exceptions previously granted to allow one time extensions of the ILRT test frequency for both Units1 and 2 and exceptions from conducting post modification ILRT following replacement of the Units 1 and 2 Steam Generators.

These exceptions were for things that have already taken place so their deletion is solely an administrative action that does not result in any change in how the units are operated.

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 5.5.16 involves the extension of the Calvert Cliffs Unit 1 and 2 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 only the extension of the interval between Type A containment leak rate tests and Type C tests for Calvert Cliffs Unit 1 and 2. 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 XI 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 both Units 1 and 2 and exceptions from conducting post modification ILRT following replacement of the Units 1 and 2 Steam Generators. These exceptions were for things that have already taken place so their deletion is an administrative action and does not change how the units are operated and maintained, thus there is no reduction in any margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Calvert Cliffs concludes that the proposed amendment presents no 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.

38

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE 4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's 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 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 Part 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 Part 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 Part 50, Appendix J, October 2008

9. Letter from M. J. Maxin (NRC) to J. C. Butler (NEI), dated June 25, 2008, 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) 39

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE

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

11. Letter from D. G. McDonald, Jr (NRC) to C. H. Cruse (CCNPP), dated March 13, 1996, Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit No. 1 (TAC No.

M94500) and Unit No. 2 (TAC No. M94501)

12. Letter from A. W. Dromerick (NRC) to C. H. Cruse (CCNPP), dated February 11, 1997, Issuance of Amendments for Calvert Cliffs Nuclear Power Plant, Unit No. 1 (TAC No.

M97341) and Unit No. 2 (TAC No. M97342)

13. Letter from D. Skay (NRC) to C. H. Cruse (CCNPP), dated May 1, 2002, Calvert Cliffs Nuclear Power Plant, Unit No. 1 - Amendment Re: One Time Extension of Appendix J, Type A, Integrated Leak Rate Test Interval and Exception from performing a Post-Modification Type A Test (TAC No. MB3929)

14. Letter from D. Skay (NRC) to P. E. Katz (CCNPP), dated June 27, 2002, Calvert Cliffs Nuclear Power Plant, Unit No. 2 - Amendment Re: Exception from Performing a Post-Modification Integrated Leakage Rate Testing (TAC No. MB3444)

15. Letter from D. V. Pickett (NRC) to J. A. Spina (CCNPP), dated August 29, 2007, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Amendment Re: Implementation of Alternative Radiological Source Term (TAC Nos. MC8845 and MC8846)

16. Letter from D. V. Pickett (NRC) to G. H. Gellrich (CCNPP), dated March 22, 2011, Calvert Cliffs Nuclear Power Plant, Unit No. 2 - Amendment Re: One Time 5-Year Extension to the Containment Integrated Leak Rate Test Interval (TAC No. ME4804)

17. Letter from N. S. Morgan (NRC) to G. H. Gellrich (CCNPP), dated July 31, 2013, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 - Issuance of Amendments Regarding Peak Calculated Containment Internal Pressure (TAC Nos. ME9081 and ME9082)

18. EPRI-1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, October 2008

19. Letter from G. H. Gellrich (CCNPP) to Document Control Desk (NRC), dated September 24, 2013, License Amendment Request re: Transition to 10 CFR 50.48(c) -

NFPA 805 Performance Based Standard for Fire Protection

20. Regulatory Guide 1.147, Revision 16, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, October 2010

21. NRC Information Notice 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner, May 5, 2014

22. Letter from R. V. Guzman (NRC) to S. L. Belcher (NMP), dated March 30, 2010, Nine Mile Point Nuclear Station, Unit No. 2 - Issuance of Amendment Re: Extension of Primary Containment Integrated Leakage Rate Testing Interval (TAC No. ME1650)

23. Letter from N. K. Kalyanam (NRC) to Vice President, Operations (ANO), dated April 7, 2011, Arkansas Nuclear One, Unit No. 2 - Issuance of Amendment Re: Technical Specification Change to Extend Type A Test Frequency to 15 Years (TAC No. ME4090) 40

ATTACHMENT (1)

EVALUATION OF THE PROPOSED CHANGE

24. Letter from M. L. Chawala (NRC) to Vice President, Operations (PNP), dated April 23, 2012, Palisades Nuclear Plant - Issuance of Amendment to Extend the Containment Type A Leak Rate Test Frequency to 15 Years (TAC No. ME5997)

25. Letter from S. Williams (NRC) to T. D. Gatlin (VCSNS), dated February 5, 2014, Issuance of Amendment Extending Integrated Leak Rate Test Interval (TAC No.

MF1385) 41

ATTACHMENT (2)

MARKED UP TECHNICAL SPECIFICATIONS PAGE Calvert Cliffs Nuclear Power Plant September 18, 2014

Programs and Manuals 5.5 5.5 Progirams and Manuals 5.5.16 Containment Leakaqe Rate Testinq Proqram A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B.

This program shall be in accordance with the guidelines contained in R.gulatory Guide 1.163, ,Pnrf.rman....

INSERT 1 Based Cntainmnt Leak Test Pr.gram," dated September 1995, ineluding errata, as moiedifie Ay tche feliewilg exeeptions:

a. Nuclear Energy institcute (NEI) 94 01 1995, S..tion 9.2.3:, The first Unit 1 Type A test perfermced a:ft the Junc 15, 1992 Type A test shall be p.rfor*..d no latter than Juno 14, 2007. The flirst Unitt 2 Type -A test performold a~fter the May 2, 2001 Type A tost shall be performold ne later than May 1, 2016.

b. I"PA Unit^A 1 .P6*is emecpted

-iA P ... *PQh n"-R pest frefm TPA^ medifioatien 4Q

• Q,~.A.*,- integrated 4 ^

T .A...

leakage rate test infg reguirefmonts associated wit steamf gemerater replaeecmont-.

e. Unit 2 is emeepted fromA pest m~edifieatien integrated

• *m* m* ._1J_

~t~ay~nrat rerpiaccecnt The peak calculated containment internal pressure for the design basis loss-of-coolant accident, Pa, is 49.7 psig.

The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, La, shall be 0.16 percent of containment air weight per day at Pa.

Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is -5 1.0 La. During the first unit startup following testing, in accordance with this program, the leakage rate acceptance criterion are -50.60 La for Types B and C tests and f=0.75 La for Type A tests.

INSERT 1: NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012 CALVERT CLIFFS - UNIT 1 5.5-17 Amendment No. 3G3 CALVERT CLIFFS - UNIT 2 Amendment No. 284

ATTACHMENT (3)

EVALUATION OF RISK SIGNIFICANCE OF PERMANENT ILRT EXTENSION Calvert Cliffs Nuclear Power Plant September 18, 2014

.IHUGHES 0ASSOCIATES ENGINEERS CONSULTANTS SCIENTISTS Calvert Cliffs Nuclear Power Plant:

Evaluation of Risk Significance of Permanent ILRT Extension 0054-0001 -000-CALC-001 Prepared for:

Calvert Cliffs Nuclear Power Plant Project Number: 0054-0001-000 Project

Title:

Permanent ILRT Extension Revision: 3 Name and Date Dtg]Wty signed by Maln Johnson Preparer Matthew Johnson Matt Johns am the author of this o 2014 09.12 09:36:14-05'00'

. D.. tal*y signed by Nicholas Reviewer: Nicholas Lovelace t K._ "' o* ace L Lol Date 2014.09.12 14:06:09-05'00' Review Method Design Review Calculation D Approved by: Richard Anoba V C '7al /I t/

Page 1 ot93 Revision 33 Page I of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension REVISION RECORD

SUMMARY

Revision Revision Summary 0 Initial Issue

.1 Incorporated True North review comments.

2 Incorporated minor comments regarding NFPA 805 transition in Section 5.1.2.

3 Removed generic QA condition statement and generic containment overpressure discussion in section 2.0.

Page 2of93 Revision 33 Page 2 of 93

IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension TABLE OF CONTENTS 1 .0 P UR P O S E...................................................................................................................... 4 2 .0 S C O P E ........................................................................................................................... 4

3.0 REFERENCES

....................................................................................................... 6 4.0 ASSUMPTIONS AND LIMITATIONS ........................................................................... 8 5.0 METHODOLOGY and analysis .................................................................................... 8 5 .1 Inp u ts ........................ ................................................................................................... 8 5.1.1 General Resources Available ............................................................................. 8 5.1.2 Plant Specific Inputs ....................................................................................... 11 5.1.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large) .......................................................................................... 14 5.1.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage

.......................................................................................................................... 16 5 .2 A n a ly s is ...................................................................................................................... 18 5.2.1 Step 1 - Quantify the Baseline Risk in Terms of Frequency per Reactor Year ..... 19 5.2.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose) ........... 23 5.2.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Y e a rs ................................................................................................................. 24 5.2.4 Step 4 - Determine the Change in Risk in Terms of LERF ............................... 27 5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability (C C F P ) ........................................................................................................ . . 30 5 .3 S e n s itiv itie s ................................................................................................................. 31 5.3.1 Potential Impact from External Events Contribution ........................................ 31 5.3.1.1 Potential Impact from External Events Contribution Using IPEEE Fire Analysis ............................................................................................... 33 5.3.2 Potential Impact from Steel Liner Corrosion Likelihood .................................... 35 5.3.3 Expert Elicitation Sensitivity ............................................................................ 36 5.3.4 Large Leak Probability Sensitivity Study ....................................................... 38 6 .0 R ES ULTS ...................................................................................................................... 40

7.0 CONCLUSION

S AND RECOMMENDATIONS ........................................................ 42 A. Atta ch m e nt 1 ................................................................................................................. 43 Page 3 of 93 Revision 33 Page 3 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 1.0 PURPOSE The purpose of this analysis is to provide a risk assessment of extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) to permanent fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for the Calvert Cliffs Nuclear Power Plant (CCNPP). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1], the methodology used in EPRI TR-104285 [Reference 2], the NEI "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals" from November 2001 [Reference 3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 as applied to ILRT interval extensions, and risk insights in support of a request for a plant's licensing basis as outlined in Regulatory Guide (RG) 1.174 [Reference 4], the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval [Reference 5],

the methodology used in EPRI 1009325, Revision 2-A [Reference 24], and the methodology improvements in EPRI 1018243 [Reference 24].

2.0 SCOPE Revisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing frequency requirement from three in ten years to at least once in ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage rate was less than limiting containment leakage rate of 1La.

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, "Performance-Based Containment Leak Test Program," September 1995 [Reference 6], provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessment of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRC's rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project TR-104285, "Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals".

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

NEI 94-01 Revision 2-A contains a Safety Evaluation Report that supports using EPRI Report No. 1009325 Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, for performing risk impact assessments in support of ILRT extensions [Reference 24].

The Guidance provided in Appendix H of EPRI Report No. 1009325 Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.

The acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 10-6 per reactor year. RG 1.174 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the Conditional Containment Failure Probability (CCFP), which helps ensure the defense-in-depth philosophy is maintained, is also calculated.

Regarding CCFP, changes of up to 1.1% have been accepted by the NRC for the one-time requests for extension of ILRT intervals. In context, it is noted that a CCFP of 1/10 (10%) has been approved for application to evolutionary light water designs. Given these perspectives, a change in the CCFP of up to 1.5% is assumed to be small.

In additional, the total annual risk (person rem/year population dose) is examined to demonstrate the relative change in this parameter. While no acceptance guidelines for these additional figures of merit are published, examinations of NUREG-1493 and Safety Evaluation Reports (SER) for one-time interval extension (summarized in Appendix G) indicate a range of incremental increases in population dose that have been accepted by the NRC. The range of incremental population dose Increases is from <0.01 to 0.2 person-rem/year and/or 0.002% to 0.46% of the total accident dose. The total doses for the spectrum of all accidents (NUREG-1493 [Reference 6], Figure 7-2) result in health effects that are at least two orders of magnitude less than the NRC Safety Goal Risk. Given these perspectives, a very small population dose is defined as an increase from the baseline interval (3 tests per 10 years) dose of <1.0 person-rem per year or 1% of the total baseline dose, whichever is less restrictive for the risk impact assessment of the proposed extended ILRT interval.

For those plants that credit containment overpressure for the mitigation of design basis accidents, a brief description of whether overpressure is required should be included in this section. In addition, if overpressure is included in the assessment, other risk metrics such as CDF should be described and reported.

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3.0 REFERENCES

The following references were used in this calculation:

1. Revision 3-A to Industry Guideline for Implementing Performance-BasedOption of 10 CFR Part 50, Appendix J, NEI 94-01, July 2012.

2. Risk Impact Assessment of Revised ContainmentLeak Rate Testing Intervals, EPRI, Palo Alto, CA EPRI TR-104285, August 1994.

3. Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Revision 4, developed for NEI by EPRI and Data Systems and Solutions, November 2001.

4. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, May 2011.

5. Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C. H.

Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, Docket No.

50-317, March 27, 2002.

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

7. Evaluation of Severe Accident Risks: Surry Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 3, Revision 1, Part 1, October 1990.

8. Letter from R. J. Barrett (Entergy) to U. S. Nuclear Regulatory Commission, IPN-01-007, January 18, 2001.

9. United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB0178), April 17, 2001.

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

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

12. Technical Findings and Regulatory Analysis for Generic Safety Issue I1.E.4.3

'Containment Integrity Check', NUREG-1273, April 1988.

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

14. Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAM TM , EPRI, Palo Alto, CA, TR-105189, Final Report, May 1995.

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

16. United States Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.

17. Calculation No. CO-QU-001, Revision 1, Calvert Cliffs Nuclear Power Plant, Unit 1, "PRA Quantification (QU) Notebook," July 2012.

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18. Calculation No. NC-94-020, "Severe Accident Analysis of Calvert Cliffs for IPE Level I1,"

December 1994.

19. Massoud, M., Calculation No. CA07463, Revision 0, "2010 Update of Dose Analysis for Level 3 PRA Release Categories," August 2001.

20. Anthony R. Pietrangelo, One-time extensions of containment integrated leak rate test interval - additional information, NEI letter to Administrative Points of Contact, November 30, 2001.

21. Letter from J. A. Hutton (Exelon, Peach Bottom) to U. S. Nuclear Regulatory Commission, Docket No. 50-278, License No. DPR-56, LAR-01-00430, dated May 30, 2001.

22. Risk Assessment for Joseph M. Farley NuclearPlant Regarding ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, P0293010002-1929-030602, March 2002.

23. Letter from D. E. Young (Florida Power, Crystal River) to U. S. Nuclear Regulatory Commission, 3F0401-11, dated April 25, 2001.

24. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA. 1018243, October 2008.

25. Risk Assessment for Vogtle Electric Generating Plant Regarding the ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, February 2003.

26. Perspectives Gained from the IPEEE Program, USNRC, NUREG-1742, April 2002.

27. Procedure STP M-662-1, Revision 6, Calvert Cliffs Nuclear Power Plant, Unit 1, "Integrated Leak Rate Test Unit 1 Containment."

28. Procedure STP M-662-2, Revision 7, Calvert Cliffs Nuclear Power Plant, Unit 2, "Integrated Leak Rate Test Unit 2 Containment."

29. Calculation No. CO-QU-002, Revision 1, Calvert Cliffs Nuclear Power Plant, Unit 2, "PRA Quantification (QU) Notebook," August 2010.

30. Calculation No. RSC 10-21, Revision 0, Calvert Cliffs Nuclear Power Plant, Unit 2, "Evaluation of Risk Significance of ILRT Extension," August 2010.

31. Armstrong, J., Simplified Level 2 Modeling Guidelines: WOG PROJECT: PA-RMSC-0088, Westinghouse, WCAP-1 6341 -P, November 2005.

32. Calculation CO-LE-001, Revision 1, CENG, Units 1 and 2, "PRA Level 2 Notebook," May 2010.

33. Landale, J., PRAER No. CO-2010-012, CENG, CO-2010-012, August 2010.

34. Harrison, D., Generic Component Fragilities for the GE Advanced BWR Seismic Analysis, International Technology Corporation, September 1988.

35. Calculation No. RAN 97-031, IPEEE, Calvert Cliffs Nuclear Power Plant, "Individual Plant Examination of External Events," August 1997.

Page 7 of 93 Revision 33 Revision Page 7 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 4.0 ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculation:

" The technical adequacy of the CCNPP PRA is consistent with the requirements of Regulatory Guide 1.200 as is relevant to this ILRT interval extension, as detailed in Attachment 1.

" The CCNPP Level 1 and Level 2 internal events PRA models provide representative results.

" It is appropriate to use the CCNPP internal events PRA model as a gauge to effectively describe the risk change attributable to the ILRT extension. An extensive sensitivity study is done in Section 5.3.1 to show the effect of including external event models for the ILRT extension. The IPEEE simplified seismic PRA [Reference 35] and the detailed Fire PRA (model 6.1 M) are used for this sensitivity analysis. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose) will not substantially differ if detailed analysis of seismic events were to be included in the calculations.

" Accident classes describing radionuclide release end states are defined consistent with EPRI methodology [Reference 2].

" 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 sequences is 10La based on the previously approved methodology performed for Indian Point Unit 3 [Reference 8, Reference 9].

" The representative containment leakage for Class 3b sequences is 100La based on the guidance provided in EPRI Report No. 1009325, Revision 2-A (EPRI 1018243)

[Reference 24].

" The Class 3b can be very conservatively categorized as LERF based on the previously approved methodology [Reference 8, Reference 9].

" The impact on population doses from containment bypass scenarios is not altered by the proposed ILRT extension, but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes 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.

5.0 METHODOLOGY AND ANALYSIS 5.1 Inputs This section summarizes the general resources available as input (Section 5.1.1) and the plant specific resources required (Section 5.1.2).

5.1.1 General Resources Available Various industry studies on containment leakage risk assessment are briefly summarized here:

1. NUREG/CR-3539 [Reference 10]

2. NUREG/CR-4220 [Reference 11]

3. NUREG-1273 [Reference 12]

4. NUREG/CR-4330 [Reference 13]

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5. EPRI TR-105189 [Reference 14]

6. NUREG-1493 [Reference 6]

7. EPRI TR-104285 [Reference 2]

8. NUREG-1150 [Reference 15] and NUREG/CR-4551 [Reference 7]

9. NEI Interim Guidance [Reference 3, Reference 20]

10. Calvert Cliffs liner corrosion analysis [Reference 5]

11. EPRI Report No. 1009325, Revision 2-A (EPRI 1018243), Appendix H [Reference 24]

This first study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and is to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different containment leakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRC's cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh study is an EPRI study of the impact of extending ILRT and LLRT test intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the basis for the consequence analysis of the ILRT interval extension for CCNPP. The ninth study includes the NEI recommended methodology (promulgated in two letters) for evaluating the risk associated with obtaining a one-time extension of the ILRT interval. The tenth study addresses the impact of age-related degradation of the containment liners on ILRT evaluations. Finally, the eleventh study builds on the previous work and includes a recommended methodology and template for evaluating the risk associated with a permanent 15-year extension of the ILRT interval.

NUREG/CR-3539 [Reference 101 Oak Ridge National Laboratory documented a study of the impact of containment leak rates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 [Reference 16]

as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

NUREG/CR-4220 [Reference 111 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.

NUREG-1273 [Reference 121 A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREG/CR-4220 database. This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected. In addition, this study noted that local leak rate tests can detect "essentially all potential degradations" of the containment isolation system.

NUREG/CR-4330 [Reference 131 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 Revision 3 Page 9 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension modeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals.

However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies:

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

EPRI TR-105189 [Reference 141 The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessment because it provides insight regarding the impact of containment testing on shutdown risk. This study contains a quantitative evaluation (using the EPRI ORAM software) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk. The conclusion from the study is that a small, but measurable, safety benefit is realized from extending the test intervals.

NUREG-1493 [Reference 61 NUREG-1493 is the NRC's cost-benefit analysis for proposed alternatives to reduce containment leakage testing intervals and/or relax allowable leakage rates. The NRC conclusions are consistent with other similar containment leakage risk studies:

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

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

EPRI TR-104285 [Reference 21 Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-1 05189 study),

the EPRI TR-104285 study is a quantitative evaluation of the impact of extending ILRT and LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with NUREG-1 150 Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-1 04285 uses a simplified Containment Event Tree to subdivide representative core damage frequencies into eight classes of containment response to a core damage accident:

1. Containment intact and isolated

2. Containment isolation failures dependent upon the core damage accident

3. Type A (ILRT) related containment isolation failures

4. Type B (LLRT) related containment isolation failures

5. Type C (LLRT) related containment isolation failures

6. Other penetration related containment isolation failures

7. Containment failures due to core damage accident phenomena

8. Containment bypass Consistent with the other containment leakage risk assessment studies, this study concluded:

"...the proposed CLRT (Containment Leak Rate Tests) frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms. For example, for the PWR analyzed, the change is about 0.04 person-rem per year...

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension NUREG-1 150 [Reference 151 and NUREG/CR-4551 [Reference 71 NUREG-1 150 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 Surry. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the CCNPP Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent CCNPP. (The meteorology and site differences other than population are assumed not to play a significant role in this evaluation.)

NEI Interim Guidance for Performinq Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals [Reference 3, Reference 201 The guidance provided in this document builds on the EPRI risk impact assessment methodology [Reference 2] and the NRC performance-based containment leakage test program

[Reference 6], and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

Calvert Cliffs Response to Request for Additional Information Concerning the License Amendment for a One-Time Integrated Leakage Rate Test Extension [Reference 51 This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting liner corrosion, and the corresponding change in risk. The methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms was factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete base-mat, each with a steel liner.

EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals [Reference 241 This report provides a generally applicable assessment of the risk involved in extension of ILRT test intervals to permanent 15-year intervals. Appendix H of this document provides guidance for performing plant-specific supplemental risk impact assessments and builds on the previous EPRI risk impact assessment methodology [Reference 2] and the NRC performance-based containment leakage test program [Reference 6], and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

The approach included in this guidance document is used in the CCNPP assessment to determine the estimated increase in risk associated with the ILRT extension. This document includes the bases for the values assigned in determining the probability of leakage for the EPRI Class 3a and 3b scenarios in this analysis, as described in Section 5.2.

5.1.2 Plant Specific Inputs The plant-specific information used to perform the CCNPP ILRT Extension Risk Assessment includes the following:

" Level 1 Model results: Unit 1 [Reference 17] and Unit 2 [Reference 29]

" Level 2 Model results [Reference 17, Reference 18, Reference 19]

" Release category definitions used in the Level 2 Model [Reference 18, Reference 19]

" Dose within a 50-mile radius [Reference 19]

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1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

" ILRT results to demonstrate adequacy of the administrative and hardware issues

[Reference 30]

" Containment failure probability data [Reference 18, References 32 and 33]

Level 1 Model The Level 1 Internal Events PRA Model that is used for CCNPP is characteristic of the as-built plant. The current Level 1 model (CCNPP PRA Model Version 6.2a) [Reference 17] is a linked fault tree model, and was quantified with the total Internal Events Core Damage Frequency (CDF) = 1.61 E-5/year for Unit 1 and CDF = 1.41 E-5/year for Unit 2. The total External Event CDF (excluding seismic) = 3.24E-5/year for Unit 1 and 3.71 E-5/year for Unit 2. Table 5-1 provides a summary of the Internal Events CDF results for CCNPP PRA Model Version 6.2a.

Table 5-2 provides a summary of the External Events CDF results. The High Winds are included in CCNPP PRA Model Version 6.2a. The Fire PRA results come from Model Version 6.1 M. The Seismic PRA results come from the IPEEE Seismic Analysis [Reference 35].

Table 5 Internal Events CDF (CCNPP PRA Model Version 6.2a)

Internal Events Unit I Frequency (per year) Unit 2 Frequency (per year)

LOCAs 5.88E-6 7.70E-6 Internal Floods 6.18E-6 1.06E-6 Transients 3.40E-6 4.70E-6 ISLOCA 1.97E-7 1.97E-7 SGTR 4.71E-7 4.60E-7 Total Internal Events CDF 1.61 E-5 1.41 E-5 Total Internal Events CDF 1.34E-5 (Excluding ISLOCA & SGTR)

Table 5 External Events CDF External Events Unit I Frequency (per year) Unit 2 Frequency (per year)

Fire 3.15E-5 3.59E-5 High Winds 9.19E-7 1.23E-6 Seismic 1.07E-5 1.07E-5 Total External Events CDF 4.31E-5 4.78E-5 Note that the above Fire PRA values reflect the anticipated configuration of the plant upon full implementation of NFPA 805 and related plant modifications to resolve fire protection issues.

Refer to Section 5.3.1.

Level 2 Model The Level 2 Model that is used for CCNPP was developed with guidance from WCAP-16341-P to calculate the LERF contribution, as well as the other release end states evaluated in the model: INTACT, SERF (small early release frequency), and LATE [Reference 31]. The current LERF model (CCNPP PRA Model Version 6.2a) [Reference 17] is a linked fault tree model and was quantified with the total Unit 1 Internal Events LERF = 1.39E-6/year and Unit 2 Internal Revision 3 Page 12 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Events LERF = 1.56E-6/year. The total Unit I External Event LERF (excluding seismic) =

2.99E-6/year and Unit 2 External Event LERF (excluding seismic) = 4.21 E-6/year. Table 5-3 provides a summary of the Internal Events LERF results for CCNPP PRA Model Version 6.2a.

Table 5-4 provides a summary of the External Events CDF results. The High Winds are included in CCNPP PRA Model Version 6.2a. The Fire PRA results come from Model Version 6.1M. The Seismic PRA results come from the IPEEE Seismic Analysis [Reference 35].

Table 5 Internal Events LERF (CCNPP PRA Model Version 6.2a)

Internal Events Unit 1 Frequency (per year) Unit 2 Frequency (per year)

LOCAs 3.26E-7 4.01 E-7 Internal Floods 2.46E-7 2.17E-7 Transients 1.50E-7 2.84E-7 ISLOCA 1.97E-7 1.97E-7 SGTR 4.71E-7 4.60E-7 Total Internal Events LERF 1.39E-6 1.56E-6 Table 5 External Events LERF External Events Unit I Frequency (per year) Unit 2 Frequency (per year)

Fire 2.97E-6 4.17E-6 High Winds 2.21E-8 3.77E-8 Seismic 1.41E-6 1.41E-6 Total External Events CDF 4.40E-6 5.62E-6 Note that the above Fire PRA values reflect the anticipated configuration of the plant upon full implementation of NFPA 805 and related plant modifications to resolve fire protection issues.

Refer to Section 5.3.1.

Population Dose Calculations The population dose calculation was performed for the CCNPP Severe Accident Mitigation Alternatives (SAMA) analyses [Reference 19] in 2010. Table 5-5 presents dose exposures calculated from methodology described in Reference 1 and data from Reference 19. Reference 19 provides the population dose (person-rem) for Classes 1, 2, 6, 7, and 8; Class 3a and 3b population dose values are calculated from the Class 1 population dose and represented as 1OLa and 1OOL a, respectively, as guidance in Reference 1 dictates.

Table 5 Population Dose Accident Class Description Release (person-rem) 1 Containment Remains Intact 3.20E+04 2 Containment Isolation Failures 2.OOE+07 3a Independent or Random Isolation Failures SMALL 3.20E+05 1 3b Independent or Random Isolation Failures LARGE 3.20E+06 2 Revision 3 Page 13 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of permanent ILRT Extension Table 5 Population Dose Accident Class Description Release (person-rem)

Isolation Failure in which pre-existing leakage is not n/a dependent on sequence progression. Type B test Failures 5 Isolation Failure in which pre-existing leakage is not n/a dependent on sequence progression. Type C test Failures Isolation Failure that can be verified by IST/IS or 7.01 E+06 6

surveillance 7 Containment Failure induced by severe accident 5.61 E+07 8 Accidents in which containment is by-passed 2.25E+07

1. 10*La

2. 100"*L Release Cateaorv Definitions Table 5-6 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology [Reference 2]. These containment failure classifications are used in this analysis to determine the risk impact of extending the Containment Type A test interval, as described in Section 5.2 of this report.

Table 5 EPRI Containment Failure Classification [Reference 2]

Class Description Containment remains intact including accident sequences that do not lead to containment failure in the 1 long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant.

Containment isolation failures (as reported in the Individual Plant Examinations) including those accidents in which there is a failure to isolate the containment.

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.

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.

Independent (or random) isolation failures including those accidents in which the pre-existing isolation 5 failure to seal is not dependent on the sequence in progress. This class is similar to Class 4 isolation failures, but is applicable to sequences involving Type C test and their potential failures.

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

Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J 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.

5.1.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large)

The ILRT can detect a number of component failures such as liner breach, failure of certain bellows arrangements and failure of some sealing surfaces, which can lead to leakage. The proposed ILRT test interval extension may influence the conditional probability of detecting these types of failures. To ensure that this effect is properly addressed, the EPRI Class 3 Revision 3 Page 14 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of permanent ILRT Extension accident class, as defined in Table 5-6, is divided into two sub-classes, Class 3a and Class 3b, representing small and large leakage failures respectively.

The probability of the EPRI Class 3a and Class 3b failures is determined consistent with the EPRI Guidance [Reference 24]. For Class 3a, the probability is based on the maximum likelihood estimate of failure (arithmetic average) from the available data (i.e., 2 "small" failures in 217 tests leads to "large" failures in 217 tests (i.e., 2/217 = 0.0092). For Class 3b, the probability is based on the Jeffrey's Non-Uniform Prior (i.e., 0.5/ 218 = 0.0023).

In a follow-up letter [Reference 20] to their ILRT guidance document [Reference 3], NEI issued additional information concerning the potential that the calculated delta LERF values for several plants may fall above the "very small change" guidelines of the NRC Regulatory Guide 1.174

[Reference 4]. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for ALERF. NEI describes ways to demonstrate that, using plant-specific calculations, the ALERF is smaller than that calculated by the simplified method.

The supplemental information states:

The methodology employed for determining LERF (Class 3b frequency) involves conservatively multiplying the CDF by the failure probability for this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, some plant-specific accident classes leading to core damage are likely to include individual sequences that eithermay already (independently)cause a LERF or could never cause a LERF, and are thus not associatedwith a postulatedlarge Type A containmentleakage path (LERF). These contributorscan be removed from Class 3b in the evaluation of LERF by multiplying the Class 3b probabilityby only that portion of CDF that may be impacted by Type A leakage.

The application of this additional guidance to the analysis for CCNPP, as detailed in Section 5.2, involves the following:

" The Class 2 and Class 8 sequences 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 2 events refer to sequences with large pre-existing containment isolation failures; Class 8 events refer to sequences with containment bypass events. These sequences are already considered to contribute to LERF in the CCNPP Level 2 PRA analysis.

" A review of Class 1 accident sequences shows that several of these cases involve successful operation of containment sprays. For calculation of the Class 3b and Class 3a frequencies, the fraction of the Class 1 CDF associated with successful operation of containment sprays could also be subtracted. Successful operation of containment sprays result in lower containment pressure with subsequent reduction in containment leakage. This conservatism was removed for the CCNPP ILRT analysis, as detailed in Section 5.2.4.

Consistent with the NEI Guidance [Reference 3], the change in the leak detection probability can be estimated by comparing the average time that a leak could exist without detection. For example, the average time that a leak could go undetected with a three-year test interval is 1.5 years (3 years / 2), and the average time that a leak could exist without detection for a ten-year interval is 5 years (10 years / 2). This change would lead to a non-detection probability that is a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRT testing.

Correspondingly, an extension of the ILRT interval to 15 years can be estimated to lead to a factor of 5 ((15/2)/1.5) increase in the non-detection probability of a leak.

Page 15 of 93 Revision 3 Revision 3 Page 15 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension It should be noted that using the methodology discussed above is very conservative compared to previous submittals (e.g., the IP3 request for a one-time ILRT extension that was approved by the NRC [Reference 9]) because it does not factor in the possibility that the failures could be detected by other tests (e.g., the Type B local leak rate tests that will still occur). Eliminating this possibility conservatively over-estimates the factor increases attributable to the ILRT extension.

5.1.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage An estimate of the likelihood and risk implications of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is evaluated using the methodology from the Calvert Cliffs liner corrosion analysis [Reference 5]. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete base-mat, each with a steel liner.

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

Differences between the containment base-mat and the containment cylinder and dome:

" The historical steel liner flaw likelihood due to concealed corrosion

" The impact of aging

" The corrosion leakage dependency on containment pressure

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

" Consistent with the Calvert Cliffs analysis, a half failure is assumed for base-mat concealed liner corrosion due to the lack of identified failures. (See Table 5-7, Step 1)

" The two corrosion events used to estimate the liner flaw probability in the Calvert Cliffs previous analysis are assumed to still be applicable.

" Consistent with the Calvert Cliffs analysis, the estimated historical flaw probability is also limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date (and have been performed since the time frame of the Calvert Cliffs analysis), and there is no evidence that additional corrosion issues were identified (See Table 5-7, Step 1).

" Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages (See Table 5-7, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.

" In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere, given that a liner flaw exists, was estimated as 1.1% for the cylinder and dome, and 0.11% (10% of the cylinder failure probability) for the base-mat. These values were determined from an assessment of the probability versus containment pressure. For CCNPP, the ILRT maximum pressure is psig 50 [References 27 and 28]

and ultimate pressure of 132 psig [References 32 and 33]. Probabilities of 1% for the cylinder and dome, and 0.1% for the base-mat are used in this analysis, and sensitivity studies are included in Section 5.3.2 (See Table 5-7, Step 4).

" Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack Revision 3 Page 16 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension formation) in the base-mat region is considered to be less likely than the containment cylinder and dome region (See Table 5-7, Step 4).

Consistent with the Calvert Cliffs analysis, a 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used.

To date, all liner corrosion events have been detected through visual inspection (See Table 5-7, Step 5).

Consistent with the Calvert Cliffs analysis, all non-detectable containment failures are assumed to result in early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

Table 5 Steel Liner Corrosion Base Case Step Description Containment Cylinder and Containment Basemat Dome (85%) (15%)

Historical liner flaw likelihood Events: 2 Events: 0 Failure data: containment location (Brunswick 2 and North Anna 2) Assume a half failure specific 2 / (70 x 5.5) = 5.19E-03 0.5 / (70 x 5.5) = 1.30E-03 Success data: based on 70 steel-lined containments and 5.5 years since the 10CFR 50.55a requirements of periodic visual inspections of containment surfaces Year Failure rate Year Failure rate Aged adjusted liner flaw likelihood During the 15-year interval, assume 1 2.05E-03 1 5.13E-04 failure rate doubles every five years average 5-10 5.19E-03 average 5-10 1.30E-03 (14.9% increase per year). The 15 1.43E-02 15 3.57E-03 average for the 5th to 10th year set to the historical failure rate. 15 year average = 6.44E-03 15 year average = 1.61 E-03 Increase in flaw likelihood between 3 and 15 years Uses aged adjusted 0.73% (1 to 3 years) 0.18% (1 to 3 years) 3 liner flaw likelihood (Step 2), 4.18% (1 to 10 years) 1.04% (1 to 10 years) assuming failure rate doubles every 9.66% (1 to 15 years) 2.41% (1 to 15 years) five years.

Likelihood of breach in containment 1% 0.1%

given liner flaw 10%

5% failure to identify visual flaws plus 5% likelihood that the flaw is Visual inspection detection failure not visible (not through-cylinder 100%

likelihood but could be detected by ILRT) All Cannot be visually inspected events have been detected through visual inspection. 5%

visible failure detection is a conservative assumption.

0.00073% (3 years) 0.000180% (3 years) 0.73% x 1% x 10% 0.18% x 0.1% x 100%

Likelihood of non-detected 0.00418% (10 years) 0.00104% (10 years) 6 containment leakage (Steps 3 x 4 x 4.18%x1%x1% 1.04%x0.1%x10%

5) 0.00966% (15 years) 0.00241% (15 years) 9.66% x 1% x 10% 2.41% x 0.1% x 100%

Revision 3 Page 17 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the containment cylinder and dome, and the containment base-mat, as summarized below for CCNPP.

Table 5 Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for CCNPP Description At 3 years: 0.00073% + 0.000180% = 0.00091%

At 10 years: 0.00418% + 0.00104% = 0.00522%

At 15 years: 0.00966% + 0.00241% = 0.01207%

The above factors are applied to those core damage accidents that are not already independently LERF or that could never result in LERF.

5.2 Analysis The application of the approach based on the guidance contained in EPRI Report No. 1009325, Revision 2-A, Appendix H [Reference 24], EPRI TR-1 04285 [Reference 2] and previous risk assessment submittals on this subject [References 5, 8, 21, 22, and 23] have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report, as described in Table 5-6.

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

" Core damage sequences in which the containment remains intact initially and in the long term (EPRI TR-104285, Class 1 sequences [Reference 2]).

" Core damage sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B or Type C test components. For example, liner breach or bellow leakage (EPRI TR-104285, Class 3 sequences [Reference 2]).

" Accident sequences involving containment bypassed (EPRI TR-1 04285, Class 8 sequences [Reference 2]), large containment isolation failures (EPRI TR-104285, Class 2 sequences [Reference 2]), and small containment isolation "failure-to-seal" events (EPRI TR-104285, Class 4 and 5 sequences [Reference 2]) are accounted for in this evaluation as part of the baseline risk profile. However, they are not affected by the ILRT frequency change.

" Class 4 and 5 sequences are impacted by changes in Type B and C test intervals; therefore, changes in the Type A test interval do not impact these sequences.

Table 5 EPRI Accident Class Definitions Accident Classes Description (Containment Release Type) 1 No Containment Failure 2 Large Isolation Failures (Failure to Close) 3a Small Isolation Failures (Liner Breach) 3b Large Isolation Failures (Liner Breach) 4 Small Isolation Failures (Failure to Seal - Type B)

Revision 3 Page 18 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 EPRI Accident Class Definitions Accident Classes Description (Containment Release Type) 5 Small Isolation Failures (Failure to Seal - Type C) 6 Other Isolation Failures (e.g., Dependent Failures) 7 Failures Induced by Phenomena (Early and Late) 8 Bypass (Interfacing System LOCA)

CDF All CET End States (Including Very Low and No Release)

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

Step 1 - Quantify the baseline risk in terms of frequency per reactor year for each of the accident classes presented in Table 5-9.

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

Step 3 - Evaluate risk impact of extending Type A test interval from 3 in 10 years to 1 in 15 years and 1 in 10 years to 1 in 15 years.

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

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

5.2.1 Step 1 - Quantify the Baseline Risk in Terms of Frequency per Reactor Year As previously described, the extension of the Type A interval does not influence those accident progressions that involve large containment isolation failures, Type B or Type C testing, or containment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks is included in the model. (These events are represented by the Class 3 sequences in EPRI TR-104285 [Reference 2].) The question on containment integrity was modified to include the probability of a liner breach or bellows failure (due to excessive leakage) at the time of core damage. Two failure modes were considered for the Class 3 sequences. These are Class 3a (small breach) and Class 3b (large breach).

The frequencies for the severe accident classes defined in Table 5-9 were developed for CCNPP by first determining the frequencies for Classes 1, 2, 6, 7, and 8. Table 5-10 provides a correlation of the adjusted release category frequencies and the EPRI release classes in Table 5-9. Table 5-10 provides the CCNPP-specific frequencies for each Level 2 release category.

Table 5-11 presents the grouping of each endstate in EPRI Classes based on the associated description. Table 5-12 presents the LERF sequence description, frequency and EPRI category for each sequence and the totals of each EPRI classification. Table 5-13 provides a summary of the accident sequence frequencies that can lead to radionuclide release to the public and have been derived consistent with the definitions of accident classes defined in EPRI TR-104285

[Reference 2], the NEI Interim Guidance [Reference 3], and guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24]. Adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion of the steel liner per the methodology described in Section 5.1.4. Note: calculations were performed with more digits than shown in this section. Therefore, minor differences may occur if the calculations in this sections are followed explicitly.

Revision 3 Page 19 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (e.g., containment liner) exists that can only be detected by performing a Type A ILRT. The probability of leakage detectable by a Type A ILRT is calculated to determine the impact of extending the testing interval. The Class 3 calculation is divided into two classes: Class 3a is defined as a small liner breach (La < leakage

< 10La), and Class 3b is defined as a large liner breach (10La < leakage < 100La).

Data reported in EPRI 1009325, Revision 2-A [Reference 24] states that two events could have been detected only during the performance of an ILRT and thus impact risk due to change in ILRT frequency. There were a total of 217 successful ILRTs during this data collection period.

Therefore, the probability of leakage is determined for Class 3a as shown in the following equation:

2 Pclass3 a -- = 0.0092 217 Multiplying the CDF by the probability of a Class 3a leak yields the Class 3a frequency contribution in accordance with guidance provided in Reference 24. As described in Section 5.1.3, additional consideration is made to not apply failure probabilities on those cases that are already LERF scenarios (i.e., the Class 2 and Class 8 contributions). Therefore, these LERF contributions from CDF are removed. Therefore, the frequency of a Class 3a failure is calculated by the following equation:

Frequiclass3 a = Pclass3 a * (CDFul - Class2u, - Class8uj) =2 *(1.61E-S - 5.01E 6.77E-7) 217 Frequlcass3a= 1.42E-7 Frequ2class3a = Pclass3 a * (CDFu2 - Class2u2 - Class8U2) 2 *(1.41E5 - 4.29E8 - 6.72E7)

Frequ2class3a= 1.23E-7 In the database of 217 ILRTs, there are zero containment leakage events that could result in a large early release. Therefore, the Jeffreys Non-Informed Prior is used to estimate a failure rate and is illustrated in the following equations:

Probability = Number of Failures+ 1/2 effreys Failure PrNumber of Tests + 1 Pctass3b -- 021+ 1/21- 0.0023 217++/-1 The frequency of a Class 3b failure is calculated by the following equation:

Frequlass3b= Pclass3b * (CDFul - Class2u, - Class8uj) = 's *(1.61E5 218

- 5.01E8 - 6.77E7)

Frequldcass3b = 3.52E-8 Frequ 2 1c ass 3 b Pclass3b * (CDFu2 - Class2u2 - Class8u2 ) = *(1.41E5 - 4.34E 6.72E7) 218 Frequ2class3b = 3.07E-8 For this analysis, the associated containment leakage for Class 3a is 1 OLa and for Class 3b is 100L,. These assignments are consistent with the guidance provided in Reference 24.

Class 1 Sequences. This group consists of all core damage accident progression bins for which the containment remains intact (modeled as Technical Specification Leakage). The frequency per year is initially determined from the EPRI Accident Class 1 frequency listed in Table 5-12 and then subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF), calculated below:

Revision 3 Page 20 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Frequiciass ý=Frequiciassi - (Frequlclass3a- Frequlclass3b)

Frequ2 ciassi = FreqU2classi - (FreqU2class3a - FreqU2class3b)

Class 2 Sequences. This group consists of core damage accident progression bins with large containment isolation failures. The frequency per year for these sequences is obtained from the EPRI Accident Class 2 frequency listed in Table 5-12.

Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis, consistent with approved methodology.

Class 5 Sequences. This group consists of all core damage accident progression bins for which a containment isolation failure-to-seal of Type C test components. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis, consistent with approved methodology.

Class 6 Sequences. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs.

These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution. For CCNPP, this class is defined as the SERF category. The frequency per year for these sequences is obtained from the EPRI Accident Class 6 frequency listed in Table 5-12.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (e.g., overpressure). For this analysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table 5-12.

Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass or SGTR occurs. For this analysis, the frequency is determined from the EPRI Accident Class 8 frequency listed in Table 5-12.

Table 5 Release Category Frequencies Release Category EPRI Category Unit I Frequency (/yr) Unit 2 Frequency (/yr)

INTACT Class 1 6.76E-06 5.12E-06 LERF Classes 2,7, 8 1.39E-06 1.56E-06 SERF Class 6 1.87E-06 1.25E-06 LATE Class 1 6.06E-06 6.17E-06 1 Total (CDF) N/A 1.61E-05 1.41 E-05

1. Unit 2 LATE was quantified at 5E-12 truncation. The other end states were quantified at 1 E-12.

Page 21 of 93 Revision 33 Revision Page 21 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension LERF quantification is distributed into EPRI categories based on end states. Table 5-11 shows this distribution.

Table 5 Release Category Frequencies CCNPP LERF Description of Outcome EPRI Unit I Unit 2 End State Category Frequency (/yr) Frequency (lyr)

LERF01 Containment pressure (HP)failure vesselfollowing breach high-(VB) 7 3.66E-07 3.80E-07 LERF02 Containment failure following HP VB 7 4.87E-10 5.45E-08 LERF03 Containment failure following low pressure 7 4.25E-1 1 1.89E-07 (LP) VB LERF04 Temperature induced (TI) SGTR 8 0.OOE+00 3.17E-09 LERF05 Containment failure following LP VB 7 5.04E-08 9.61 E-08 LERF06 Pressure induced (PI) SGTR 8 0.00E+00 0.OOE+00 LERF07 Containment failure following LP VB 7 1.09E-08 9.68E-09 LERF08 Loss of isolation 2 3.34E-08 3.72E-08 LERF09 Containment bypass 8 6.68E-07 6.56E-07 LERF10 Containment failure following LP VB 7 1.37E-07 5.59E-08 LERF11 Containment failure following HP VB 7 2.01 E-08 1.40E-08 LERF12 Containment failure following LP VB 7 5.14E-08 2.85E-08 LERF13 TI-SGTR 8 8.75E-09 1.21 E-08 LERF14 Containment failure following LP VB 7 1.27E-08 7.63E-09 LERF15 PI-SGTR 8 0.OOE+00 0.OOE+00 LERF16 Containment failure following LP VB 7 0.OOE+00 0.OOE+00 LERF17 Loss of isolation 2 3.05E-08 1.71 E-08 LERF18 Containment bvyass 8 4.94E-10 5.23E-10 Contribution to EPRI Classification 2 6.39E-08 5.43E-08 Contribution to EPRI Classification 7 6.49E-07 8.35E-07 Contribution to EPRI Classification 8 6.77E-07 6.72E-07 Total LERF 11.39E-06 1.56E-06 Table 5 Release Category Frequencies Release Category EPRI Category Unit I Frequency (/yr) Unit 2 Frequency (/yr)

INTACT + LATE 1 Class 1 1.28E-05 6 1.13E-05 6 LERF 2 Class 2 5.01E-08 6 4.34E-08 6 SERF 3 Class 6 1.87E-06 1.25E-06 4

LERF Class 7 6.49E-07 8.35E-07 5

LERF Class 8 6.77E-07 6.72E-07 Total (CDF) 1.61E-5 1.41E-5

1. The EPRI Class 1 category consists of INTACT and LATE failures. A LATE failure is classified as intact due to the long time until failure and is consistent with guidance in Reference 24.

2. The EPRI Class 2 category consists of CCNPP assigned LERF contribution associated with isolation failures as re-categorized in Table 5-11 with pre-event containment liner failure removed (see note 6).

3. The EPRI Class 6 category consists of CCNPP assigned scrubbed isolation failures in SERF.

4. The EPRI Class 7 category consists of the CCNPP assigned LERF contribution associated with phenomenological failures as re-categorized in Table 5-11.

5. The EPRI Class 8 category consists of the CCNPP assigned LERF contribution associated with bypass or SGTR failures as re-categorized in Table 5-11.

6. The level 2 model contains a bounding contribution associated with pre-event containment liner failure. To preclude influencing the current detailed assessment, the contribution associated with this failure is adjusted by removal of the bounding estimate from Class 2 and adding it to the intact containment case (Class 1).

The Unit 1 pre-event containment liner failure value is 1.385E-8; the Unit 2 value is 1.094E-8. These values are the LERF contributions from events FAILLEAK and FAILLEAK_2 for Units 1 and 2, respectively.

Revision 3 Page 22 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Baseline Risk Profile Class Descripticrn Unit I Frequency (/yr) Unit 2 Frequency (lyr) 2 2 1 No containment failure 1.27E-05 1.12E-05 2 Large containment iso lation failures 5.01 E-08 4.34E-08 3a Small isolation failures (liner breach) 1.42E-07 1.23E-07 3b Large isolation failures (liner breach) 3.52E-08 3.07E-08 4 Small isolation failures - fail ure to seal (type B) El E1 5 Small isolation failures - failure to seal (type C) El El 6 Containment isolation failures (dependent failure, 1.87E-06 1.25E-06 personnel errors) 7 Severe accident phenomena induced failure (early 6.49E-07 8.35E-07 and late) 8 Containment bypass 6.77E-07 6.72E-07 Total 1.61 E-05 1.41E-05

1. E represents a probabilistically insignificant value.

2. The Class 3a and 3b frequencies are subtracted from Class 1 to preserve total CDF.

5.2.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose)

Plant-specific release analyses were performed to estimate the person-rem doses to the population within a 50-mile radius from the plant. The releases are based on CCNPP-specific dose calculations summarized on Table 5-5. Table 5-14 provides a correlation of CCNPP population dose to EPRI Accident Class. Table 5-15 provides population dose for each EPRI accident class.

The population dose for EPRI Accident Classes 3a and 3b were calculated based on the guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24] as follows:

EPRI Class 3a PopulationDose = 10

• 3.20E+4 = 3.20E+5 EPRI Class 3b PopulationDose = 100

• 3.20E+4 = 3.20E+5 Table 5 Mapping of Population Dose to EPRI Accident Class Release EPRI Unit I Frequency Unit I Dose Unit 2 Frequency Unit 2 Dose Category Category (/yr) (person-rem) (/yr) (person-rem)

INTACT + Class 1 1.28E-05 3.20E+04 1.13E-05 3.20E+04 LATE LERF Class 2 5.01 E-08 2.OOE+07 4.34E-08 2.OOE+07 SERF Class 6 1.87E-06 7.01E+06 1.25E-06 7.01 E+06 LERF Class 7 6.49E-07 5.61E+07 8.35E-07 5.61E+07 LERF Class 8 6.77E-07 2.25E+07 6.72E-07 2.25E+07 Page 23 of 93 Revision 3 Page 23 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Baseline Population Doses Class Description Population Dose (person-rem) 1 No containment failure 3.20E+04 2 Large containment isolation failures 2.OOE+07 3a Small isolation failures (liner breach) 3.20E+05 1 3b Large isolation failures (liner breach) 3.20E+06 2 4 Small isolation failures - failure to seal (type B) N/A 5 Small isolation failures - failure to seal (type C) N/A 6 Containment isolation failures (dependent failure, personnel errors) 7.01 E+06 7 Severe accident phenomena induced failure (early and late) 5.61 E+07 8 Containment bypass 2.25E+07

1. 10*La

2. 100*La 5.2.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Years The next step is to evaluate the risk impact of extending the test interval from its current 10-year interval to a 15-year interval. To do this, an evaluation must first be made of the risk associated with the 10-year interval, since the base case applies to 3-year interval (i.e., a simplified representation of a 3-to-10 interval).

Risk Impact Due to 10-Year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and Class 3b sequences is impacted. The risk contribution is changed based on the NEI guidance as described in Section 5.1.3 by a factor of 10/3 compared to the base case values. The Class 3a and 3b frequencies are calculated as follows:

10 2 I Frequclass3alOyr 10 3

• 2 217

• (CDFu1 - Class2u, - Class8uj) = 3 217

• 1.54E-5 = 4.72E-7 10 21 1.34E-5 = 4.11E-7 C2ass8U2 ) == 2

• 10 22 10

• Frequ2class3alOyr * * (CDFU2 - Class2U2 -

3217 3 217 10 .5.

Frequlclass3blOyr = 10 3

0*

• 218 (CDFul - Class2u, - Class8uj) =

3 218

• 1.54E5 = 1.17E7 FreqU20ass~bl~r F~qucas~l~r = 10 3 218__* (CDFU2 - Class2U2 - Class8u2) = 1* 3 218

• 1.34E5 = 1.02E-7 The results of the calculation for a 10-year interval for Units 1 and 2 interval are presented in Tables 5-16 and 5-17, respectively.

Risk Impact Due to 15-Year Test Interval The risk contribution for a 15-year interval is calculated in a manner similar to the 10-year interval. The difference is in the increase in probability of leakage in Classes 3a and 3b. For this case, the value used in the analysis is a factor of 5 compared to the 3-year interval value, as described in Section 5.1.3. The Class 3a and 3b frequencies are calculated as follows:

Frequ1c5 i 22 2 F--ass3alSyr = -3

• 217 - * (CDFul - Class2u, - Class8uj) = 5 * -

• 1.54E5 = 7.08E7 217 Revision 3 Page 24 of 93

IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 15 2-- 2 F~q~lssa5r FreqU2Cass3alSyr 3 is 217 2 * (CDFu2 - Class2u2 - Class8u2 ) = 5 *2217

• 1.34E-5 = 6.17E-7 15 5 5 Frequlcass3blyr * * (CDFul - Class2u, - Class8uj) = 5 *

• 1.54E-5 = 1.76E-7 3 218 218 1s 5 5 FreqU2Cass3blSyr 3 3218

• 2' * (CDFu2 - Class2u2 - Class8u2 ) = 5

• 2 218

• 1.34E-5 = 1.53E-7 The results of the calculation for a 15-year interval for Units 1 and 2 are presented in Table 5-18 and 5-19.

Table 5 Unit I Risk Profile for Once in 10 Year ILRT Class Description Frequency Contribution (%) Population Population

(/yr) Dose Dose Rate (person- (person-rem) rem/yr) 1 No containment failure 1 1.23E-05 76.17% 3.20E+04 3.92E-01 2 Large containment isolation failures 5.01E-08 0.31% 2.00E+07 1.00E+00 3a Small isolation failures (liner 4.72E-07 2.93% 3.20E+05 1.51 E-01 breach) 3b Large isolation failures (liner 1.17E-07 0.73% 3.20E+06 3.76E-01 breach) 4 Small isolation failures - failure to E1 E El E seal (type B) 5 Small isolation failures - failure to E1 E1 El E seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1.87E-06 11.61% 7.01 E+06 1.31 E+01 errors) 7 Severe accident phenomena 6.49E-07 4.04% 5.61 E+07 3.64E+01 induced failure (early and late) 8 Containment bypass 6.77E-07 4.21% 2.25E+07 1.52E+01 Total 1.61 E-05 6.67E+01

1. £ represents a probabilistically insignificant value.

2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

Table 5 Unit 2 Risk Profile for Once in 10 Year ILRT Class Description Frequency Contribution (%) Population Population

(/yr) Dose Dose Rate (person- (person-rem) rem/yr) 1 No containment failure' 1.08E-05 76.51% 3.20E+04 3.45E-01 2 Large containment isolation failures 4.34E-08 0.31% 2.00E+07 8.67E-01 3a Small isolation failures (liner breach) 4.11 E-07 2.92% 3.20E+05 1.32E-01 3b Large isolation failures (liner breach) 1.02E-07 0.73% 3.20E+06 3.28E-01 Small isolation failures - failure to seal (type B)

Revision 3 Page 25 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Unit 2 Risk Profile for Once in 10 Year ILRT Class Description Frequency Contribution (%) Population Population

(/yr) Dose Dose Rate (person- (person-rem) rem/yr)

Small isolation failures - failure to seal (type C) 6 Containment isolation failures 1.25E-06 8.85% 7.01E+06 8.75E+00 (dependent failure, personnel errors) 7 Severe accident phenomena 8.35E-07 5.92% 5.61 E+07 4.69E+01 induced failure (early and late) 8 Containment bypass 6.72E-07 4.76% 2.25E+07 1.51 E+01 Total 1.41 E-05 7.24E+01

1. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

Table 5 Unit I Risk Profile for Once in 15 Year ILRT Class Description Frequency (/yr) Contribution (%) Population Population Dose Dose Rate (person- (person-rem) rem/yr) 1 No containment failure1 1.20E-05 74.34% 3.20E+04 3.83E-01 2 Large containment isolation 5.01E-08 0.31% 2.00E+07 1.OOE+00 failures 3a Small isolation failures (liner 7.08E-07 4.40% 3.20E+05 2.26E-01 breach) 3b Large isolation failures (liner 1.76E-07 1.09% 3.20E+06 5.64E-01 breach) 4 Small isolation failures - failure E1 E1 El E to seal (type B) 5 Small isolation failures - failure El El El El to seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1.87E-06 11.61% 7.01 E+06 1.31 E+01 errors) 7 Severe accident phenomena 6.49E-07 4.04% 5.61 E+07 3.64E+01 induced failure (early and late) 8 Containment bypass 6.77E-07 4.21% 2.25E+07 1.52E+01 Total 1.61 E-05 6.69E+01

1. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

Table 5 Unit 2 Risk Profile for Once in 15 Year ILRT Class Description Frequency (Iyr) Contribution (%) Population Population Dose Dose Rate (person- (person-rem) rem/yr) 1 No containment failure' 1.05E-05 74.69% 3.20E+04 3.37E-01 Revision 3 Page 26 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Unit 2 Risk Profile for Once in 15 Year ILRT Class Description Frequency (/yr) Contribution (%) Population Population Dose Dose Rate (person- (person-rem) rem/yr) 2 Large containment isolation 4.34E-08 0.31% 2.OOE+07 8.67E-01 failures 3a Small isolation failures (liner 6.17E-07 4.37% 3.20E+05 1.97E-01 breach) 3b Large isolation failures (liner 1.54E-07 1.09% 3.20E+06 4.91E-01 breach) 4 Small isolation failures - failure F1 El El E to seal (type B) 5 Small isolation failures - failure El El El El to seal (type C)

Containment isolation failures 6 (dependent failure, personnel 1.25E-06 8.85% 7.01 E+06 8.75E+00 errors) 7 Severe accident phenomena 8.35E-07 5.92% 5.61E+07 4.69E+01 induced failure (early and late) 8 Containment bypass 6.72E-07 4.76% 2.25E+07 1.51 E+01 Total 1.41 E-05 7.26E+01

1. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.

5.2.4 Step 4 - Determine the Change in Risk in Terms of LERF The risk increase associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from an intact containment could, in fact, result in a larger release due to the increase in probability of failure to detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.

Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 [Reference 4] defines very small changes in risk as resulting in increases of CDF less than 10-6/year and increases in LERF less than 10-7/year, and small changes in LERF as less than 10-6/year. Since containment overpressure is not required in support of ECCS performance to mitigate design basis accidents at CCNPP, the ILRT extension does not impact CDF. Therefore, the relevant risk-impact metric is LERF.

For CCNPP, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a 10-year test interval from Tables 5-16 and 5-17, the Class 3b frequency is 1.17E-7/year for Unit 1 and 1.02E-7 for Unit 2; based on a 15-year test interval from Tables 5-18 and 5-19, the Class 3b frequency is 1.76E-7 for Unit 1 and 1.54E-7 for Unit 2. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is 1.41 E-7/year for Unit 1 and 1.23E-7 for Unit 2. Similarly, the increase due to increasing the interval from 10 to 15 years is 5.87E-8/year for Unit 1 and 5.12E-8 for Unit 2. As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology), the Revision 3 Page 27 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension estimated change in LERF is below the threshold criteria for a small change when comparing the 15-year results to the current 10-year requirement, and slightly greater than the criteria when compared to the original 3-year requirement. Table 5-20 summarizes these results.

Table 5 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Unit 1:3 Unit 1:10 Unit 1:15 Unit 2:3 Unit 2:10 Unit 2:15 Interval Years Years Years Years Years Years (baseline) (baseline)

Class 3b (Type A 3.52E-08 1.17E-07 1.76E-07 3.07E-08 1.02E-07 1.54E-07 LERF)

ALERF (3 year baseli(ea , 8.22E-08 1.41E-07 7.16E-08 1.23E-07 baseline)

ALERF (10 year 5.87E-08 5.12E-08 baseline)

The increase in the overall probability of LERF due to Class 3b sequences being slightly greater than 1E-7 is not unexpected. Since the target is exceeded, some refinement is necessary. One method to remove some conservatism is to examine the source term expected to be available for release during the accident sequence. The source term is greatly reduced if the debris expelled from the reactor remains covered with water. Therefore, ifthe accident sequence contains containment spray success, the source term is not considered to lead to a large early release. The methodology developed in Reference 33 is used for this containment spray success sensitivity. Excluding INTACT scenarios where containment spray is credited and therefore scrubbing the source term release results in a frequency reduction.

Conservatisms are further reduced by analyzing the source term release times. Early release timing is defined by time short enough that ability to evacuate nearby population is impaired such that a fatality is possible. For this assessment, an early release is defined as occurring before 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. By reviewing CCNPP's MAAP runs in the Level 2 severe accident report

[Reference 18], it was determined three cases had source terms released after the 6.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> mark. The first case is HRIF, which simulates a loss of main feedwater due to a station blackout (SBO). The last two cases, GIOY and MRIF, evaluate small LOCAs inside containment. These three MAAP cases are matched with a corresponding plant damage state (PDS) in CCNPP's Level 2 notebook [Reference 32]. Table 5-21 displays CCNPP's PDSs.

Table 5 Summary of CCNPP Plant Damage States PDS Containment RCS Pressure at Feedwater Pressurizer CHR? AC Power Bypass? Time of Core Availability? PORV/SRV Status? Available?

Damage?

1 No High Not Not stuck open Not Available available Available Not 4 No Low Available Not stuck open Available Available Not 5 No High Available Not stuck open Available Available 6 No Low Available Not stuck open Available Available Revision 3 Page 28 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Summary of CCNPP Plant Damage States PDS Containment RCS Pressure at Feedwater Pressurizer CHR? AC Power Bypass? Time of Core Availability? PORV/SRV Status? Available?

Damage?

7 SGTR N/A N/A N/A N/A N/A 8 ISLOCA N/A N/A N/A N/A N/A Not Not 9 No High Available Not stuck open Available Available 10 NoHighNot 10 No High Available Not stuck open Available Available Not Not Not 14 No High Available Stuck open Available Available ot NNot Not 15 No High Available Not stuck open Available Available Not Not 16 No High Available Stuck open Available Available Not Not 17 No Low Available Not Stuck Open Available Available The HRIF MAAP case models a SBO that leads to a loss of main feedwater. The analysis assumes a loss of containment heat removal and AC power. The reactor coolant system is isolated and the containment remains intact. Core damage occurs while the reactor coolant system is at high pressure. Based on the information in Table 5-21, this case can be used to represent PDS 15. Table 2-2 of Reference 32 contains the list of all the Level 1 core damage accident sequences and how each is mapped to a PDS. Using the correlation of the HRIF case and the SBO cases that contain a loss of feedwater (PDS 15), it is determined that the frequency contribution can be removed from the LERF contribution because the release time of the source terms is greater than 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> [Reference 18]. The impacted sequences are SBO004, SBO005, SBOO10, SBO013, SBOO15, SBOO18, SBOO19, and SB0039.

Another MAAP case evaluated is GIOY, which involves a Small LOCA inside containment with an equivalent break size of 0.005 ft2 and the containment isolated. The reactor coolant system is at high pressure with auxiliary feedwater (AFW) and AC power available; containment air cooling (CAC) provides containment cooling and maintains containment pressure. The radionuclide release occurs after 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> [Reference 18]. Therefore, this case can be used to represent PDS 5. The following sequences are Small LOCA cases assigned to PDS 5 and are excluded based on their late release: SLOCA002, SLOCA003, and SLOCA012.

Another MAAP case evaluated is MRIF, which involves a Small LOCA inside containment with an equivalent break size of 0.02 ft2 . Containment is isolated; the reactor coolant pressure is high; AFW and containment heat removal are not available; AC power is available. These characteristics map to PDS 1. The following sequences are Small LOCA cases assigned to PDS 5 and are removed from the PDS 1 frequency: TRAN003, TRAN004, TRAN005, TRAN007, TRAN008, TRAN009, SLOCA007, and SLOCA01 1 [Reference 18].

Revision 3 Page 29 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension The exclusion of these frequencies yields new Level 2 results. Table 5-22 shows adjusted release category frequencies after some conservatisms from containment spray success and release timing are excluded.

Table 5 Adjusted Release Category Frequencies Release Category EPRI Category Unit I Frequency (/yr) Unit 2 Frequency (/yr)

INTACT Class 1 3.28E-06 1.18E-06 LERF Classes 2, 7, 8 9.51 E-07 1.08E-06 SERF Class 6 1.54E-06 5.74E-07 LATE Class 1 2.37E-06 2.18E-06 1 Total (CDF) N/A 8.14E-06 5.O1E-06

1. Unit 2 LATE was quantified at 5E-12 truncation. The other end states were quantified at 1 E-12.

Substituting these values into the previously defined equations and calculation method yields the final results displayed in Table 5-23.

Table 5 Impact on LERF due to Extended Type A Testing Intervals with Adjusted CDF ILRT Inspection Unit 1: 3 Unit 1:10 Unit 1:15 Unit 2: 3 Unit 2:10 Unit 2:15 Interval Years Years Years Years Years Years (baseline) (baseline)

Class 3b (Type A 1.14E-08 3.78E-08 5.68E-08 6.14E-09 2.05E-08 3.07E-08 LERF)

ALERF (3 year baseline) 2.65E-08 4.54E-08 1.43E-08 2.46E-08 ALERF (10 year baseline) 1.89E-08 1.02E-08 The adjusted containment spray and PDS inputs allow the Unit 1 and 2 values to be much less than the 1 E-7 LERF metric. The delta LERF between the 3 years and the 15 years is 4.54E-8/yr for Unit 1 and 2.46E-8/yr for Unit 2. These values show that the proposed extension meets the definition of a very small change in risk as defined in Regulatory Guide 1.174.

5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability (CCFP)

Another parameter that the NRC guidance in RG 1.174 [Reference 4] states can provide input into the decision-making process is the change in the conditional containment failure probability (CCFP). The CCFP is defined as the probability of containment failure given the occurrence of an accident. This probability can be expressed using the following equation:

CCFP = 1 - f(ncf)

CDF where f(ncf) is the frequency of those sequences that do not result in containment failure; this frequency is determined by summing the Class 1 and Class 3a results Since CCFP is only concerned with a containment failure and not whether the release is small or large, the Class 1 results without refinement must be used to calculate the CCFP. Table 5-24 Revision 3 Page 30 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension shows the steps and results of this calculation. The difference in CCFP between the 3-year test interval and 15-year test interval is 0.88% for Unit 1 and 0.87% for Unit 2.

Table 5 Impact on CCFP due to Extended Type A Testing Intervals ILRT Inspection Unit 1: 3 Unit 1:10 Unit 1:15 Unit 2: 3 Unit 2:10 Unit 2:15 Interval Years Years Years Years Years Years (baseline) (baseline) f(ncf) (/yr) 1.28E-05 1.27E-05 1.27E-05 1.13E-05 1.12E-05 1.12E-05 f(ncf)/CDF 0.796 0.791 0.787 0.799 0.794 0.791 CCFP 0.204 0.209 0.213 0.201 0.206 0.209 ACCFP (3 year baseline) 1 0.511% 0.876% 0.508% 0.871%

ACCFP (10 year 0.363%

baseline) 0.365%

As stated in Section 2.0, a change in the CCFP of up to 1.5% is assumed to be small. The increase in the CCFP from the 3 in 10 year interval to 1 in 15 year interval is 0.876% for Unit 1 and 0.871% for Unit 2. Therefore, this increase is judged to be very small.

5.3 Sensitivities 5.3.1 Potential Impact from External Events Contribution An assessment of the impact of external events is performed. The primary basis for this investigation is the determination of the total LERF following an increase in the ILRT testing interval from 3 in 10 years to 1 in 15 years.

Calvert Cliffs is transitioning to NFPA 805 licensing basis for fire protection and submitted a License Amendment Request (LAR) on September 24, 2013 (ADAMS Accession No. ML13301A673). This transition included performing a Fire PRA and committing to modifications to reduce the fire-induced core damage and large early release frequencies to those reported in the NFPA 805 LAR. Compensatory actions have been implemented to reduce the fire risk until the modifications are implemented. The Unit 1 ILRT is scheduled for 2016, which is prior to the scheduled implementation of all the modifications by 2018. It is anticipated that many, but not all, of the NFPA 805 modifications will be completed by the Unit 1 refueling outage. Risk mitigation strategies will be in place for any open modification. These strategies may be actions to reduce fire initiating event probabilities, actions to improve suppression probability, and/or actions to recover or protect systems that mitigate core damage and large early release accident sequences. The Unit 2 ILRT is scheduled for 2023, so the NFPA 805 modifications will be implemented prior to the extension. The section evaluates the fire risk using the Fire PRA.

Section 5.3.1.1 uses the IPEEE fire risk values to evaluate fire risk.

The Fire PRA model 6.1M was used to obtain the fire CDF and LERF values. To reduce conservatism in the model, the plant damage state methodology described in Section 5.2.4 was also applied to the CDF portion of the Fire PRA model. The following shows the calculation for Class 3b for Units 1 and 2:

0.5 Frequlciass3bFreulcassb Plas~b

=ciass3b * (CF1 D~c~z)218 (CDFl - PDSCDFl) = *-(3.18E-05 - 2.20E-5) = 2.25E-8 0.5

... cass3b * (CDF2 - PDSCDF2 ) = * (3.62E 2.26E-5) = 3.12E-8 Revision 3 Page 31 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Freqciass3 blOyr 3

• 10 Pclassi * (CD F1 - PDSCDF-) = 13

• 218 * (3.18E 2.20E-5) = 7.49E-8 10 10 0.5 Freqclass3 bloyr 3

• Pclass3b

• 10 (CDF2 - PDScDF2 ) = 10 3

• 0-

• 218 (3.62E 2.26E-5) = 1.04E-07 150.

Freqclass3 blSyr is

• Pclass3b * (CDF1 - PDScDFl) = 5 *0. * (3.18E 2.20E-5) = 1.12E-07 3 218 Freqcass~b~yr =150.

Freqcls 3 blSyr is

• Pclass3b * (CDF2 - PDScDF2 ) = 5

• 0- * (3.62E 2.26E-5) = 1.56E-07 3 218 Seismic events were addressed through a simplified seismic PRA in Section 3 of the IPEEE for CCNPP [Reference 35]. The Seismic PRA method screened all the components that met a high confidence low probability of failure (HCLPF) for the review level seismic event occurring with a magnitude of 0.3g. The remaining components were grouped together as a proxy component. It was assumed that if this proxy component failed it would result in core damage. This method is considered conservative.

Table 5-25 shows data from Table 3-6 of the IPEEE Seismic Analysis [Reference 35].

Table 5 Seismic Contribution to Frequencies of Containment Failure Categories Containment Failure Category Associated Seismic CDF (/yr)

I. Intact Containment 4.62E-07 I1.Late Containment Failure 8.63E-06 I1l. Early Small Containment Failure 1.70E-07 to 1.27E-06 IV.Early Large Containment Failure 3.13E-07 to 1.41 E-06 V. Small Containment Bypass 0 VI. Large Containment Bypass 0 Total 1.07E-05 Note: The Seismic contribution to Containment failure categories III and IV is shown as a range of values. A range is shown because the contribution of a certain PDS will be apportioned between the small and large early containment failures, but the ratio is unknown. Therefore, we show a range of values which reflect the contribution of this PDS from being attributed entirely to early-large containment failures (conservative) to early-small containment failures. See section 3.1.6.1 of the IPEEE Seismic Analysis for a more detailed explanation.

Using this seismic data, the Class 3b frequency can be calculated by the following formulas:

Freqcass3b = Pclass3b * (CDF - CatIV - CatVI) = 0 * (1.07E 3.13E 0) = 2.38E-8 10 10 0.5 Freqclass3 bloyr 1" Pclass3b * (CDF - CatIV - CatVI) = 10

• 0 *(1.07E 3.13E-6-0)=7.92E-8 15 Freqclass3blsyr = is

• PcLass3b3 * (CDF - CatIV - CatVI) = 5

• 218 52°*(1.07E 3.13E-6-0)= 1.19E-7 CNNPP topographical location presents the opportunity for high wind events. These events include tornadoes, thunderstorms, freezing precipitation, and hurricanes. Hurricanes pose approximately one threat per year and one significant threat per 10 years (Reference 24). These natural disasters are modeled in the internal events model. As shown in Table 5-2 and 5-4 show that high wind risk is approximately two orders of magnitude lower than fire risk. Since high wind risk is already included in the internal events PRA, no further analysis is necessary to include its contribution to Class 3b frequency.

The seismic and fire contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The change in LERF is calculated for the 1 Revision 3 Page 32 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension in 10 year and 1 in 15 year cases and the change defined for the external events in Tables 5-26 and 5-27 for Units 1 and 2, respectively.

Table 5 CCNPP Unit I External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year 1 per 15 years per 15 years)

External Events 4.62E-08 1.54E-07 2.31E-07 1.85E-07 Internal Events 1.14E-08 3.78E-08 5.68E-08 4.54E-08 Combined 5.76E-08 1.92E-07 2.88E-07 2.30E-07 Table 5 CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year I per 15 years per 15 years)

External Events 5.50E-08 1.83E-07 2.75E-07 2.20E-07 Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08 Combined 6.11E-08 2.04E-07 3.05E-07 2.44E-07 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined the total change in Unit 1 and 2 LERF meet the guidance for small change in risk, as it exceeds the 1.OE-7/yr and remains less than 1.OE-6 change in LERF for both units. For this change in LERF to be acceptable, total LERF must be less than 1.OE-5.

Conservatively using the highest seismic LERF value and not crediting containment spray success or plant damage state adjustments for the Internal Events or Fire PRA, the total LERF values are calculated below:

Unit 1: LERFui = LERFuiinternal + LERFuiseismic + LERFulfire + LERFulclass3Bincrease

= 1.39E-6/yr + 1.41E-6/yr + 2.97E-6/yr + 2.30E-07/yr = 6.OOE-6/yr Unit 2: LERFU2 = LERFu2internal + LERFu2seismic + LERFU2fire + LERFu2class3Bmincrease

= 1.56E-6/yr + 1.41E-6/yr + 4.17E-6/yr + 2.44E-07 = 7.38E-6/yr Since the total LERF for both units is less than 1.OE-5, it is acceptable for the ALERF to be between 1.OE-7 and 1.OE-6.

5.3.1.1 Potential Impact from External Events Contribution Using IPEEE Fire Analysis An assessment of the impact of external events is also performed using fire risk analysis from the IPEEE [Reference 35] rather than the Fire PRA model 6.1 M. Table 4-7 from the simplified IPEEE fire PRA shows the frequencies of major containment failure categories for Unit 1

[Reference 35]. The same containment failure category percentages are assumed for Unit 2; as given in Section 4.6.8.3 of the IPEEE, the estimated Unit 2 fire CDF is 1.1E-5/yr. The Level 2 results are shown here in table 5-28.

Page 33 of 93 Revision 33 Page 33 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of permanent ILRT Extension I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent IIRT Extension Table 5 Fire Contribution to Frequencies of Containment Failure Categories Containment Failure Category Percentage Unit 1 Fire CDF (/yr) Unit 2 Fire CDF (/yr)

I. Intact Containment 37.1% 2.67E-05 4.08E-06 I1.Late Containment Failure 56.5% 4.07E-05 6.22E-06 Ill. Early Small Containment Failure 1.7% 1.21 E-06 1.87E-07 IV. Early Large Containment Failure 6.5% 4.67E-06 7.15E-07 V. Small Containment Bypass 0.0% 0.00E+00 0.00E+00 VI. Large Containment Bypass 0.0% 0.00E+00 0.00E+00 Total 7.32E-05 1.10E-05 Using the IPEEE fire data, the Class 3b frequency can be calculated by the following formulas:

Unit 1: Frequ1d1ss3b = Pc1ass3b * (CDF - CatIV - CatVI) = 0-5 * (7.32E 4.67E 0)= 1.57E-7 218 10 Unit 1: Frequlclass3bloyr= 3

• Pclass3b * (CDF - CatIV - CatVI) 10 0.5

-- * -0 * (7.32E 4.67E 0) 5.24E-7 3 218 Unit 1: Frequlclass3blSyr is

• Pclass3b * (CDF - CatIV - CatVI)

=5 * * (7.32E 4.67E 0)= 7.86E-7 218 Unit 2: Frequ2 cjass3b = Pctass3b * (CDF - CatIV - CatVI) = .* (1.1E-5 - 7.15E-7 - 0)= 2.36E-8 218 10E-5 - - 7.1 E- 0)t.6 -

Unit 2: Frequ2class3blOyr 3

• Pciass3b * (CDF - Cat/V - CatVI) 10, * -L- * (1.10E 7.15E 0) = 7.86E-8 3 15 218 Unit 2: FreqU2class3blSyr = 1

• Pclass3b * (CDF - CatIV - CatVl)

= 0 * (1.10E 7.15E 0)= 1.18E-7 218 As done in Section 5.3.1, the IPEEE seismic and fire contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The change in LERF is calculated for the 1 in 10 year and 1 in 15 year cases and the change defined for the external events in Tables 5-29 and 5-30 for Units 1 and 2, respectively.

Table 5 CCNPP Unit I External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year 1 per 10 year I per 15 years per 15 years)

External Events 1.81 E-07 6.03E-07 9.04E-07 7.24E-07 Internal Events 1.14E-08 3.78E-08 5.68E-08 4.54E-08 Combined 1.92E-07 6.41 E-07 9.61 E-07 7.69E-07 Page 34 of 93 Revision 33 Page 34 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 3 per 10 year I per 10 year I per 15 years per 15 years)

External Events 4.73E-08 1.58E-07 2.37E-07 1.89E-07 Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08 Combined 5.35E-08 1.78E-07 2.67E-07 2.14E-07 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined the total change in Unit 1 and 2 LERF meet the guidance for small change in risk, as it exceeds the 1.OE-7/yr and remains less than 1.OE-6 change in LERF for both units. For this change in LERF to be acceptable, total LERF must be less than 1.OE-5.

Conservatively using the highest seismic LERF value and not crediting containment spray success or plant damage state adjustments for the Internal Events PRA, the total LERF values are calculated below:

Unit 1: LERFui = LERFulinternal + LERFuiseismic + LERFulfire + LERFulclass3Bincrease

= 1.39E-6/yr + 1.41E-6/yr + 4.67E-6/yr + 7.69E-07 = 8.24E-6/yr Unit 2: LERFU2 = LERFU2internaj + LERFu2seismic + LERFu2flre+ LERFu2class3Bincrease

= 1.56E-6/yr + 1.41E-6/yr + 7.15E-7/yr + 2.14E-07 = 3.90E-6/yr Since the total LERF for both units is less than 1.OE-5, it is acceptable for the ALERF to be between 1.OE-7 and 1.OE-6.

5.3.2 Potential Impact from Steel Liner Corrosion Likelihood A quantitative assessment of the contribution of steel liner corrosion likelihood impact was performed for the risk impact assessment for extended ILRT intervals. As a sensitivity run, the internal event CDF was used to calculate the Class 3b frequency. The impact on the Class 3b frequency due to increases in the ILRT surveillance interval was calculated for steel liner corrosion likelihood using the relationships described in Section 5.1.4. The EPRI Category 3b frequencies for the 3 per 10-year, 10-year and 15-year ILRT intervals were quantified using the internal events CDF. The change in the LERF risk measure due to extending the ILRT interval from 3 in 10 years to 1 in 10 years, orto 1 in 15 years is provided in Tables 5-31 and 5-32. The steel liner corrosion likelihood was increased by a factor of 1000, 10000, and 100000. Except for extreme factors of 10000 and 100000, the corrosion likelihood is relatively insensitive to the results.

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I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Unit 1 Steel Liner Corrosion Sensitivity Cases 3b 3b 3b LERF LERF LERF Frequency Frequency Frequency Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-10 to (1-per-10 to year ILRT) year ILRT) year ILRT) I-per-10) 1-per-15) 1-per-15)

Internal Event 3B 1.14E-08 3.78E-08 5.68E-08 2.65E-08 4.54E-08 1.89E-08 Contribution Corrosion Likelihood 1.15E-08 3.98E-08 6.36E-08 2.84E-08 5.22E-08 2.38E-08 X 1000 Corrosion Likelihood 1.24E-08 5.76E-08 1.25E-07 4.52E-08 1.13E-07 6.77E-08 X 10000 Corrosion Likelihood 2.17E-08 2.35E-07 7.42E-07 2.14E-07 7.20E-07 5.06E-07 X 100000 Table 5 Unit 2 Steel Liner Corrosion Sensitivity Cases 3b 3b 3b LERF LERF LERF Frequency Frequency Frequency Increase Increase Increase (3-per-10 (1-per-10 (1-per-15 (3-per-10 to (3-per-M0 to (1-per-10 to year ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

Internal Event 3B 6.14E-09 2.05E-08 3.07E-08 1.43E-08 2.46E-08 1.02E-08 Contribution Corrosion Likelihood 6.19E-09 2.15E-08 3.44E-08 1.53E-08 2.82E-08 1.29E-08 X 1000 Corrosion Likelihood 6.70E-09 3.11E-08 6.77E-08 2.44E-08 6.1OE-08 3.66E-08 X 10000 Corrosion Likelihood 1.17E-08 1.27E-07 4.01E-07 1.16E-07 3.89E-07 2.74E-07 X 100000 5.3.3 Expert Elicitation Sensitivity Another sensitivity case on the impacts of assumptions regarding pre-existing containment defect or flaw probabilities of occurrence and magnitude, or size of the flaw, is performed as described in Reference 24. In this sensitivity case, an expert elicitation was conducted to develop probabilities for pre-existing containment defects that would be detected by the ILRT only based on the historical testing data.

Using the expert knowledge, this information was extrapolated into a probability versus magnitude relationship for pre-existing containment defects. The failure mechanism analysis also used the historical ILRT data augmented with expert judgment to develop the results.

Details of the expert elicitation process and results are contained in Reference 24. The expert elicitation process has the advantage of considering the available data for small leakage events, which have occurred in the data, and extrapolate those events and probabilities of occurrence to the potential for large magnitude leakage events.

The expert elicitation results are used to develop sensitivity cases for the risk impact assessment. Employing the results requires the application of the ILRT interval methodology Revision 3 Page 36 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension using the expert elicitation to change in the probability of pre-existing leakage in the containment.

The baseline assessment uses the Jefferys non-informative prior and the expert elicitation sensitivity study uses the results of the expert elicitation. In addition, given the relationship between leakage magnitude and probability, larger leakage that is more representative of large early release frequency, can be reflected. For the purposes of this sensitivity, the same leakage magnitudes that are used in the basic methodology (i.e., 10 La for small and 100 La for large) are used here. Table 5-33 presents the magnitudes and probabilities associated with the Jefferys non-informative prior and the expert elicitation use in the base methodology and this sensitivity case.

Table 5 CCNPP Summary of ILRT Extension Using Expert Elicitation Values (from Reference 24)

Leakage Size (La) Jefferys Non-Informative Expert Elicitation Percent Reduction Prior Mean Probability of Occurrence 10 2.70E-02 3.88E-03 86%

100 2.70E-03 9.86E-04 64%

Taking the baseline analysis and using the values provided in Tables 5 5-19 for the expert elicitation yields the results in Tables 5-34 and 5-35 for Units 1 and 2, respectively, are developed.

Table 5 CCNPP Unit I Summary of ILRT Extension Using Expert Elicitation Values Accident ILRT Interval Class 3 per 10 Years 1 per 10 Years I per 15 Years Base Adjusted Dose Dose Frequency Dose Frequency Dose Frequency Base (person- Rate Rate Rate Frequency rem) (person- (person- (person-remlyr) rem/yr) rem/yr) 1 1.28E-05 1.28E-05 3.40E+02 2.70E-04 1.26E-05 2.50E-04 1.25E-05 2.36E-04 2 5.01E-08 5.01E-08 2.OOE+07 1.OOE+00 5.01E-08 1.OOE+00 5.01E-08 1.OOE+00 3a N/A 5.98E-08 3.40E+03 2.03E-04 1.99E-07 6.78E-04 2.99E-07 1.02E-03 3b N/A 1.52E-08 3.40E+04 5.17E-04 5.06E-08 1.72E-03 7.60E-08 2.58E-03 6 1.87E-06 1.87E-06 7.01E+06 1.31E+01 1.87E-06 1.31E+01 1.87E-06 1.31E+01 7 6.49E-07 6.49E-07 5.61 E+07 3.64E+01 6.49E-07 3.64E+01 6.49E-07 3.64E+01 8 6.77E-07 6.77E-07 2.25E+07 1.52E+01 6.77E-07 1.52E+01 6.77E-07 1.52E+01 Totals 1.61E-05 1.61E-05 1.06E+08 6.58E+01 1.61E-05 6.58E+01 1.61E-05 6.58E+01 ALERF (3 per 10 N/A 3.55E-08 6.08E-08 yrs base)

ALERF (1 per 10 N/A N/A 2.53E-08 yrs base)

CCFP 20.26% 20.48% 20.64%

Page 37 of 93 Revision 3 3 Page 37 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 CCNPP Unit 2 Summary of ILRT Extension Using Expert Elicitation Values Accident ILRT Interval Class 3 per 10 Years I per 10 Years I per 15 Years Base Adjusted Dose Dose Rate Frequency Dose Rate Frequency Dose Rate Frequency Base (person (person- (person- (person-Frequency -rem) rem/yr) remlyr) rem/yr) 1 1.13E-05 1.12E-05 3.40E+02 3.81E-03 1.10E-05 3.73E-03 1.08E-05 3.67E-03 2 4.34E-08 4.34E-08 2.OOE+07 8.67E-01 4.34E-08 8.67E-01 4.34E-08 8.67E-01 3a N/A 5.21 E-08 3.40E+03 1.77E-04 1.74E-07 5.90E-04 2.60E-07 8.86E-04 3b N/A 5.21 E-08 3.40E+04 1.77E-03 1.74E-07 5.90E-03 2.60E-07 8.86E-03 6 1.25E-06 1.25E-06 7.01E+06 8.75E+00 1.25E-06 8.75E+00 1.25E-06 8.75E+00 7 8.35E-07 8.35E-07 5.61E+07 4.69E+01 8.35E-07 4.69E+01 8.35E-07 4.69E+01 8 6.72E-07 6.72E-07 2.25E+07 1.51E+01 6.72E-07 1.51E+01 6.72E-07 1.51E+01 Totals 1.41E-05 1.41E-05 1.06E+08 7.16E+01 1.41E-05 7.16E+01 1.41E-05 7.16E+01 ALERF (3 per 10 N/A 1.22E-07 2.09E-07 yrs base)

ALERF (1 per 10 N/A N/A 8.69E-08 yrs base)

CCFP 20.21% 21.08% 21.69%

The results illustrate how the expert elicitation reduces the overall change in LERF and the overall results are more favorable with regard to the change in risk.

5.3.4 Large Leak Probability Sensitivity Study The large leak probability is a vital portion of determining the Class 3b frequency. CCNPP had previously calculated the large leak probability using the WCAP method. Table 5-36 presents the large leak probabilities for the baseline test, 10 year test interval, and 15 year test interval.

Table 5-37 was developed using the same process as to calculate Class 3b.

Table 5 CCNPP Large Leak Probabilities Using the WCAP Method Test Interval WCAP Large Leak EPRI Accident EPRI Accident Class 3b Probability Class 3b Frequency: Unit 2 Frequency: Unit 1 3 per 10 years 2.47E-4 1.38E-09 8.21 E-10 10 years 7.41E-4 4.05E-09 2.41 E-09 15 years 1.11E-3 5.96E-09 3.55E-09 Using the same EPRI approach, but with an updated Class 3b frequency calculated from the WCAP large leak probability data, Table 5-37 contains the final results for both units.

Page 38 of 93 Revision 33 Revision Page 38 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Impact on LERF due to Extended Type A Testing Intervals with WCAP CDF ILRT Inspection Unit 1:3 Unit 1:10 Unit 1: 15 Unit 2:3 Unit 2: 10 Unit 2:15 Interval Years Years Years Years Years Years (baseline) (baseline)

Class 3b (Type A 1.38E-09 4.05E-09 5.96E-09 8.21E-10 2.41 E-09 3.55E-09 LERF)

ALERF (3 year 2.67E-09 4.57E-09 1.59E-09 2.73E-09 baseline) -.

ALERF (10 year 1.91E-09 1.14E-09 baseline) ___

These results demonstrate that the EPRI methodology is conservative when used to calculate a large leak probability as compared to the WCAP method.

Page 39 of 93 Revision 33 Revision Page 39 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 6.0 RESULTS The results from this ILRT extension risk assessment for CCNPP are summarized in Table 6-1 for Unit 1 and Table 6-2 for Unit 2.

Table 6 Unit I ILRT Extension Summary Class Dose Base Case Extend to Extend to (person- 3 in 10 Years I in 10 Years I in 15 Years rem)

CDF/Year Person- CDF/Year Person- CDF/Year Person-Rem/Year Rem/Year Rem/Year 1 3.20E+04 5.59E-06 1.79E-01 5.46E-06 1.75E-01 5.37E-06 1.72E-01 2 2.00E+07 3.33E-08 6.66E-01 3.33E-08 6.66E-01 3.33E-08 6.66E-01 3a 3.20E+05 4.56E-08 1.46E-02 1.52E-07 4.87E-02 2.28E-07 7.30E-02 3b 3.20E+06 1.14E-08 3.63E-02 3.78E-08 1.21E-01 5.68E-08 1.82E-01 6 7.01E+06 1.54E-06 1.08E+01 1.54E-06 1.08E+01 1.54E-06 1.08E+01 7 5.61E+07 2.49E-07 1.40E+01 2.49E-07 1.40E+01 2.49E-07 1.40E+01 8 2.25E+07 6.68E-07 1.50E+01 6.68E-07 1.50E+01 6.68E-07 1.50E+01 Total 8.14E-06 4.07E+01 8.14E-06 4.08E+01 8.14E-06 4.09E+01 ILRT Dose Rate from 3a and 3b From 3 N/A 1.15E-01 1.96E-01 ATotal Years Dose Rate From 10 N/A Years YearsN/A 8.18E-02 From 3 N/A 0.282% 0.483%

%ADose Years Rate From 10 N/A Years YearsN/A 0.201%

3b Frequency (LERF)

From 3 N/A 2.65E-08 4.54E-08 Years ALERF From 10 N/A N/A 1.89E-08 Years CCFP %

From 3 N/A 0.326% 0.558%

Years ACCFP%

From 10 N/A N/A 0.233%

Years Page 40 of 93 Revision 3 Page 40 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 6 Unit 2 ILRT Extension Summary Class Dose Base Case Extend to Extend to (person- 3 in 10 Years 1 in 10 Years I in 15 Years rem)

CDF/Year Person- CDF/Year Person- CDF/Year Person-Rem/Year Rem/Year Rem/Year 1 3.20E+04 3.32E-06 1.06E-01 3.25E-06 1.04E-01 3.20E-06 1.02E-01 2 2.OOE+07 1.84E-08 3.69E-01 1.84E-08 3.69E-01 1.84E-08 3.69E-01 3a 3.20E+05 2.47E-08 7.89E-03 8.22E-08 2.63E-02 1.23E-07 3.95E-02 3b 3.20E+06 6.14E-09 1.96E-02 2.05E-08 6.55E-02 3.07E-08 9.82E-02 6 7.01E+06 5.74E-07 4.02E+00 5.74E-07 4.02E+00 5.74E-07 4.02E+00 7 5.61 E+07 4.01 E-07 2.25E+01 4.01 E-07 2.25E+01 4.01E-07 2.25E+01 8 2.25E+07 6.60E-07 1.49E+01 6.60E-07 1.49E+01 6.60E-07 1.49E+01 Total 5.01E-06 4.19E+01 5.01E-06 4.19E+01 5.01E-06 4.20E+01 ILRT Dose Rate from 3a and 3b From 3 N/A 6.19E-02 1.06E-01 ATotal Years Dose Rate From 10 N/A Years YearsN/A 4.42 E-02 From 3 N/A 0.148% 0.254%

%ADose Years Rate From 10 N/A Years YearsN/A 0.106%

3b Frequency (LERF)

From 3 N/A 1.43E-08 2.46E-08 Years ALERF From 10 N/A N/A 1.02E-08 Years CCFP %

From 3 N/A 0.286% 0.490%

Years ACCFP%

From 10 N/A N/A 0.204%

Years Revision 3 Page 41 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

7.0 CONCLUSION

S AND RECOMMENDATIONS Based on the results from Section 5.2 and the sensitivity calculations presented in Section 5.3, the following conclusions regarding the assessment of the plant risk are associated with extending the Type A ILRT test frequency to 15 years:

Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.OE-06/year and increases in LERF less than 1.OE-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 4.54E-8/year for Unit 1 and 2.46E-8/year for Unit 2 using the EPRI guidance. As such, the estimated change in LERF is determined to be "very small" for both units using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].

" The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.20 person-rem/year for Unit 1 and 0.11 person-rem/year for Unit 2. EPRI Report No. 1009325, Revision 2-A [Reference 24] states that a very small population dose is defined as an increase of < 1.0 person-rem per year, or <

1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This results of this calculation meet these criteria for both units. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

" The increase in the conditional containment failure from the 3 in 10 year interval to 1 in 15 year interval is 0.558% for Unit 1 and 0.490% for Unit 2. EPRI Report No. 1009325, Revision 2-A [Reference 24] states that increases in CCFP of < 1.5% is very small.

Therefore, this increase is judged to be very small.

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

Previous Assessments The NRC in NUREG-1493 [Reference 6] has previously concluded that:

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

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

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

Revision 3 Page 42 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension A. ATTACHMENT 1 PRA Quality Statement for Permanent 15-Year ILRT Extension The Calvert Cliffs Internal Events and Wind Model, Calvert-CAFTA-TREE-6.2a, was used for this analysis.

An independent PRA peer review was conducted under the auspices of the Pressurized Water Reactor Owners Group in June of 2010, and was performed against the guidance of Regulatory Guide 1.200, Revision 2, and requirements of American Society of Mechanical Engineers (ASME)/American National Standards (ANS) RA-Sa-2009. The scope of the review was a full-scope review of the Calvert Cliffs Nuclear Plant (Calvert Cliffs) at-power, internal initiator PRA.

Findings (generally, documentation issues or model concerns that have been evaluated as not significant using a sensitivity study) have been captured in the PRA Configuration Risk Management Program (CRMP) database. On an on-going basis, other potential PRA model and documentation changes are captured and prioritized in the CRMP database.

To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following configuration control activities are routinely performed:

" Design changes and procedure changes are reviewed for their impact on the PRA model. PRA screening is required for all design and procedure changes.

" New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

" Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated based upon reviews of plant program data, particularly data supporting the Maintenance Rule.

The Calvert Cliffs Internal Events model is also updated to support the Calvert Cliffs Fire PRA.

The Calvert Cliffs Internal Events PRA is based on a detailed model of the plant developed from the Individual Plant Examination for Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities." The model is maintained and updated in accordance with Calvert Cliffs procedures, and has been updated to meet the ASME PRA Standard and Revision 2 of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities."

The Calvert Cliffs internal events PRA model was peer reviewed in June 2010. All findings which had significant impact on this analysis have been addressed. This assessment is provided as Table 1. The ILRT application was determined to be an application requiring a Capability Category II PRA model per the Regulatory Guide 1.200 criteria, Revision 2. This is based on the requirement for numerical results for CDF and LERF to determine the risk impact of the requested change and the fact that this change is risk-informed, not risk-based. Table 1 includes discussion of all findings from the industry peer review along with the assessment and evaluation of the finding that shows that they have either been addressed or have no material impact on the ILRT interval extension request.

The peer review found that 97% of the SR's evaluated Met Capability Category IIor better.

There were 3 SRs that were noted as "not met" and eight that were noted as Category I. As noted in the peer review report, the majority of the findings were documentation related. Of the 11 SRs which did not meet Category IIor better, seven were related to conservatisms or documentation in LERF and two were related to internal floods. There were 39 findings. All findings which could be relevant to the ILRT extension evaluation were updated in the internal events model used to quantify the Level 2 release states. Thus, with the exception of minor Revision 3 Page 43 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension documentation concerns, the internal events model meets Capability Category II or causes conservative results for all SRs relevant to the ILRT extension evaluation results. No significant changes have been implemented in the internal events PRA. As there are no new methods applied, no follow on or focused peer reviews were required.

The Calvert Cliffs Fire PRA peer review was performed January 16-20, 2012 using the NEI 07-12 Fire PRA peer review process, the ASME PRA Standard (ASME/ANS RA-Sa-2009) and Regulatory Guide 1.200, Rev. 2. The purpose of this review was to establish the technical adequacy of the Fire PRA for the spectrum of potential risk-informed plant licensing applications for which the Fire PRA may be used. The 2012 Calvert Fire PRA peer review was a full-scope review of all of the technical elements of the Calvert Cliffs at-power FPRA (2012 model of record) against all technical elements in Section 4 of the ASME/ANS Combined PRA Standard, including the referenced internal events SRs. The peer review noted a number of facts and observations (F&Os). The findings and their dispositions are provided in Table 2. All findings are being provided and have been dispositioned. All F&Os that were defined as suggestions have been dispositioned and will be available for NRC review. The Fire PRA is adequate to support the ILRT extension.

The Calvert Cliffs seismic PRA model is relatively conservative and, other than the high magnitude acceleration event, is not a dominant contributor. The Calvert Cliffs high winds PRA model is very conservative in the tornado area in that all tornados are grouped into the most conservative event. PRA risk for tornadoes and high winds are based upon IPEEE values. Calvert Cliffs has maintained and updated a high wind PRA model in order to perform risk assessment of tornado missile impacts and hurricane force winds. Although this model has not been peer reviewed in compliance with the ASME/ANS RA-Sa-2009 standard, the model is based upon accepted methodology and utilizes the ASME/ANS RA-Sa-2009 compliant internal events model. High winds updates are not expected to cause a significant increase in CDF or LERF. A more detailed assessment would be expected to cause a decrease in CDF.

Page 44 of 93 Revision 33 Revision Page 44 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 1-16 AS-B3 Systems Based on Sections 2.4 and 2.10 of Complete The PRA Internal Events No impact on SY-B6 Analysis the System Analysis Introduction Accident Sequence ILRT analysis.

Notebook (CO-SY-00, Rev. 0) this Notebook, CO-AS-001, Subsequent SR appears to be met. However, Section 3.3, has been analysis has there is a potential issue related to updated with an engineering found this issue this SR. Did not find reference to analysis of this issue. The to be non-any engineering analysis needed analysis identifies that significant: 1) the to support Containment Air Cooler during the Loss of Offsite temperature rise operation when this system is Power sequences, the is not likely to assumed to be available during Containment Air Coolers are challenge the LOSP when the containment heats credited for SBO conditions containment air up prior to electrical recovery. where the containment coolers, and 2) heats up, and then, after the importance of (This F&O originated from SR SY- power recovery, the air the air coolers is B6) coolers are credited for significantly containment pressure and reduced by the temperature control. For redundant these accident sequences, function provided offsite power is restored in by containment one hour, and the spray containment pressure and corresponding saturation temperature remain well below containment design parameters that would challenge the CACs.

Furthermore, failure of CACs is not risk significant, due to the potential availability of containment spray.

REFERENCE CO-AS-001 Page 45 of 93 Revision 33 Page 45 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension 1-17 IFSO-Al Internal Examined Internal Flooding Complete An engineering analysis has Due to the QU-E3 Flooding Notebook (CO-IF-001, Rev. 1) subsequently been relatively low Sections 3.0 and 3.1. Part of the performed for AFW contribution to Internal Flood analysis may not be discharge piping flooding. CDF, this flood complete for assessing the Aux The fraction of at-power time has no impact on Feedwater Discharge Piping as a during which the AFW ILRT analysis.

Flood Source. system is in operation 0.6%

and the AFW Discharge (This F&O originated from SR Piping flood may be IFSO-A1) screened due to their low impact on CDF (<1 E-9).

REFERENCE CO-IF-001 1-18 IFSO-A4 Internal Examined Internal Flooding Open Human-induced impacts on No impact on IFEV-A7 Flooding Notebook (CO-IF-001, Rev. 1) the flood initiating event ILRT analysis.

Section 3.3 and 5.3. Consideration frequencies are not well This is a of human-induced mechanisms as documented. The issue has documentation potential flood sources not clear, been captured in the PRA issue.

Regarding human-induced impacts configuration control on the flood frequency, Section 5.3 database (CRMP), but not of the IF report states that they yet addressed.

were included, but their inclusion should be better documented or referenced from IF (e.g., a sample calculation showing human contribution would be helpful)

(This F&O originated from SR IFSO-A4) 1-19 IFEV-B3 Internal While some items are included in Open In the Internal Flood No impact on IFPP-B3 Flooding Section 7.0 of the IFreport, many notebook, the discussions ILRT analysis.

IFQU-B3 other instances of uncertainties on uncertainties and This is a Revision 3 Page 46 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension IFSN-B3 and assumptions are cited assumptions should be documentation IFSO-B3 throughout the report, but not expanded. This issue has issue.

included in the discussion of been captured in the PRA Section 7.0 nor are the configuration control implications of these other database (CRMP), but not uncertainties and assumptions are yet addressed.

discussed.

(This F&O originated from SR IFPP-B3) 1-25 DA-C7 Data For the most part actual plant- Complete The ESFAS logic train The low risk specific data is used as a basis for testing has a very low risk significance of the number of demands significance and generally ESFAS logic train associated actual plant does not take the logic OOS. testing is experiences (See basis for DA- The train does go to 2-out- considered to C6), which includes both actual of-3 logic. Occurrences have no impact planned and unplanned activities, where the train is in 2-out-of- on ILRT analysis.

However, there are a few ESFAS 3 logic is incorporated into testing and/or other logic channel the PRA Data Analysis testing that are not tracked via the Notebook, CO-DA-plant computer. 001, Section 2.6 and 3.5.

For the logic relays there is Created this F&O on non- a RAW of <1.04 and documentation of ESFAS/logic Birnbaum on the order of train testing, which needs to 4E-07. Any logic relay include actual practice. unavailability that does not cause the ESFAS channel to (This F&O originated from SR DA- be OOS and bypassed, is C7) therefore of low significance.

REFERENCE CO-DA-001 Page 47 of 93 Revision 33 Page 47 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 2-7 IFPP-A5 Internal Flood Section 2.3 provides a discussion Complete A walkdown was performed No impact on that walkdowns used to confirm to assess the susceptibility ILRT analysis.

plant arrangement. The following to jet impingement or spray This is an note is contained in section 2.3: in rooms 105A and 203. All Internal Flood equipment is considered documentation Unfortunately, the walk-down failed by spray or issue.

documentation from the original impingement for flood flooding analysis no longer exists. sources originating in the A plant walk-down was performed room. Notebook CO-IF-001 as a part of this analysis to provide was updated with this familiarity with the plant design as additional documentation.

well as confirm findings from the original walk-down. This walk- REFERENCE down is documented in a set of CO-IF-001 notes and photographs included in Appendix B.

Walkdown photos for room 105A and 203 show equipment and potential flood propagation paths.

However, there is not enough spatial information to develop specific targets for flood impingement or spray.

(This F&O originated from SR IFPP-A5) 2-9 DA-D4 Data Evidence of meeting this SR at Complete Table 2-6 of the Data No impact on CC-Il/Ill is found in the PRA Data Notebook CO-DA-001 listed ILRT analysis.

Notebook (CO-DA-001, Rev. 1) in incorrect data and Bayesian This was a Sections 2.1 and 2.7. Found update results for the documentation inconsistencies in the value of total SACMs. However, the issue. The number components of different correct values were used in Internal Events types (for both units) in Table 2-5 the models for peer review.

Revision 3 Page 48 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension of the PRA Data Notebook with the model includes actual total number for Calvert For the SACM EDGs in the correct data.

Cliffs. Also, found an inconsistency Table 2-6, the correct plant-between the prior distribution and specific data are in Table 2-posterior distribution for SACM 5. Table 2-6 lists incorrect EDG fail to start in Table 2-6 of the data and Bayesian update Data Notebook. results for the SACMs.

However, the correct values (This F&O originated from SR DA- are used in the models.

D4)

The above errors have been corrected in CO-DA-001.

Other minor typographical errors were identified and corrected in the notebook.

REFERENCE CO-DA-001 3-3 SY-C2 Systems Section 2.3 of each system Complete Marked-up system boundary No impact on Analysis notebook states that marked up drawings were generated for ILRT analysis.

plant system drawings are each system notebook. This is an provided as supplements to the Where Unit 1 and Unit 2 are Internal Events system notebook, which depicts similar, just the Unit 1 documentation the boundary of the system in boundary is depicted. In issue that has terms of PRA modeling. The addition, the system been addressed.

drawings are not in the notebooks. notebooks include drawing snippets, sketches, and (This F&O originated from SR SY- descriptive text that also C2) depict the system boundary.

REFERENCES CO-SY-[AII]

Revision 3 Page 49 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 3-5 SY-Al 1 Systems The fault tree does not include Complete A bounding sensitivity case This finding does SY-A6 Analysis potential failures of the AFW was run to include failure of not impact the accumulator system. the AFW accumulators ILRT extension.

failing short-term AFW The random (This F&O originated from SR SY- operation. This issue has an failure probability Al1) insignificant contribution to of the CDF. Short-term failure of accumulators is the AFW operation is two orders of dominated by failure of magnitude lower electrical support systems than active and failure of active hardware failures hardware (i.e. valves and that support the instrumentation). The same system applicable system function.

notebooks were updated.

REFERENCES CO-SY-036 CO-SY-019 CO-SY-000 3-8 SY-C1 Systems Several system notebooks were Complete Some new flow diversions No impact on SY-A13 Analysis reviewed (AFW, EDG, SI, 120 were identified as part of the ILRT analysis.

VAC electrical, etc.). In general, Fire PRA Multiple Spurious This is an the documentation is complete and Operation review, and these Internal Events thorough. In most cases it clearly were added to the system documentation follows the RG 1.200 SRs. models and system issue that has In some places, assumptions were notebooks. Furthermore, a been addressed.

imbedded in the documentation comprehensive review of without sufficient reference or PRA mechanical systems justification. Examples include: notebooks and drawings was performed to identify SI notebook page 11, last bullet and document potential flow

'Only one of the three HPSI pumps diversions. Flow diversion functions - For a cold leg break, it discussions were added to Revision 3 Page 50 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension is assumed that only one-fourth Sections 3.4.d of the pump discharge is spilled via the applicable system break. For a hot leg break, the notebooks.

entire pump discharge reaches the core.'

SI notebook page 12, 2nd bullet

'The maximum time assumed for operation for the safety injection pumps is 30 seconds following SIAS initiation.'

CO-SY-000 states that each system notebook addresses flow diversions (where applicable) in section 3.4.d. Although flow diversions appear to be addressed (for example, the SW notebook talks about flow diversion), there is no consistent discussion in each system notebook.

(This F&O originated from SR SY-Cl) 3-9 DA-B1 Data DA notebook table 2-5 contains Complete The model has been No impact on the grouping of components for updated to add additional ILRT analysis.

plant specific failure data. Many of component types and failure The model used the groupings appear to take into modes to better reflect for the ILRT account differences in such things service conditions. Service analysis includes as size, type, mission type (e.g., Water and Salt Water the updated data FW TDP run vs. AFW TDP pumps were broken out. and failure standby). However, in some cases, AFW pumps and Safety modes.

it is not clear what the basis for the Injection pumps were broken grouping is. For example, SW out. This resulted in changes MDP RUN and SRW MDP RUN to the associated failure are grouped together even though rates. The change has been Revision 3 Page 51 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension they are of different service reflected in the Data conditions (salt water vs. clean Notebook, CO-DA-001.

water), voltages (480 VAC vs.

4160 VAC), size, etc. Similarly, REFERENCE AFW MDP is included with HPSI CO-DA-001 MDP and LPSI MDP, even though the two SI pumps are pumping borated water, while the AFW pump is pumping condensate grade water. No documentation of the appropriateness of these groupings is provided.

(This F&O originated from SR DA-B1) 3-11 QU-B7 Quantification The mutually exclusive cutsets for Complete A comprehensive review of No impact on each system are described in the mutual exclusive modeling ILRT analysis.

system notebook section 3.4.e. was performed. Each The PRA model Several SY notebooks were system notebook and each that was updated reviewed to determine system model was reviewed as part of this appropriateness of the mutually to validate the review was used exclusive cutsets. All appeared appropriateness of the as the model for reasonable. A review was modeling and reconcile any the ILRT performed of the MUTEX gate differences, and to verify analysis.

within the fault tree model and the that a documented basis appropriate combinations identified exists for each mutually in the SY notebooks appear to exclusive event. The PRA have been included in the model. model was updated to reflect There are two gates under the new, deleted, or re-MUTEX gate which contain organized mutually exclusive mutually exclusive cutsets which modeling identified as part of are not documented in the system this review.

notebooks. While the majority of these are intuitively obvious (e.g.,

Revision 3 Page 52 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 11 Steam Generator Tube Rupture REFERENCE occurs as an IE AND 12 Steam SY-CO-[ALL]

Generator Tube Rupture occurs as an IE), these should be included in an appropriate system notebook.

(This F&O originated from SR QU-B7) 3-12 QU-D3 Quantification A review of the top cutsets from Complete Documentation of the cutset No impact on each event tree was performed. reviews was presented to ILRT analysis.

The utility stated that during this the peer review team; The original review, cutsets were reviewed to although, the documentation internal events determine if any mutually exclusive was separate from the cutset review events were contained within formal QU notebook notes have now cutsets, ifany flag settings were package. A note was added been archived.

inappropriate or ifany recoveries to the QU notebook directing were overlooked or added the reader to the location of inappropriately. A review of a the cutest review notes and sampling of cutsets did not indicate spreadsheets. The PRA any inappropriate results. configuration control However, the QU notebook does procedure, CNG-CM-1.01-not include a discussion of this 3003, requires a review of review. cutsets for PRA changes. In (This F&O originated from SR QU- practice, the top CDF and D3) LERF cutsets are examined for even the most innocuous model changes.

REFERENCE CNG-CM-1.01-3003 CO-QU-001 CO-FRQ-001 4-5 IE-A10 Complete To address this finding, the No impact on SY-Al0 Initiating Events The onlysystems shared mentionbetween in CO-SC-00l of the units Diesel Generator modeling ILRT analysis.

Revision 3 Page 53 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension IE-C3 is the SBO EDG, noted in Section was updated as described in The finding has SC-A4 4.1.2. Itstates that the SBO diesel Appendix H of C0-SY-023- been addressed can power any one bus on either 024, PRA DG System in the Internal unit. However, in the CAFTA Notebook. EOP-7 directs to Events model, model, there is an assumed bus align the OC DG to the unit which, in turn, is preference of 11, then 24, then 12, with redundant safety used in the ILRT then 23.* This is noted in the EDG equipment out-of-service, analysis.

system notebook but no basis is with a goal to restore at least provided. The procedures do not one 4KV bus. Since 4KV actually have a preference, which Buses 11 and 24 support yields a potentially non- AFW, those busses would conservative analysis. For have a preference over example, ifthere is a LOOP, the Busses 14 and 21, all else U2 diesels fail to start and the Ul being equal. No unit diesels fail to run after 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The preference is modeled. If SBO diesel would then be aligned there is a conflict in the to U2, and it is non-conservative to order-of-preference, for give the U1 bus 11 full credit. If example, both 4KV Bus 11 such non-conservatism is and 4KV Bus 24 are not negligible, some analysis should powered, then a 50-50 be performed to demonstrate this. probability is assumed as to the preferred bus.

(This F&O originated from SR IE-Al 0) REFERENCE C0-SY-023-024

• Note: Peer review finding was not precise. It should have stated bus preference for Unit 1 is 11, then 24, and for Unit 2, is 24 then 21.

4-12 HR-Cl Human One basic event calculated in the Complete The basic event has been No impact on Reliability appendix (ESFOHFCISZEFG) was added to the model. A ILRT analysis.

not included in the fault tree sensitivity run with the basic The missing models. CCNPP staff noted that it event included in the current basic event has Revision 3 Page 54 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension had previously been modeled, but model showed no increase been added to inadvertently deleted in an update. in risk. The system notebook the internal CO-SY-048 was updated. events model (This F&O originated from SR HR- used in the ILRT Cl) REFERENCE analysis.

CO-SY-048 4-15 IFEV-A6 Internal The internal flooding analysis did Open This finding has been No impact on Flooding not have a formal process to identified in the PRA ILRT analysis.

gather plant specific design configuration control The review of information, operating practices, database (CRMP), but has condition reports etc. that could potentially affect the not yet been addressed. did not identify generic flooding frequencies. In any design response to an NRC RAI on the issues or CCNPP ISI program plan, CCNPP operating mentioned a review of Condition practices that Reports that did not find any items would affect the that would increase the flooding generic flooding frequency. frequencies.

The CR review meets part of the requirement, but the SR also calls for reviews of plant design, operating practices, etc. that should be considered. The evaluation should be documented in the PRA.

(This F&O originated from SR IFEV-A6)

Page 55 of 93 Revision 33 Page 55 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 4-19 LE-C13 Large Early The sources of uncertainty are well Complete Dominant LERF cutsets No significant LE-F3 Release identified in Table 5-1 of the LE were reviewed to identify impact on ILRT LE-G4 notebook and quantified in Table uncertainties that could be analysis. The 5-2 of the QU notebook. However, addressed. Two changes dominant LERF no discussion of the uncertainties have been implemented to contributors were or insights from them is provided, address significant reviewed and For example, Sensitivity 1 shows a uncertainties and reduced model changes 74% reduction in LERF, but this LERF. First, a reverse-flow implemented.

large reduction is not investigated, check valve in the CVCS The Calvert Cliffs Letdown line was credited LERF Also, conservatisms in the ISLOCA as a potential ISLOCA contribution is analyses were discussed in the AS recovery. Second, a new now similar to review. SGTR was treated in an human action was added other PWRs.

overly conservative manner by with realistic timing for categorizing all SGTR as LERF. Steam Generator isolation and RCS depressurization (This F&O originated from SR LE- on a SGTR. These and less F3) significant model updates resulted in a LERF-to-CDF ratio change from approximately 17% to approximately 10%. This newer ratio is in the typical range for other PWRs.

REFERENCE CO-LE-001 4-20 LE-F1 Large Early The relative contribution to LERF Complete The contributions to LERF No impact on LE-G3 Release is presented in the QU notebook are documented in the ILRT analysis.

by PDS and by initiating event, but Quantification Notebook and This is an internal not by accident progression are noted as such in the events sequence, phenomena, Level 2 Notebook. Accident documentation containment challenges or progression sequences are issue.

containment failure mode. located in Section 4.2.3 and Revision 3 Page 56 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension Appendix C. The Level 2 (This F&O originated from SR LE- notebook has been updated G3) to point to additional phenomena and containment challenges and failure mode Tables/Figures in the QU Notebook REFERENCE CO-QU-001 CO-LE-001 4-21 LE-G5 Large Early The LE notebook states that Complete Section 5.5.2.7 of CO-LE- This internal Release limitations in the LE analysis that 001, Revision 2 - added events finding could impact applications are discussion of results of does not impact documented in the QU notebook, impact on application of the the ILRT but it is not. Given the conservative Unit 2 ILRT extension analysis.

modeling of SGTR and ISLOCA, request.

the impact on applications should include assessment of how this REFERENCE conservatism can skew the LERF CO-LE-001 results.

(This F&O originated from SR LE-G5) 4-22 LE-Cl0 Large Early The LERF contributors have not Complete The LERF results were No significant LE-C12 Release been reviewed for reasonableness reviewed for conservatisms impact on ILRT LE-F2 (per SR LE-F2). The QU notebook as described in the SRs. analysis. The LE-C3 discusses the top 20 LERF cutsets After conservatisms were dominant LERF (which total 73% of the total addressed (see discussion contributors were LERF). It notes conservatism in for F&0 4-19 above), no reviewed and the cutsets and says it will be significant issues were model changes evaluated in Section 5.2, but is not. identified. implemented.

Revision 3 Page 57 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension Section 4.3.6 of the QU notebook The Calvert Cliffs compares the total LERF of REFERENCE LERF CCNPP to St. Lucie, but does not CO-LE-001 contribution is even break the results down by now similar to contributor (e.g., SGTR, ISLOCA, other PWRs.

etc.).

Also, the ASME PRA Standard SRs C-3, C-10 and C-13 require a review of the LERF results for conservatism in the following areas:

1. Engineering analyses to support continued equipment operation or operator actions during severe accident progression that could reduce the LERF

2. Engineering analyses to support continued equipment operation or operations after containment failure.

3. Potential credit for repair of equipment.

No such review has been performed, despite the large conservatism noted in the containment bypasses.

(This F&O originated from SR LE-F2) 5-10 LE-D7 Large Early Following the failure of one or Complete The merits have been No impact to Release more containment penetrations to considered of adding an ILRT analysis.

isolate on CIAS, a feasible operator action in order Modeling of an operator action is to manually close containment operator action to close the failed valves from the penetration from the Main manually close Main Control Room. Control Room to recover failed valves from Revision 3 Page 58 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension from a containment isolation the main control (This F&O originated from SR LE- failure. A review of cutsets room would not D7) shows that a recovery is not significantly feasible for top LERF reduce LERF, as sequences, because the such an action is sequence includes either 1) not feasible for a loss of CR indication, 2) the significant includes a station black-out sequences where condition, or 3) includes containment non-recoverable pipe isolation has breaks. failed.

REFERENCE CO-LE-001 Attachment S 5-17 IE-C1 Initiating Bayesian updates of non-time- Complete CENG understands the No impact on IE-C13 Events based LOCA data were improper. general concern on ILRT analysis.

IE-C4 The small and medium LOCA Bayesian updating of rare The approach frequencies were obtained from events. However, the used for LOCA draft NUREG 1829 then Bayesian method used was based on frequencies has updated (in App E) with CCNPP a white paper developed by been validated by experience from 2004 to 2008. The industry experts regarding industry experts Very Small LOCA prior having LOCA frequencies. These and is the same alpha = 0.4, Mean = 1.57E-03; was experts included INL, NRC approach as was Bayesian updated to a Posterior and Industry experts. In used for the having a mean value of 7.02E-04. addition, the approach used NRC's SPAR This represents an excessive drop for the Calvert PRA was the model.

associated with CCNPP same as used for the NRC experience of 4 to 5 years. SPAR model. This issue is Similarly, the Small and Medium captured in the PRA LOCAs were Bayesian updated configuration control with the whole industry experience database (CRMP).

rcy data. The draft NUREG 1829 LOCA frequencies were obtained Revision 3 Page 59 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension from expert elicitations (not time- REFERENCE based) that included crack CO-IE-001 propagation analysis. The Bayesian update for VSLOCA used the Alpha parameter and the mean value to justify that the prior mean was based on 255 rcy. This may not have been the basis for the expert elicitations in NUREG 1829.

Also, the Medium LOCA frequency may be classified as extremely rare event. Itwould require no Bayesian updating. The current CCNPP SLOCA and MLOCA frequencies are very close even though the source data in NUREG 1829 indicates a negative exponential drop in these frequencies.

(This F&O originated from SR IE-Cl)

(Note: rcy - reactor year) 5-18 IE-C2 Initiating Justify the exclusion of LOOP Complete The event is not counted No impact on IE-C7 Events event at CCNPP in 1987. No time following guidance provided ILRT analysis.

trend analysis was provided to in NUREG/CR-6928, based The data analysis justify the exclusion, upon trend analysis. A full is acceptable.

discussion is included in the (This F&O originated from SR IE- Initiating Event notebook, C2) CO-IE-001.

REFERENCE CO-IE-001 Revision 3 Page 60 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 5-23 HR-A2 Human The Pre-Initiator HRAs did not Complete It is agreed that the No impact on Reliability include the miscalibration of SIT miscalibration of SIT ILRT analysis.

pressure. For example, in the pressure could have a Given the event where SIT pressure is negative impact on various pressure of the miscalibrated high, various accident scenarios involving CCNPP SITs accident scenarios requiring SI are LLOCA and VLLOCA they are only negatively impacted. Add SIT initiators. However, this required and pressure miscalibrated high or, instrumentation is not provide justify no impact on CDF / LERF. modeled explicitly and is significant benefit therefore deemed included on Large LOCAs.

(This F&O originated from SR HR- within the component The frequency of A2) boundary for the SIT. As a Large LOCA such the miscalibration times the pre-probability would be initiator included in the SIT frequency is unavailability, negligible.

REFERENCE CO-HR-001 5-25 SC-C1 Success Simplify the traceability of Tsw. In Complete Where applicable, the Tsw No impact on HR-12 Criteria the post initiator HRA details, the of each HFE that could be ILRT analysis.

SC-C2 HRA success criteria are often traced to the Success This is an internal provided as a positive re-statement Criteria notebook (CS-SC- events of the HRA title. And, the 001) was updated and documentation consequence of failure is often referenced in the HRA finding.

stated as core damage. Consider Calculator. CO-HR-001 was adding Tsw to the success criteria also updated.

and linking that to the PCTran case where Tsw was developed. REFERENCE Also, in the SC report (Table B-3), CO-HR-001 consider adding the actual time to core uncovery (or core damage)

Revision 3 Page 61 of 93

IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension instead of providing a "Yes" entry in the column of "core damage?"

(This F&O originated from SR HR-12) 5-30 LE-D1 Large Early Section 3.2.11 discussed the Complete CCNPP's Level 2 PRA No impact on LE-B2 Release containment challenge from follows the analysis in ILRT analysis.

Hydrogen Combustion. It WCAP-16341-P, Simplified The methodology concluded that the challenge may Level 2 Modeling in WCAP-16341-be significant for some accident Guidelines. In the industry- P is appropriate scenarios. The CCNP entry in supported analysis, the for Calvert Cliffs Table 6.11-2 of the Level 2 WCAP percentage of cladding level 2 analysis showed a potentially significant oxidation is the main factor for internal impact from Hydrogen burn. used to develop a maximum events initiators.

Provide an estimate of the impact H2 concentration in the of Hydrogen burn on containment containment, and, in turn, a pressure. Use an accident containment pressure is scenario that is likely to produce calculated ifthe H2 larger amounts of H2 with failed completely burns. These are containment spray. The optimal then mapped to site-specific time to estimate the impact of containment failure Hydrogen burn is approximately at probabilities.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> which is the time when the EOF and TSC personnel have A simplifying assumption is convened and 'are ready to guide made that "no pre-burning of the Main Control Room into hydrogen generated in the periodic Hydrogen burns before core melt progression is the formation of explosive considered." Calvert Cliffs' mixtures. severe action management procedures do include (This F&O originated from SR LE- actions to reduce H2 D1) concentration in the containment, but these Revision 3 Page 62 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension actions are not credited in the PRA model. Also, Containment Spray is not questioned for the LERF accident sequences.

Containment Spray is a factor in LATE containment failure accident sequences.

REFERENCES CO-LE-001 5-31 DA-D4 Data The summary table for Bayesian Complete The aforementioned No impact on the updated parameters (on Page 53 footnote was incorporated ILRT analysis for of the PRA Data Notebook, CO- into Table 2-6 of CO-DA-001. this minor DA-001, Rev. 1) shows the CS- internal events MDP was Bayesian updated with REFERENCE documentation plant experience containing 1 CO-DA-001 issue and no failure and Zero run-hours. The changes were CCNPP PRA staff responded to required for the this issue as an isolated case. CS-MDP failure There is an actual FTR > 1 hr rate.

(This F&O originated from SR DA-D4) 6-3 SC-B2 Success Expert judgment was not used as Complete The approach for SLOCA The existing Criteria the sole basis for any success break size analysis is analysis meets criteria. However, upon inspection discussed in the Success the intent of the of the PCTran run tables in the SC Criteria notebook. SR and therefore report appendices, many instances Furthermore, a review was there is no of surrogate or inferred results conducted of this issue; in impact on the were found. Instead of running addition, TH analyses were ILRT analysis.

specific PCTran calculations to completed to verify the Revision 3 Page 63 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension cover the whole SLOCA break size break-size ranges. It was spectrum, intermediate break sizes found that the computer have been calculated simulations adequately supplemented with expert represented the various judgment to derive limiting time break-size ranges.

delay for operators to actuate SI (30 min) or limiting time delay for REFERENCE OTCC (SGL<350'+10min). CO-SC-001 (This F&O originated from SR SC-B2) 6-5 SY-A20 Systems When appropriate, the Complete AFW basic event No impact on Analysis simultaneous unavailability within a AFW0TMMAINT6-F7 was ILRT analysis.

system is documented in the determined to not be needed The offending system notebooks and included in in the plant model. The basic basic event was the PRA model. However, a further event was removed. All removed from the review of these items is required remaining AFW equipment model. A review for completeness. unavailability events in the did not discover model and notebooks were other missing or (This F&O originated from SR SY- reviewed for consistency. incorrect A20) AFWOTMMAINT-TF was simultaneous determined to be modeled unavailability correctly, its description was events.

found to be in error in the system notebook. Notebook C0-SY-036 was updated. A review for concurrent maintenance was previously performed and documented in the Data Notebook.

REFERENCE C0-SY-036 Revision 3 Page 64 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension 6-8 HR-H2 Human Some recovery actions included in Complete For each screening HRA, No impact on Reliability the model (thus credited) are set to the internal events analysis ILRT analysis.

screening values. In the HEP was updated to include a The evaluation (appendices of the HR specific reference to the documentation report) there are no indications that earlier HRA analysis. for internal procedures, training, or other Included are the applicable events HRAs shaping factors are available on a success criteria for each was updated to plant-specific basis. recovery. Refer to CO-HR- address this 001, Internal Events Human finding.

(This F&O originated from SR HR- Reliability Analysis, and the H2) associated HRA Calculator file.

For Fire PRA development, the internal events HRAs with screening values were analyzed to assure that they were sufficiently conservative for fire scenarios. Refer to Section 4.1 of CO-HRA-001, Fire PRA Human Reliability Analysis. Documentation for fire HRA actions are similar Page 65 of 93 Revision 33 Page 65 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension to that done for the updated internal events HRA actions.

REFERENCE CO-HR-001 CO-HRA-001 6-9 HR-I1 Human The HR report is well documented Complete Updated the notebooks in No impact on Reliability in general andupgrdes will facilitate soe hoever bsicthe the ncebooksin reference section so Fire PRA. This is upgrades, however, some basic HRA designator names and a documentation event names are not consistent d finding. HRA between the HR report and the the HR Calculator, HtR names in the system notebooks. notebook, CAFTA Model model and 6.0. Changes included notebook are (This F&O originated from SR HR- adding the "-B" extension now consistent.

11) and removing the "(-2)"

event where applicable.

REFERENCE CO-HR-001 CO-SY-[Many]

6-10 IFPP-A2 Internal Plant design features such as Complete The Internal Flood notebook No impact on IFSN-A2 Flooding open rooms or as built divisions has been updated to ILRT analysis.

are used to define the flood areas incorporate an analysis This is a and was well documented. More describing the screening of documentation detail is needed as to why the the containment building finding for the containment buildings were from flooding analysis. Internal Flood screened from the analysis. Essentially, the containment notebook.

is designed for LOCA condition, which screens Revision 3 Page 66 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension (This F&O originated from SR reactor coolant system and IFPP-A2) related piping system. Other piping systems have limited inventory, are normally isolated, or have a low flow rate. Reference CO-IF-001.

REFERENCE CO-IF-001 6-14 IFSO-B1 Internal While the flooding calculations Open This is a documentation No impact on IFSN-A9 Flooding have been performed and are finding for the internal floods ILRT analysis.

thought to be correct and well notebook. The issue has This is a done, additional documentation of been captured in the PRA documentation data would enhance the IF report. configuration control issue.

Itappears that the input reports database (CRMP), but not and references are based on yet closed-out.

poorly documented or non-officially revisioned reports and information sources.

(This F&O originated from SR IFSN-A9) 6-16 IFQU- Internal Walkdowns have been conducted Open This is an internal floods No impact on Al1 Flooding and are documented in Appendix documentation finding. The ILRT analysis.

IFPP-B2 B of the IF report. It is stated in the finding has been captured in This is a IF report that prior information is the PRA configuration documentation no longer available; this fact control database (CRMP), issue.

should be corrected as required for but not yet addressed.

analysis updates and information verifications.

Revision 3 Page 67 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table I Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension (This F&O originated from SR IFQU-A1 1) 6-17 IFQU- Internal By including the flooding events Open The level of modeling detail No impact on A10 Flooding under the transient fault tree, the in the CCPRA is sufficiently ILRT analysis.

LERF impacts are automatically robust such that the model This is a accounted for in the same manner logic for flood impacts documentation as the general transient events in propagate appropriately issue.

the LERF analysis. Very little through the system fault documentation is found related to trees so that the equivalent the IF analysis in the LE report, general transient initiator although the IFreport states that (e.g loss of CCW) is the LERF impacts due to flooding appropriately defined in the are documented and analyzed in transient fault tree. In the LE report. addition, cutset reviews have not revealed the (This F&O originated from SR current modeling to be IFQU-A10) deficient in this regard.

This documentation of the above basis is captured in the PRA configuration control database (CRMP),

but not yet addressed.

6-18 HR-H2 Human The system time window Tsw for Complete Itwas determined that the No impact on Reliability post initiator HRAs was frequently text in Section 3.1.5.7 was ILRT analysis. As associated with 'core damage'. incorrect and does not described in this Post initiator HRAs that appear in capture how stress is F&O for internal the top cutsets may require actually applied in the EPRI events, the stress success criteria linked to beginning HRA Calculator. CO-HR-001, levels in the of core uncovery (about 20 Internal Events PRA Human model are minutes before 'core damage'). Or, Reliability Analysis, has appropriate, but the operator actions that may fall been updated to show the updates to the into that final 20-minute time stress level applied to each documentation Revision 3 Page 68 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic FindinglObservation Status Disposition Impact to ILRT Extension period should be overridden to HFE and the justification for are required. The assume a high stress level. While stress selection. Also internal events section 3.1.5.7 described this included is a correlation documentation approach, there is no evidence of between stress level and was updated.

its proper application in the HRA failure of execution quantifications. probability. New text has been provided for inclusion (This F&O originated from SR HR- in a future update of the H2) HRA notebook.

For the Fire PRA development, the internal events HRA stress levels were carried forward. As described in CO-HRA-001, Fire PRA Human Reliability Analysis, additional stresses were evaluated and incorporated due to the fire initiator.

REFERENCE CO-HR-001 CO-HRA-001 6-22 HR-El Human Upon RAS, LPSI stops and EOP- Complete As documented in CR-2009- No impact on Reliability 5, Step S.1(d) requires the 005881, shutting the RWT ILRT analysis.

Operators to 'Shut RWT OUT outlet valves upon a RAS The system is Valves SI-4142, 4143'. This does not impact station operable without manual action was not modeled in operability. The Safety the manual the PRA. The CCNPP PRA staff Injection Pumps and action to shut the provided reasonable response to Containment Spray Pumps RWT outlet this issue. Based on CR-2009- will not fail ifthe RWT valves. There is 005581, there is no impact on isolation valves do not close no impact on Revision 3 Page 69 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension pump operability. Also, the staff with a RAS signal. A design internal events will continue to track the CR. If margin issue has been CDF. The issue there are any changes to the identified. This issue has was added to the disposition of pump operability, been added to the plant's plant's margin then a new HRA may be added to margin management management the PRA model (ifwarranted). program. No model changes program.

have been made, but the (This F&O originated from SR HR- PRA configuration El ) management program, CNG-CM-1.01-3003, would capture any design changes concerning this issue.

REFERENCE CO-SY-052 CR-2009-005881 CNG-CM-1.01-3003 6-23 HR-G7 Human When the Calculator reads in the Open New HRA events, No impact on Reliability combinations, it assumes that CVCOHFBHEOTA-B-8HRS ILRT actions occur in the order of the and AFW0HF-CC-SGDEC- analysis. The time delay (Td). However, the time 8HR were added to model new HRA events delay is not the same for all Td variances where CST are not sequences, and care must be depletion occurs early and significant.

taken to make the combinations when it occurs later. This appropriate for the sequences in accounts for appropriate which they occur. Page 88 of the sequencing of events.

HRA notebook indicates this was considered, since the Td was This specific issue with time modified for events occurring prior delay and CST depletion to reactor trip, and also for OTCC has been addressed and after SG overfill. However, not all incorporated into the PRA occurrences have been model. An updated addressed. The combination dependency analysis has Revision 3 Page 70 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension examined by the review team is been performed, which Combination 770 (OTCC after CST includes these new HRA depletion). Inthis event the CST events. The dependency depletion should come first. analysis shows that these new HRA actions are not (This F&O originated from SR HR- significant for CDF or G7) LERF. A PRA configuration control database (CRMP) item has been initiated to formally incorporate the updated dependency analysis into the model.

REFERENCE CO-HR-001 CO-HRA-001 7-13 QU-A2 Quantification Discrepancy between Complete The top flood cutset was No impact on documentation and result files. incorrectly flagged as being ILRT analysis for SB0037 and SB0038 sequences SBO sequence 37 (offsite this internal appear to be inverted in Tables D- power recovered < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />) events 1, 4.2.2, 4.2.4, 4.2.5, B-3). instead of sequence 38 documentation (offsite power not issue.

(This F&O originated from SR QU- recovered). Updated tables A2) B-2, C-1, and D 1 in CO-QU-001. Spot-check was performed to identify other errors. In CO-QU-002, fixed sequence 12 table 4.2-5, which incorrectly showed sequence 37 instead of 38.

Page 71 of 93 Revision 3 3 Page 71 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations F&O ID SR Topic Finding/Observation Status Disposition Impact to ILRT Extension REFERENCE CO-QU-001 CO-QU-002 Page 72 of 93 Revision 33 Revision Page 72 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis PP- PP-B3 Plant Complete The containment is partitioned CO-PP-001, Calvert Cliffs Fire PRA Plant No impact to B3- PP-B6 Partitioning into 2 PAUs. There are Partitioning Notebook, was updated to ILRT analysis, 01 PP-C3 intervening combustibles and include an analysis that justifies the as this affects this was accounted for in the partitioning of the containment into two the FPRA plant PRA by treating the 20 feet as plant partitioning units with a 20-foot spatial partitioning an overlap region and failing separation (known as the buffer zone). The analysis.

components affected in both only potential intervening combustibles in PAUs. There is no justification this buffer zone were identified as qualified given for the 20 foot cables that were verified to be encased assumption. The turbine deck within marinate covered raceways. The is continuous from unit 1 to unit covers prevent the cables from becoming

2. This area is divided into 2 potential combustibles and therefore are PAUs, TURB1 and TURB2, but not considered intervening combustibles.

there is no discussion for the basis of the partitioning. The unit 1 and unit 2 Turbine Deck was Finding level of significance is walked down to assess for the acceptability baseparaon wredithino r site of the Appendix R partitioning into distinct separation with no requisite PAUs. The boundary was assessed to justification. have at least a 20-foot separation between Maintain the containment as 1 potential ignition sources and potential PAU and discern the targets, assessed for intervening separation of east from west in combustibles, and the Turbine deck the fire modeling. Document volume assessed for damaging hot gas the spatial separation and no layer development. The partitioning was intervening combustibles for found acceptable and consistent with the turbine deck. NUREG/CR-6850, Section 1.5.2, where main turbine decks are typical applications where spatial separation has been credited.

PP- PP-B5 Plant Complete The water curtain in the CCW The Component Cooling Water room water No impact to 35- PP-C3 Partitioning room was credited as an active curtain is an approved Appendix R ILRT analysis, 01 fire barrier. The justification exemption, as identified in the exemption as this affects Reiin3Pae7 f9 Revision 3 Page 73 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis was that the water curtain was issued by the NRC in response to Calvert the FPRA plant part of the original regulatory Cliffs exemption request ER820816. The partitioning fire protection program. This validity of crediting CCW Room Water analysis.

meets CAT 1, but needs Curtains is discussed in Southwest enhancement for CAT Il/111. Research Institute Report No. 01-0763-Finding level was used 201. A reference to the Southwest because the requirements for Research Institute report was added to CO-CAT Il/111 were not met. PP-001, Plant Partitioning Notebook.

Calvert Cliffs should provide a direct reference to their Appendix R program as the basis for the acceptability for this or provide a design basis justification for the water curtain and document that in the PP notebook ifthe Appendix R program reference cannot be found.

PP- PP-B7 Plant Complete 1. The walk down A table was created to correlate the No impact to B7- PP-C3 Partitioning nomenclature does not match building or area nomenclature that was ILRT analysis, 01 PP-C4 Qualitative the PP notebook. Example used for the plant walkdown as this affects QLS- Screening page 561 of the walkdown documentation, to the plant analysis unit the FPRA plant documentation uses identifiers used in the Fire PRA analysis. partitioning Al nomenclature in the This table was added to CO-PP-001, documentation.

containment that does not Calvert Cliffs Fire PRA Plant Partitioning match the PP notebook. Notebook as Table 17.

2. There are many areas inaccessible such as: #23 The facilities and rooms that were not Charging Pump Room, U1 originally walked-down were reviewed.

Service Water Pump Room, Supplemental walkdowns were performed U1 East Battery Room, E/W and supplemental walkdown datasheets Revision 3 Page 74 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis Corridor. These areas appear were generated. For areas that were not to be accessible with a little accessible at the time of the supplemental effort. In some of the areas walkdowns (for radiological safety reasons, screened out in QLS, the areas personnel safety concerns, or access were inaccessible and did not otherwise denied), The reason for have a confirmatory walkdown. inaccessibility was added to Table 17.

Finding level assessed due to the incompleteness of the walkdown documentation.

1. Prepare a table that correlates the PAUs from the PP notebook with the area nomenclature used in the walkdown documentation.

2. Complete the walkdowns, particularly for areas screened in the QLS task.

CS- CS-B1 Fire PRA Complete Current Breaker coordination The breaker coordination study has been No impact to B1- CS-C4 Cable study still in progress. This completed. As described in ECP ILRT analysis, 01 Selection study needs to be completed in 000321, Form 12, Engineering Evaluation, as this affects and order to receive a category II all PRA common power supplies are the FPRA plant Location met for CS-BI. assumed to meet - or will meet - the Cable coordination requirements of NFPA 805, Selection Complete the breaker except as noted in CO-CS-001, Fire PRA analysis and coordination study. Cable Selection Notebook. As described in the item has the cable selection notebook, two 120VAC been lighting panels are not validated as completed.

coordinated, and these panels are assumed to fail for all Fire PRA scenarios.

Also, as described in the PRA notebook a breaker for 480V motor control center Revision 3 Page 75 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis MCC101BT has not been validated as coordinated. This breaker, 52-10150, is modeled so that a fire-induced electrical fault on the breaker's power cabling will fail MCC101BT. Finally, the notebook identifies that selected 120V power panels have coordination issues, but that these will be addressed by design changes and referenced in Attachment S - Modifications and Implementation Items.

PRM- PRM- Fire Complete The FPRA model did not Loss of Control Room HVAC can affect the No impact to B3- B3 PRA/Plant address events involving loss operability and availability of equipment in the ILRT 01 PRM- Response of both HVAC trains to the the control room and cable spreading analysis, as the B4 Model MCR, long term heatup of room. As described in Calvert PRA System loss of MCR MCR and need for operator Analysis Notebooks CO-SY-002, CO-SY- HVAC PRM- actions outside the MCR to 017, and C0-SY-030, loss of HVAC is modeling has B5 compensate for the loss of modeled to have the effect of increasing been electronic controls in the MCR, the failure rate of 120VAC and 125VDC implemented in which was assumed as a instruments and controls in the cable the models CCDP of 1.0 for the plant. The spreading room. For the control room, used in the basis for excluding this degradation of the 125VDC system is used ILRT analysis potential Core Damage as a conservative surrogate for control sequence was addressed in room I&C degradation.

questions to the Calvert Cliffs PRA team. This sequence is a Loss of Control Room HVAC and new sequence outside the subsequent temperature increases may current FPRA model logic adversely affect operator responses. The trees.

model reflects degradation of human actions by the degradation of the 125VDC Consider using a combination system used for instruments and controls.

of MCR heatup calculations to Loss of Control Room HVAC is not define the time when operators expected to cause abandonment by would leave MCR and consider Page 76 of 93 Revision 33 Revision Page 76 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis a recovery action for restoring operations staff of the control room due to cooling the MCR. high temperatures. On complete loss of HVAC with no mitigation, such as no use of emergency fans, calculation CA02725 shows a CR temperature of 123 deg F at 24-hours. While this is a challenging environment, this temperature is assessed as insufficient to solely drive a complete CR abandonment scenario. NUREG/CR-6738 describes operational experience where operators will continue to occupy the control room even under severe environments.

Operations staff says that in consideration of high temperatures in the control room, that Operations would do what was needed to keep the cores safe and covered. The site safety director says that for a temperature of 123 deg F, the site would implement a mitigation strategy which would include stay-times, assessment of individuals for heat-related conditions, use of ice vests, and call-in of additional qualified operations staff to rotate into the control room.

The above discussion was included in CO-SY-030, Control Room HVAC PRA System Notebook.

Page 77 of 93 Revision 33 Page 77 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension I RCA-54001 -OOO-CALC-OO1 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis FSS- FSS- Fire Complete A range of ignition source / FDS modeling was used for fire scenario No impact to A5- A5 Scenario target set combinations has evaluations in the Cable Spreading Rooms the ILRT 01 Selection been represented for and Switchgear Rooms. In both cases, analysis, as and unscreened PAUs. These thermocouple location was adjusted as this affects the Analysis combinations are identified in identified in F&O FSS-D3-02. For the CSR, FPRA model relevant calculation sheets for consequences were divided into scenarios and the item is unscreened PAUs. In some based on mitigation potential. First, if the complete.

PAUs, sub-PAUs are defined scenario was suppressed by the Halon and damage from a potential system then the limit of damage was based fire within the sub-PAU is on what was predicted by FDS in terms of addressed. However, it is not temperature and energy. If it was clear how or why damage unsuppressed it went to total room bum, would be limited to the which assumes failure of all targets in the specified sub-PAU because room, regardless of the initial scenario there are no physical barriers boundary. For the Switchgear Room FDS between specified sub-PAUs. analysis, the analysis was updated to add The documentation is such clarity to the analysis. A discussion of the that it cannot be determined if application of sub-PAUs has been added to the selected fire scenarios Addendum 1 to CO-FSS-004, Fire PRA provide reasonable assurance Detailed Fire Modeling Notebook. Damage that the risk contribution of was not limited to specified sub-PAUs.

each unscreened PAU can be Specific examples of the treatment of fire characterized. Another issue growth and the application of sub-PAUs that influences the potential for have been provided.

fire propagation across sub-PAU boundaries is that the As described in CO-FSS-004, the sub-PAU temperature measurement analysis included spatial information from locations specified in the walkdown, along with engineering detailed FDS fire modeling judgment, to determine if fire sources could evaluations do not generally fail additional components, cables, or other coincide with locations where combustibles, potentially leading to more maximum temperature are damage to surrounding equipment or cables. For scenarios that leveraged FDT Page 78 of 93 Revision 33 Page 78 of 93

1RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis expected (e.g., within the fire modeling, the issue related to whether the plume). analysis had correctly addressed the impact of transients along the edge of a As a consequence, for some boundary interface for a sub-PAU. A fire csecenarios dame fire scenarios damage tom to comparable to secondaryconsideration wasoilalso combustion and related fires.

targets is not predicted when it Resolution involved selection of several shoulfedb mased oither. representative PAUs for a sensitivity study specified damage criteria, that expanded the existing sub-PAUs and Somthe bascios artepercreed examined secondary ignition potential.

on the basis of temperature measurements that do not represent conditions at targets within the fire plume. (See F&O FSS-D3-02) This could have a significant impact on the potential for fire propagation across sub-PAU boundaries and needs to be discussed more thoroughly.

FSS- FSS- Fire Complete There were indications that The PAUs were considered representative No impact to A5- A5 Scenario Calvert Cliffs had the tools and of the work performed based on several the ILRT 01 Selection information in place to properly criteria. The analysis indicated that the analysis, as and evaluate the propagation of methods mentioned were indeed this affects the Analysis fires across the sub-PAU appropriate. Sub-PAU impacts did not FPRA model boundaries given no physical change from the expanded assessment and the item is barriers but there were no and that secondary ignition was bounded complete.

examples showing that this by the existing analysis and was evaluation was performed or appropriately addressed. The analysis was any explicit descriptions of how incorporated into the documentation for they were performed in CO-FSS-004.

general. The concern here is that without an explicit Revision 3 Page 79 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension I RCA-54001 -OOO-CALC-OO1 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis description of the process for evaluating the spread of fires across sub-PAU boundaries with no physical barriers and detailed examples, there is the potential that in the future, new people updating the PRA may not know that they have to evaluate this.

Calvert Cliffs needs to describe their process for evaluating fire growth and propagation between sub-PAUs and as applicable, between PAUs.

Specific examples of the sub-PAU fire growth need to be provided. If fire propagation from sub-PAU to sub-PAU was not treated, Calvert Cliffs needs to evaluate all sub-PAUs to determine ifthere is any potential for fire spread and then model the potential for spreading fires and for damage occurring across sub-PAU boundaries.

FSS- FSS- Fire Complete Where used, the FDS model FDS modeling was used for fire scenario No impact to D2- D2 Scenario was generally used with a level evaluations in the Cable Spreading Rooms the ILRT 01 Selection of grid resolution that was and Switchgear Rooms. analysis, as and below the level of grid this affects the Analysis resolution documented in the FPRA model Revision 3 Page 80 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis NUREG-1824 Verification and For the Cable Spreading Room FDS fire and the item is Validation study for the FDS scenarios, a grid study was performed on complete.

model. A validation study was the updated FDS model. The study not conducted to support the recommended a grid size that was within use of this lower level of grid the range in NUREG/CR-1824. That grid resolution. Grid resolution has size was used for CSR FDS scenario a bearing on the results of FDS evaluations. The study and results were calculations. Grid resolutions incorporated into CO-FSS-004, Fire PRA outside the validation range in Detailed Fire Modeling Notebook.

NUREG-1824 should be justified and validated. The Unit 1 27' and 45' Switchgear Rooms Increase thethelevel of o grid Inreaselutin were updated to increase the level of grid ll g Uid e resolution to a value that is within the resolution in the FDS PAU Fire vldto ag ouetdi Evaluations (C0-FSS-004 R1) validation range documented in EvthaluationsidCreFolutiRi) NUREG/CR-1824. Results calculated in so that the grid resolution is the Unit 1 FDS models were applied to Unit documented in NUREG-1824. 2. Results of the updated model are incorporated into C0-FSS-004 as Addendum 1.

FSS- FSS- Fire Complete This SR is not met because FDS modeling was used for fire scenario No impact to D3- D3 Scenario detailed FDS fire modeling evaluations in the Cable Spreading Rooms the ILRT 01 FSS- Selection evaluations of PAUs 302, 306, and Switchgear Rooms. analysis, as B2 and 311, 317, 407 and 430 assume this affects the FSS- Analysis that material surfaces are For the Cable Spreading Room FDS fire FPRA model D4 "inert." As noted C0-FSS-004 R1, on thisp. 44 of scenarios, the Unit 1 CSR was modified to and the item is complete.

include actual material properties and assumption was made "... so sensitivity analysis. Actual material the PAU structure (walls, floor, properties were used in the updated or ceiling) itself would absorb U1CSR FDS model rather than the prior anyceiling)mitself vaoulbsofrb use of "inert" material conditions. Adiabatic any heat from the various fire conditions were used for any items with scenarios, producing a more material properties that are unknown or of Revision 3 Page 81 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis conservative or worst case a high uncertainty to bound the analysis result for all fire scenarios' and prevent heat transfer into those impacts to the components objects. The CSR FDS model was and cables within the PAU executed and the results compared to the model. As such, no detailed baseline results. This study was then material properties were documented in FSS-004. The results were required to be defined in FDS applied to Unit 2 CSR. This study was then for the scenarios to function documented in FSS-004, Fire PRA correctly." However, Detailed Fire Modeling Notebook.

specification of material surfaces as "inert" in FDS does The Unit 1 27' and 45' Switchgear Rooms not prevent heat absorption were updated to specify representative into material surfaces. On the material properties as referenced by contrary, this specification NUREG 1805. This adjustment enabled the maintains material surfaces at analysis to obtain more realistic estimates ambient temperature in FDS, of environmental conditions for these fire which tends to maximize heat scenarios. Results calculated in the Unit 1 absorption into these surfaces. FDS models were applied to Unit 2.

To prevent heat absorption into Results of the updated model are material surfaces, they should incorporated into CO-FSS-004 as have been specified as Addendum 1.

"adiabatic" rather than as "inert." The "inert" parameter in FOS maximizes heat transfer to surfaces rather than minimize it. This can result in lower calculated gas temperatures.

Specify materials surfaces as "adiabatic" rather than as "inert" in EDS to prevent them from absorbing heat in order to achieve the stated goal of Revision 3 Page 82 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension I RCA-54001 -OOO-CALC-OO1 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis producing a more conservative or worst case result. This may prove to be overly conservative, in which case specification of realistic material properties could be used to achieve more realistic estimates of environmental conditions for these fire scenarios.

FSS- FSS- Fire Complete Temperature measurement FDS modeling was used for fire scenario No impact to D3- D3 Scenario locations specified in the evaluations in the Cable Spreading Rooms the ILRT 02 FSS- Selection detailed FDS fire modeling and Switchgear Rooms. analysis, as A5 and evaluations do not generally this affects the Analysis coincide with locations where FPRA model For the Cable Spreading Room FDS fire maximum temperature are and the item is scenarios, new measurement devices were expected (e.g., within the fire complete.

included in the updated U1CSR FDS plume). As a consequence, for model. The thermocouples were placed some fire scenarios damage to directly above the fire source in the targets is not predicted when it updated FDS model and the scenarios re-should be based on the evaluated. The results were applied to Unit specified damage criteria.

2 CSR. This study and the results were Some scenarios are screened then documented in FSS-004, Fire PRA on the basis of temperature Detailed Fire Modeling Notebook.

measurements that do not represent conditions at targets within the fire plume. The Unit 1 27' and 45' SWGR rooms were updated to alter the location of the Re-run FDS simulations with thermocouples such that the centerline temperature measurement plume temperature was recorded and used probes located within the fire to determine target impacts. Results Page 83 of 93 Revision 33 Page 83 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis plume or use other fire calculated in the Unit 1 FDS models were modeling tools such as FDTs applied to Unit 2. Results of the updated to calculate fire plume model are incorporated into CO-FSS-004 temperatures for these as Addendum 1.

scenarios.

FSS- FSS- Fire Complete Fire detection timing is FDS modeling was used for fire scenario No impact to D8- D8 Scenario evaluated for detailed fire evaluations in the Cable Spreading Rooms the ILRT 01 Selection modeling cases that use FDS. and Switchgear Rooms. analysis, as and This fire detection timing is For the updated Cable Spreading Room this affects the Analysis then used to estimate FDS fire scenarios, cable tray obstructions FPRA model automatic fire suppression were placed in the ceiling area of the and the item is timing and fire brigade updated UICSR FDS model. Additional complete.

response timing for these thermocouple and heat flux data recording scenarios. However, the fire devices were added to the U1CSR model detection timing is based on under the new cable tray obstructions in modeling that does not include the vicinity of the fire source. The scenarios obstructions located beneath were re-evaluated. The results were the ceiling that could have an applied to Unit 2. A sensitivity study was impact on fire detector also performed. The study and new response. The fire detection scenario results were incorporated into CO-timing is also based on an FSS-004, Fire PRA Detailed Fire Modeling unjustified assumption Notebook.

regarding the type of smoke detectors installed in the affected PAUs. Obstructions to The Unit 1 27' and 45' SWGR rooms were the flow of fire gases can have also updated to include significant an impact on smoke obstructions such as cable trays and beam concentrations and velocities, pockets within the switchgear rooms.

which in turn influence smoke Results calculated in the Unit 1 FDS detector response. Without models were applied to Unit 2. Results and including such obstructions in details of this analysis are documented in fire modeling simulations, their CO-FSS-004 as Addendum 1.

Revision 3 Page 84 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis impact on fire detection times is not evaluated.

Include obstructions located beneath the ceiling for the affected fire scenarios in order to evaluate their impact on fire detection timing. Provide justification for the selection of the type of smoke detector specified in the FDS simulations for these fire scenarios.

FSS- FSS- Fire Complete To achieve CC Il/111 for this SR, The Turbine Building was reviewed for No impact to F3-01 F3 Scenario a quantitative assessment of potential fire scenarios where structural the ILRT Selection the risk of the selected fire steel can be adversely affected. From the analysis, as and scenarios involving a) exposed scenarios examined, those that can this affects the Analysis structural steel and b) the damage structural steel were selected for FPRA model presence of a high-hazard fire further analysis. The frequency, severity and the item is sources must be completed factor and non-suppression probability of complete.

consistent with the FQ each scenario were developed and requirements including the included in the Structural Failure Analysis collapse of the exposed Notebook. These impacts were then added structural steel and any to FRANX database and quantified as part attendant damage. Such an of the final Fire PRA risk quantification in assessment has not been Fire Quantification Notebooks CO-FRQ-001 done or was not documented and CO-FRQ-002.

in a readily discernible manner.

This has a potential impact on fire risk quantification.

Page 85 of 93 Revision 33 Page 85 of 93

1 RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis Complete a quantitative assessment of the risk of the selected exposed structural steel fire scenarios consistent with the FQ requirements.

FSS- FSS- Fire Complete An assessment of the Generic probabilities were used for No impact to G4- G4 Scenario effectiveness, reliability and credited passive fire barrier features in the the ILRT 01 Selection availability of credited passive multi-compartment analysis. At Calvert analysis, as and fire barrier features has not Cliffs, the fire barriers are verified to be this affects the Analysis been documented in the multi- effective through test procedures. An FPRA model compartment analysis. To unreliability value was applied to all and the item is achieve a CC II capability normally closed doors that represents the complete.

assessment, the effectiveness, probability of the door being propped open reliability and availability of given a fire in the exposing compartment.

credited passive fire barrier The probability of finding a failed sealed features must be assessed. wall penetration is assumed to be very small to warrant propagation scenarios. A discussion of theof effectiveness, reliability, Assess the effectiveness, and availability fire barriers was added to reliability and availability of CO-FSS-008, Calvert Fire PRA Multi-credited passive fire barrier Compartment Analysis.

features and document this assessment.

FSS- FSS- Fire Complete The effectiveness, reliability Active fire barriers were evaluated as No impact to G5- G5 Scenario and availability of credited effective in studies used to support the ILRT 01 Selection active fire barrier features have Appendix R analysis. An unreliability value analysis, as and not been quantified in the has been applied to all normally open, self this affects the Analysis multi-compartment analysis. closing dampers and doors; A discussion FPRA model To achieve a CC II capability of the effectiveness of credited active fire and the item is assessment, the effectiveness, barriers was added to CO-FSS-008, Calvert complete.

reliability and availability of Fire PRA Multi-Compartment Analysis.

credited active fire barrier Revision 3 Page 86 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis features must be quantified.

Quantify the effectiveness, reliability and availability of credited active fire barrier features and document this assessment.

HRA- HRA- Human Complete Improve documentation of the CO-HRA-001, Fire Human Reliability No impact to B2- B2 Reliability adverse operator actions notebook, was updated to detail the the ILRT 01 Analysis needed to address the impact adverse operator actions added to the analysis, as of grounded or shorted model following the fire AOP review this affects the electrical buses that might process. Table 3 was added to Section 2.2 FPRA model have an impact on other plant detailing each basic event, set to true (1.0) documentation buses ifnot isolated and re used in the model to annotate the adverse and the item is energized in the areas operator actions in the model. These complete.

identified. Very difficult to find include actions to de-energize electrical the information within the HRA busses to isolate them from potential notebook alone, because the shorts and grounds. Table 2 shows the actions are modeled as inputs HFEs added to the model as part of the to FRANX. AOP review, including actions to restore AC power to busses lost due to fire failure Provide new tables listing the sequences.

actions considered or references to specific locations.

HRA- HRA- Human Complete Documentation for what was CO-HRA-001, Fire Human Reliability No impact to El- El Reliability done was very good, however, Notebook, was updated detailing the Alarm the ILRT 01 Analysis the details for not selecting any Response Procedure review process. analysis, as spurious alarms is not clear. Table 12 was expanded to show the ARP this affects the The documentation of the review of alarm impact and operator FPRA model adverse actions put into the interview notes for CR annunciators that and Revision 3 Page 87 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension I RCA-54001 -OOO-CALC-OO1 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis model as "true" are not in the could result in a manual reactor trip. No documentation HRA report, actions identified annunciators were identified that would and the item is in the cutset reviews are not cause the operator to terminate a systems complete.

clearly identified, rational for or components operation based solely on not using specific HFEs in the the alarm itself, but several were identified RCP trip actions, for identifying that could potentially result in the operator actions from procedures and tripping the Unit unnecessarily.

the process for assigning uncertainty range for the CO-HRA-001 was also updated to detail the combos. Doesn't permit adverse operator actions added to the verification of the rational for model following the fire AOP review judgments made in deciding process. Table 3 was added to Section 2.2 what is in and out of the Fire detailing each basic event, set to true (1.0)

HRA. Also, from the used in the model to annotate the adverse calculation viewpoint the need operator actions in the model. These to know the use of all include actions to de-energize electrical manpower requirements during busses to isolate them from potential early time after fire initiator for shorts and grounds. Table 2 shows the dependency analysis.

HFEs added to the model as part of the AOP review, including actions to restore Enhance documentation of the AC power to busses lost due to fire failure specific issues needed to sequences.

reproduce the assumptions and calculations used in the HRA. New HFEs added as part of the cutset review process are identified in Table 1 of CO-HRA-001, Fire Human Action Reliability notebook. These are annotated with "identified during the development of the PRM Notebook." The cutset reviews are described in CO-QNS-001, Fire PRA Quantitative Screening Notebook. A new dependency analysis was performed after the new HFEs were added to the model, Page 88 of 93 Revision 33 Page 88 of 93

IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis ensuring new dependency combinations are considered.

Additional information was added to Table 1 of the Human Reliability Analysis Notebook, CO-HRA-001, detailing why each HFE was either retained or removed.

For example, event FGAFWOSGTRISOL, Operator Feeds Affected SG with SGTR to Assure Heat Removal, was "Not retained for fire scenarios, because these actions are SGTR specific. Modeling was not necessary to ensure these actions did not appear in the cutsets, because the SGTR initiator is not being used for fire scenarios."

Combination event multipliers are used in cutsets of multiple HEP actions to account for dependencies between HEP actions. To account for the uncertainty in HEP actions, an uncertainty parameter is added to the HEP action. When performing uncertainty analysis, the uncertainty parameters for combination events is increased proportionally when they are multiplied by the combination event multipliers.

Based on interviews, there are sufficient non-control room personnel for fire recovery actions. Appendix D of CO-HRA-001 notes that there are no control room Revision 3 Page 89 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis operators assigned to the fire brigade.

There were no identified staffing issues or interferences between operators performing fire recovery actions and members of the fire brigade.

FQ- FQ-A1 Fire Risk Complete Treatment of 0 CCDPs The fire risk quantification process has No impact to Al- Quantificati scenarios is not clear and been updated in notebooks CO-FRQ-001 the ILRT 01 on appears to result in an and C0-FRQ-002 to address the issue with analysis, as underestimate of total risk (the FRANX fire scenarios having a zero this affects the underestimate appears to be conditional probability for CDF and LERF. FPRA model small based on the sensitivity and the item is evaluations performed): 1. When documented analysis shows that complete.

1 - with respect to opposite unit selected fire scenarios for one unit are quantification, use CCDP for screened from impact for the opposite unit reactor trip initiator unless (typically, no trip would be initiated), then confirmation of no trip is that scenario may be excluded from the documented; opposite unit's fire risk quantification.

2 - address use of 0 CCDP for Otherwise, a nominal conditional control room HVAC loss probability, as described in item 3 below, scenarios, apply CCDP would apply.

consistent with control room abandonment 2. F&O PRM-B3-01 identifies the concern 3 - for scenarios with limited with loss of Control Room HVAC with impact with a 0 CCDP, due to control room abandonment. As discussed cutsets below truncation limit, in more detail with the resolution to PRM-apply a baseline CCDP based B3-01, subsequent investigation revealed on reactor trip initiator that loss of CR HVAC is not expected to cause abandonment by the operations staff More than 50% of the of the control room due to high scenarios have a 0 CCDP but temperatures. Loss of CR HVAC and no clear discussion of the subsequent temperature increases may adversely affect operator responses, and Revision 3 Page 90 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis basis for the 0 CCDP is the model reflects degradation of human provided, actions with loss of CR HVAC. CO-SY-030, Control Room HVAC PRA System Treatment of 0 CCDPs Notebook, was updated to include this Trenatmtof scenarios:

0discussion.

1 - with respect to opposite unit quantification, use CCDP for 3. The new quantification process reactor trip initiator unless described in the FRQ notebooks is to confirmation of no trip is assure a nominal conditional value is documented; calculated for these low significant scenarios by 1) recalculating the zero-2 - address use of 0 CCDP for conditional scenarios at a lower truncation scntrol, room HVACy losvalue to assure resolution in the scenario scenarios, apply CCDP cutset file and conditional probabilities, consistent with control room and/or to 2) use a baseline conditional abandonment probability for CDF and LERF for the 3 - for scenarios with limited internal events reactor trip initiating vent -

impact with a 0 CCDP, due to IEOPT for Unit 1 or IEOPT-2 for Unit 2 cutsets below truncation limit, apply a baseline CCDP based on reactor trip initiator FQ- FQ-B1 Fire Risk Complete We observed zero CCDPs for The fire risk quantification process has No impact to B1- Quantificati some PAU CDF and LERF been updated in notebooks C0-FRQ-001 the ILRT 01 on values in the FRANX tables and C0-FRQ-002 to address the issue with analysis, as (e.g., PAU 512) which FRANX fire scenarios having a zero this affects the eliminated loss of HVAC to the conditional probability for CDF and LERF. FPRA model MCR as a potential MCR and the item is abandonment sequence. 1. When documented analysis shows that complete.

Treatment of 0 CCDPs selected fire scenarios for one unit are scenarios: screened from impact for the opposite unit 1 - with respect to opposite unit (typically, no trip would be initiated), then quantification, use CCDP for that scenario may be excluded from the Revision 3 Page 91 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis reactor trip initiator unless opposite unit's fire risk quantification.

confirmation of no trip is Otherwise, a nominal conditional documented; probability, as described in item 3 below, 2 - address use of 0 CCDP for would apply.

control room HVAC loss scenarios, apply CCDP 2. F&O PRM-B3-01 identifies the concern consistent with control room with loss of Control Room HVAC with abandonment (F&O FQ-Al-01 control room abandonment. As discussed (F)) in more detail with the resolution to PRM-3 - for scenarios with limited B3-01, subsequent investigation revealed impact with a 0 CCDP, due to that loss of CR HVAC is not expected to cutsets below truncation limit, cause abandonment by the operations staff apply a baseline CCDP based of the control room due to high on reactor trip initiator temperatures. Loss of CR HVAC and Allowing zero CCDPs allows subsequent temperature increases may scenarios in the fire model to adversely affect operator responses, and quantify with no contribution to the model reflects degradation of human the CDF or LERF value and actions with loss of CR HVAC. CO-SY-030, this under represents those Control Room HVAC PRA System frequencies especially when Notebook, was updated to include this considering delta risk discussion.

evaluations.

3. The new quantification process Replace the zero entries with described in the FRQ notebooks is to the lowest CCPD for a plant assure a nominal conditional value is trip with only random failures of calculated for these low significant the safety equipment as in the scenarios by 1) recalculating the zero-internal events model. We conditional scenarios at a lower truncation discussed this with the Calvert value to assure resolution in the scenario Cliffs PRA team and some of cutset file and conditional probabilities, the zeros are due to fire areas and/or to 2) use a baseline conditional in one unit potentially probability for CDF and LERF for the contributing to the CCDP of the Page 92 of 93 Revision 33 Page 92 of 93

I RCA-54001 -000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations - Findings F&O SR Topic Status Finding Disposition Impact on ID ILRT Analysis opposite unit. With the internal events reactor trip initiating vent -

exception of these cases a IEOPT for Unit 1 or IEOPT-2 for Unit 2 method for handling the zeros needed to be developed and applied in the frequency quantifications.

Page 93 of 93 Revision 33 Revision Page 93 of 93

ATTACHMENT (4)

REGULATORY COMMITMENT Calvert Cliffs Nuclear Power Plant September 18, 2014

ATTACHMENT (4)

REGULATORY COMMITMENT The table below lists the action committed to in this submittal. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

Regulatory Commitment Date Complete repairs to address the effects of concrete weathering to July 1, 2015 Unit 1 and Unit 2 containment structure dome area.

Complete repairs to address concrete delamination around the sloped July 1, 2015 surface above the equipment hatch on Unit 1 and Unit 2.

1