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:	Gellrich G H Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML14265A219 (142)

Text

' N George GellrichExeon Generation Site Vice President Calvert Cliffs Nuclear Power Plant1650 Calvert Cliffs ParkwayLusby, MD 20657410 495 5200 Office717 497 3463 Mobilewww.exeloncorp.com george.gellrich@exeloncorp.com 10 CFR 50, Appendix JSeptember 18, 2014U. S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2Renewed Facility Operating License Nos. DPR-53 and DPR-69NRC Docket Nos. 50-317 and 50-318

Subject:

License Amendment Request:

Revise Technical Specification Section 5.5.16 forPermanent 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 forCalvert Cliffs Units No. 1 and 2. The proposed amendment revises Calvert CliffsTechnical Specification 5.5.16, "Containment Leakage Rate Testing Program" to allow forpermanent extensions of the Type A Integrated Leak Rate Testing and Type C Leak RateTesting frequencies.

The proposed amendment and significant hazards discussion are provided inAttachment (1). The marked up page of the affected Technical Specification isprovided 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 therisk impact of the proposed amendment.

A copy of the risk assessment is provided inAttachment (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 animplementation period of 75 days. Approval by this time will allow Calvert Cliffs to avoidperforming 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 March2016.

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

I declare under penalty of perjury that the foregoing is true and correct.September 18, 2014.Respectfully, GHG/KLG/bjd George H. GellrichSite Vice President E. Lauver atExecuted onAttachments:

(1)(2)(3)(4)Evaluation of the Proposed ChangeMarked up Technical Specifications PageEvaluation of Risk Significance of Permanent ILRT Extension Regulatory Commitment cc: NRC Project Manager, Calvert CliffsNRC Regional Administrator, Region INRC Resident Inspector, Calvert CliffsS. Gray, MD-DNR ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGETABLE OF CONTENTS1.0 SUMMARY DESCRIPTION 2.0 DETAILED DESCRIPTION 3.0 TECHNICAL EVALUATION 4.0 REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria

4.2 Precedent

4.3 Significant Hazards Consideration

4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

Calvert Cliffs Nuclear Power PlantSeptember 18, 2014 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGE1.0 SUMMARY DESCRIPTION This evaluation supports a request to amend Operating License Numbers DPR-53 andDPR-69 for Calvert Cliffs Nuclear Power Plant (Calvert Cliffs) Units 1 and 2. The proposedchange would revise the Operating Licenses by amending Technical Specification (TS)Section 5.5.16, Containment Leakage Rate Testing Program.

The proposed change to theTechnical 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 NuclearEnergy Institute (NEI) Topical Report NEI 94-01 Revision 3-A (Reference

2) [NuclearRegulatory Commission (NRC)-approved version specified in the 10 CFR Part 50,Appendix J Program Plan] as the implementation document used by Calvert Cliffs toimplement 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 thelisting of one time exceptions previously granted to Integrated Leak Rate Test (ILRT) testfrequencies and exceptions from post modification ILRTs when Calvert Cliffs replacedSteam Generators.

These exceptions are historical in that the one-time extensions of ILRTtest frequencies have already been used and replacement of both Unit 1 and 2 SteamGenerators have already been completed.

Additional information on these exceptions isprovided in Section 3.2.3 of this document.

The proposed change would allow an increase in the ILRT test interval from its current10 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 inaccordance 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 LicenseAmendment

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, inpart:"A program shall be established to implement the leakage testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B. This programshall be in accordance with the guidelines contained in RG 1.163, "Performance-Based Containment Leak-Test Program,"

dated September 1995, including errata, as modified bythe following exceptions:

a. Nuclear Energy Institute (NEI) 94-01 -1995, Section 9.2.3: The first Unit 1Type 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 theMay 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 testingrequirements associated with steam generator replacement.

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

1 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEThe proposed change to Calvert Cliffs TS 5.5.16, "Containment Leakage Rate TestingProgram" will remove exceptions (a), (b), and (c) and replace the reference to RG 1.163 with areference to NEI Topical Report NEI 94-01 Revision 3-A. The proposed change will reviseTS 5.5.16 to state, in part:"A program shall be established to implement the leakage testing of the containment asrequired by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B. This program shall bein accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline forImplementing 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 fromthe 2016 refueling outage to a subsequent refueling outage (no later than May 3, 2021) whenit can be performed in a refueling outage that involves fewer conflicts with other plannedactivities 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 thanMarch 17, 2028.Attachment (3) contains the plant specific risk assessment conducted to support thisproposed 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 wellwithin 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 lowstrain elastic response such that its behavior will be predictable under all design loadingsand 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 domeconnected to and supported by a reinforced concrete foundation slab. The interior surface ofthe 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 numerousmechanical and electrical systems that penetrate the containment structure wall through weldedsteel 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 designedto withstand the internal pressure and coincident temperature resulting from the energyreleased in the event of the loss-of-coolant accident (LOCA) associated with rated full poweroperation.

The current design conditions for the structure are an internal pressure of 50 psig,a coincident concrete surface temperature of 2760F and a leak rate of 0.16% by weight perday at design temperature and pressure.

2 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGE3.1.1 Containment LinerThe containment liner is a 1/4" thick welded steel plate that is attached to the inside faceof 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.

The1/4"-thick liner plate is attached to the concrete by means of an angle grid system stitchwelded to the liner plate and embedded in the concrete.

The frequent anchoring is designedto prevent significant distortion of the liner plate during accident conditions and to ensure thatthe liner maintains its leak-tight integrity.

The liner plate is protected from corrosion on theinside 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 thatcomes in contact with the concrete.

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

3.1.2 Electrical Penetrations Two types of electrical penetration assemblies are used; canister and unitized header. Allelectrical 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 ofsuitable 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 thepipe and to the liner plate reinforcement.

In the case of piping that carries hot fluid, the pipe isinsulated and cooling is provided to restrict the concrete maximum temperature to 1500F.The anchorage of penetration closure connecting pipes to the containment wall weredesigned as Seismic Category I structures to resist all forces and moments caused by apostulated pipe rupture.

The design conditions include the maximum pipe reactions andpipe rupture forces.The penetration

assembly, consisting of pipe cap and the assembly welds and welds to theliner plate, utilizes full penetration welds. The assembly is anchored into the wall concreteand designed to accommodate all forces and moments due to pipe rupture and thermalexpansion.

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

However, those bellows do not form part of the containment building vessel or pressureboundary.

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 inthe containment structure and the spent fuel pool in the Auxiliary Building.

The penetration 3

ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEconsists of a 36" stainless steel pipe installed inside a 42" pipe sleeve. The inner pipeacts as the refueling tube and is fitted with a gate valve in the spent fuel pool and anencapsulating pipe sleeve, which is welded to the refueling pool liner and sealed off fromthe Containment with a testable double 0-ring blind flange in the refueling pool. Thisarrangement 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 anydifferential movement between the tube and the building structures.

The bellows donot form any part of the containment boundary so they are unaffected by this proposedchange.3.1.6 Moisture BarrierA layer of compressible material covers both sides of the containment liner on thecontainment wall where the finished concrete floor joins the wall. This cork layer, coveredwith a waterproof seal, serves as an expansion joint to accommodate any relativemovement between the containment wall, floor, and liner.3.1.7 Containment TendonsThere are four types of Containment tendons:

Dome, Hoop, Vertical (Original

-Undisturbed),

and Vertical (Replaced or Restressed).

The total active population oftendons 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 steelsheathing.

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 forcorrosion 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 verticaltendons for each unit were re-stressed or replaced with new tendons between 2001 and2002.In the concept of a post-tensioned containment structure, the internal pressure load isbalanced 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 pressureso that a margin of external pressure exists beyond that required to resist the designpressure.

Nominal, bonded reinforcing steel was also provided to distribute strains due toshrinkage 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 externalbearing 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 thestructure.

4 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGE3.2 Justification for the Technical Specification Change3.2.1 Chronology of Testing Requirements of 10 CFR Part 50, Appendix JThe testing requirements of 10 CFR Part 50, Appendix J, provide assurance that leakage fromthe containment, including systems and components that penetrate the containment, does notexceed the allowable leakage values specified in the TS. Title 10 CFR Part 50, Appendix J alsoensures that periodic surveillance of reactor containment penetrations and isolation valves isperformed so that proper maintenance and repairs are made during the service life of thecontainment and the systems and components penetrating primary containment.

The limitation on containment leakage provides assurance that the containment would perform its designfunction following an accident up to and including the plant design basis accident.

Appendix Jidentifies three types of required tests: (1) Type A tests, intended to measure the primarycontainment overall integrated leakage rate; (2) Type B tests, intended to detect local leaks andto measure leakage across pressure-containing or leakage limiting boundaries (other thanvalves) for primary containment penetrations; and (3) Type C tests, intended to measurecontainment isolation valve leakage rates. Type B and C tests identify the vast majority ofpotential containment leakage paths. Type A tests identify the overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure byevaluating 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 forWater-Cooled Power Reactors,"

was amended to provide a performance-based Option B for thecontainment 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 thecomponent and resulting risk from its failure.

The use of the term "performance-based" in10 CFR Part 50, Appendix J refers to both the performance history necessary to extend testintervals 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., twoconsecutive, 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 onan NRC risk assessment contained in NUREG-1493, (Reference

6) and Electric PowerResearch Institute (EPRI) TR-104285 (Reference

7) both of which showed that the riskincrease associated with extending the ILRT surveillance interval was very small. In additionto the 10-year ILRT interval, provisions for extending the test interval an additional 15 monthswas considered in the establishment of the intervals allowed by RG 1.163 and NEI 94-01, butthat this "should be used only in cases where refueling schedules have been changed toaccommodate other factors."

In 2008, NEI 94-01, Revision 2-A, (Reference

8) was issued. This document describes anacceptable approach for implementing the optional performance-based requirements ofOption B to 10 CFR Part 50, Appendix J, subject to the limitations and conditions noted inSection 4.0 of the NRC Safety Evaluation Report (SER) on NEI 94-01. The NRC SER wasincluded 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 theregulatory positions stated in RG 1.163 (September 1995). It delineates a performance-based 5

ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEapproach 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 to10 CFR Part 50, Appendix J and includes provisions for extending Type A ILRT intervals to upto 15 years. Nuclear Energy Institute 94-01 has been endorsed by RG 1.163 and NRC SERsof 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 areincorporated 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 twoconsecutive 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 maximumof 120 months for Type B tests (except for containment airlocks) and up to a maximum of75 months for Type C tests. If a licensee considers extended test intervals of greater than60 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 pastthe 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 beperformed during a period of reactor shutdown at a frequency of at least once per 15 yearsbased on acceptable performance history."

However, Section 9.1 states that the "required surveillance intervals for recommended Type A testing given in this section may beextended by up to 9 months to accommodate unforeseen emergent conditions but shouldnot be used for routine scheduling and planning purposes."

The NRC staff believes thatextensions of the performance-based Type A test interval beyond the required 15 yearsshould be infrequent and used only for compelling reasons.

Therefore, if a licensee wantsto use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have todemonstrate 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 Band 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 RequiredSurveillances, 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 beextended 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) arepermissible only for non-routine emergent conditions.

This provision (nine month6 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEextension) does not apply to valves that are restricted and/or limited to 30 month intervals inSection 10.2 (such as BWR MSIVs) or to valves held to the base interval (30 months) dueto unsatisfactory LLRT performance."

The NRC has also provided the following concerning the extension of ILRT intervals to15 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 theenhanced and robust primary containment inspection program and the local leakage ratetesting of penetrations.

Most of the primary containment leakage experienced has beenattributed to penetration leakage and penetrations are thought to be the most likely locationof 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 tochoose containment leakage testing under either Option A, "Prescriptive Requirements,"

orOption B, "Performance Based Requirements."

On March 13, 1996 the NRC-approved LicenseAmendment No. 212 for Calvert Cliffs Unit 1 and Amendment 189 for Unit 2 authorizing theimplementation of 10 CFR Part 50, Appendix J, Option B for Type A tests. On January 11, 1997the NRC-approved License Amendment No. 219 for Calvert Cliffs Unit 1 and Amendment 196for Unit 2 authorizing the implementation of 10 CFR Part 50, Appendix J, Option B for Type Band Type C tests. Current TS 5.5.16 requires that a program be established to comply with thecontainment 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 inaccordance with the guidelines contained in RG 1.163. Regulatory Guide 1.163 endorses, withcertain exceptions, NEI 94-01 Revision 0 as an acceptable method for complying with theprovisions of Appendix J, Option B.Regulatory Guide 1.163, Section C. 1 states that licensees intending to comply with 10 CFRPart 50, Appendix J, Option B, should establish test intervals based upon the criteria in Section11.0 of NEI 94-01 (Reference

5) rather than using test intervals specified in American NationalStandards Institute (ANSI)/American Nuclear Society (ANS) 56.8-1994.

Nuclear EnergyInstitute 94-01, Section 11.0 refers to Section 9, which states that Type A testing shall beperformed during a period of reactor shutdown at a frequency of at least once per ten yearsbased on acceptable performance history.

Acceptable performance history is defined ascompletion of two consecutive periodic Type A tests where the calculated performance leakagewas 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 todetermine performance shall be at least 24 months.Adoption of the Option B performance based containment leakage rate testing program alteredthe frequency of measuring primary containment leakage in Types A, B, and C tests but did notalter the basic method by which Appendix J leakage testing is performed.

The test frequency isbased on an evaluation of the "as found" leakage history to determine a frequency for leakagetesting which provides assurance that leakage limits will not be exceeded.

The allowedfrequency for Type A testing as documented in NEI 94-01, is based, in part, upon a genericevaluation documented in NUREG-1493.

The evaluation documented in NUREG-1493 includeda 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 that7 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEreducing the frequency of Type A tests (ILRT) from the original three tests per ten years to onetest per 20 years was found to lead to an imperceptible increase in risk. The estimated increasein risk is very small because ILRTs identify only a few potential containment leakage paths thatcannot be identified by Types B and C testing, and the leaks that have been found by Type Atests have been only marginally above existing requirements.

Given the insensitivity of risk tocontainment leakage rate and the small fraction of leakage paths detected solely by Type Atesting, NUREG-1493 concluded that increasing the interval between ILRTs is possible withminimal impact on public risk.3.2.3 Calvert Cliffs 10 CFR Part 50, Appendix J Option B Licensing HistoryMarch 13, 1996The Commission issued on March 13, 1996 Amendment No. 212 to Facility Operating LicenseNo. DPR-53 and Amendment No. 189 to Facility Operating License No. DPR-69 for CalvertCliffs Units 1 and 2, respectively (Reference 11). The amendment revised TSs to reflect theapproval 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, 1997The 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 forCalvert Cliffs Units 1 and 2 (Reference 12).The amendments adopted Option B of 10 CFR Part 50, Appendix J, approving Type B andType C containment leakage testing to be performed on a performance-based testing schedule.

May 1, 2002The 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 theperformance of the next Unit 1 ILRT following the June 15, 1992 Type A test. This test was tobe performed no later than June 14, 2007 (ILRT was conducted on May 3, 2003). Theamendment also exempted Calvert Cliffs Unit 1 from the requirement to perform a post-modification containment ILRT associated with steam generator replacement.

The Calvert CliffsUnit 1 steam generators were replaced during the 2002 refueling outage.June 27, 2002The 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 steamgenerators.

The Calvert Cliffs Unit 2 steam generators were replaced during the 2003 refueling outage.8 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEAugust 29, 2007The Commission issued on August 29, 2007 Amendment No. 281 to Renewed FacilityOperating 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), atcontainment peak pressure, to a value of 0.16% per day as expressed in TS 5.5.16.March 22, 2011The 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 revisedTS 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 toperform its next ILRT no later than May 1, 2016 (ILRT was performed on March 17, 2013).July 31, 2013The 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 internalpressure (Pa) from 49.4 pounds per square inch gauge (psig) to 49.7 psig for the design basisLOCA. In support of the revised Pa, the amendment also revised TS 3.6.4 by decreasing theupper 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 inaccordance with RG 1.163, which endorses the methodology for complying with Option B. Theperformance leakage rates are calculated in accordance with NEI 94-01, Section 9.1.1 forType A testing.

The performance leakage rate includes the Type A Upper Confidence Limit at95% plus the as-left minimum pathway leakage rate for all Types B and C pathways not inservice,

isolated, or not lined up in their test position.

Tables 3.2-1 and 3.2-2 list the Type AILRT past results for Units 1 and 2, respectively.

Table 3.2-1, Unit I Type A ILRT HistoryLeakage Rate (1) As Left Type C Minimum(Containment air Path Contribution Test Date (weight %/day) (weight %/day)12/01/1973 0.0466 0.0003/06/1978 0.136 0.02806/22/1982 0.0514 0.025405/20/1985 0.032 0.00205/27/1988 0.036 0.006The results of the last two Type A ILRTs for Calvert Cliffs Unit 1:07/05/1992 0.1564(3) 0.0742105/03/2006 0.09515 0.001689 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGETable 3.2-2, Unit 2 Type A ILRT HistoryLeakage Rate (1) As Left Type C Minimum(Containment air Path Contribution Test Date (weight %/day) (weight %/day)03/14/1976 0.019 0.000111/15/1979 0.052 0.0008412/22/1982 0.025 0.001511/24/1985 0.142(2) 0.08101/16/1991 0.061 0.001The results of the last two Type A ILRTs for Calvert Cliffs Unit 2:05/02/2001 0.0738 0.00103/17/2013 0.0802 0.0025(1) On August 29, 2007 Calvert Cliffs design basis containment leak rate, La, was changedfrom a value of 0.2 wt%/day at containment peak pressure, to a value of 0.16 wt%/day asexpressed in TS 5.5.16.(2) Local Leakage Rate Test, repair, and adjustments of containment isolation valves wasperformed prior to the 11/24/1985 ILRT. The minimum pathway leakage improvement dueto repairs and adjustments was 140,591.24 sccm or 0.081 wt%/day.

The as foundcontainment leakage rates, the sum of the minimum pathway leakage improvement andILRT 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 overallperformance of the containment totaling 123,236 sccm, or 0.07421 wt%/day.

The adjustedILRT 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 thetechnical 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 thecurrent maximum allowable containment leakage rate of 0.16 wt%/day at the test pressure of50 psig. As a result, since both tests for both units were successful, the current ILRT intervalfrequency for Calvert Cliffs Units I and 2 are ten years.3.3 Plant Specific Confirmatory Analysis3.3.1 Methodology An evaluation has been performed to assess the risk impact of extending the Calvert CliffsUnits 1 and 2 ILRT intervals from the current 10 years to 15 years. The purpose of this analysisis 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-TimeExtensions for Containment Integrated Leakage Rate Test Surveillance Intervals" fromNovember 2001,10 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGE" NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated inRG 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 ofcorrosion induced leakage of steel liners going undetected during the extended testinterval,

" Methodology used in EPRI Report No. 1009325, Revision 2-A (Reference 18), themethodology 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 amendtheir TS to extend the ILRT surveillance interval to 15 years, subject to the limitations andconditions 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)1. The licensee submits documentation indicating that the technical adequacy oftheir PRA is consistent with therequirements of RG 1.200 relevant to theILRT extension

2. The licensee submits documentation indicating that the estimated risk increaseassociated with permanently extending theILRT surveillance interval to 15 years issmall, and consistent with the clarification provided in Section 3.2.4.5 of this SE.Specifically, a small increase in population dose should be defined as an increase inpopulation dose of less than or equal toeither 1.0 person-rem per year or 1 % ofthe total population dose, whichever isrestrictive.

In addition, a small increase in CCFPshould be defined as a value marginally greater than that accepted in a previousone-time ILRT extension requests.

Thiswould require that the increase in CCFPbe less than or equal to 1.5 percentage point.Calvert Cliffs ResponseCalvert Cliffs PRA quality is addressed inSection 3.3.2 and Attachment (3), "CalvertCliffs Nuclear Power Plant: Evaluation ofRisk Significance of Permanent ILRTExtension" Attachment 1, "PRA QualityStatement for Permanent 15-Year ILRTExtension" EPRI Report No. 1009325, Revision 2-A,incorporates these population dose andCCFP acceptance guidelines, and theseguidelines have been used for the CalvertCliffs plant specific assessments.

The increase in population dose is0.20 person-rem/year for Unit 1 and0.11 person-rem/year for Unit 2.The increase in CCFP is 0.558% for Unit 1and 0.490% for Unit 2.11 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGELimitation/Condition (From Section 4.2 of SE)3. The methodology in EPRI Report No.1009325, Revision 2, is acceptable exceptfor the calculation of the increase inexpected population dose (per year ofreactor operation).

In order to make themethodology acceptable, the average leakrate accident case (accident case 3b)used by the licensees shall be 100 Lainstead of 35 La.4. A licensee amendment request (LAR) isrequired in instances where containment over-pressure is relied upon foremergency core cooling system (ECCS)performance Calvert Cliffs ResponseEPRI Report No. 1009325, Revision 2-A,incorporated the use of 100 La as the averageleak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has been used in theCalvert Cliffs plant specific risk assessment.

For Calvert Cliffs, containment over-pressure is NOT relied upon for emergency corecooling 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 forthis analysis.

An independent PRA peer review was conducted under the auspices of thePressurized Water Reactor Owners Group in June of 2010, and was performed against theguidance of RG 1.200, Revision 2, and requirements of ASME/ANS RA-Sa-2009.

The scope ofthe 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 notsignificant using a sensitivity study) have been captured in the PRA Configuration RiskManagement Program (CRMP) database.

On an on-going basis, other potential PRA modeland documentation changes are captured and prioritized in the CRMP database.

The Calvert Cliffs Internal Events model was also updated to support the Calvert Cliffs FirePRA.The Calvert Cliffs Internal Events model was peer reviewed in June 2010. All findings, whichhad significant impact on this analysis, have been addressed.

This assessment is provided inAttachment (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 therequirement for numerical results for CDF and LERF to determine the risk impact of therequested change and the fact that this change is risk-informed, not risk-based.

Table 1includes discussion of all findings from the industry-peer review along with the assessment andevaluation of the finding that shows that they have either been addressed or have no materialimpact 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 notedas Category I. As noted in the peer review report, the majority of the findings weredocumentation related.

Of the 11 SRs, which did not meet Category II or better, seven wererelated to conservatisms or documentation in LERF and two were related to internal floods.12 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEThere 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 thereare 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 theNEI 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 ofthe Fire PRA for the spectrum of potential risk-informed plant licensing applications for whichthe Fire PRA may be used. The 2012 Calvert Fire PRA peer review was a full-scope review ofall of the technical elements of the Calvert Cliffs at-power FPRA (2012 model of record) againstall technical elements in Section 4 of the ASME/ANS Combined PRA Standard, including thereferenced 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. Allfindings are being provided and have been dispositioned.

All F&Os that were defined assuggestions have been dispositioned and will be available for NRC review. The Fire PRA isadequate to support the ILRT extension.

The Calvert Cliffs seismic PRA model is relatively conservative and, other than the highmagnitude acceleration event, is not a dominant contributor.

The Calvert Cliffs high winds PRAmodel is very conservative in the tornado area in that all tornados are grouped into the mostconservative 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 riskassessment of tornado missile impacts and hurricane force winds. Although this model has notbeen peer reviewed in compliance with the ASME/ANS RA-Sa-2009

standard, the model isbased upon accepted methodology and utilizes the ASME/ANS RA-Sa-2009 compliant internalevents model. High winds updates are not expected to cause a significant increase in CDF orLERF. A more detailed assessment would be expected to cause a decrease in CDF.3.3.3 Summary of Plant-Specific Risk Assessment ResultsThe findings of the Calvert Cliffs Unit 1 and 2 Risk Assessment confirm the general findings ofprevious studies that the risk impact associated with extending the ILRT interval from three inten years to one in 15 years is small. The Calvert Cliffs plant-specific results for extending ILRTinterval 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" thefollowing conclusions regarding the assessment of the plant risk associated with extending theType 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 thelicensing basis. RG 1.174 defines very small changes in risk as resulting in increases ofCDF less than 1.OE-06/year and increases in LERF less than 1.OE-07/year.

Since the ILRTdoes not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from achange in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as4.54E-8/year for Unit 1 and 2.46E-8/year for Unit 2 using the EPRI guidance.

As such, theestimated change in LERF is determined to be "very small" for both units using theacceptance 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, measuredas an increase to the total integrated plant risk for those accident sequences influenced byType 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 definedas 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 impactfor 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 to1 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, thisincrease is judged to be very small.Calvert Cliffs is transitioning to NFPA 805 licensing basis for fire protection and submitted aLicense 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 coredamage 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 areimplemented.

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 theNFPA 805 modifications will be completed by the end of Unit 1 2016 refueling outage. Riskmitigation strategies will be in place for any open modifications.

These strategies may beactions 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 releaseaccident sequences.

The next Unit 2 ILRT is scheduled for 2023, so the NFPA 805modifications will be implemented prior to the extension.

• An assessment of the impact of external events was performed using fire risk analysis fromthe Fire PRA. The total LERF value for Unit 1 is 6.OOE-6/yr for Unit 1 and 7.38E-6/yr forUnit 2. Since the total LERF for both units is less than 1.OE-5, it is acceptable for theALERF to be between 1.OE-7 and 1.OE-6." An assessment of the impact of external events was also performed using fire risk analysisfrom the Individual Plant Examination of External Events (IPEEE) rather than the Fire PRAmodel. 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 bebetween 1.OE-7 and 1.OE-6.Therefore, increasing the ILRT interval to 15 years is considered to be insignificant since itrepresents 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 wasfound to lead to an imperceptible increase in risk. The estimated increase in risk is verysmall because ILRTs identify only a few potential containment leakage paths that cannot beidentified by Type B or Type C testing, and the leaks that have been found by Type A testshave been only marginally above existing requirements.

14 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEGiven the insensitivity of risk to containment leakage rate and the small fraction of leakagepaths 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 ILRTfrequency 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 basisconsidering 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 thissubmittal.

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

CalvertCliffs 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 ProgramThe requirements of 10 CFR Part 50, Appendix B are implemented through specification ofappropriate technical and quality requirements for the Service Level 1 coatings program whichincludes ongoing maintenance activities.

Calvert Cliffs has implemented controls for theprocurement, application, and maintenance of Service Level I protective coatings used insidethe 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 aspart of the safety-related and controlled protective coatings program.

Inspections of coatingssystems are scheduled every outage on a pre-established basis to verify containment linercoating thickness and condition.

This program also satisfies the License Renewal Application commitment contained inTable 16-2 of Calvert Cliffs Updated Final Safety Analysis Report (UFSAR),

as managinggeneral corrosion of the containment wall and dome liner plates.3.4.2 Containment Inservice Inspection ProgramThe purpose of the Calvert Cliffs Containment Inservice Inspection (CISI) program is toperiodically perform destructive and nondestructive examination of ASME Class MC and CCcomponents 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 beendeveloped to comply with ASME Section Xl 2004 Edition, except where specific writtenalternatives from Code requirements have been requested by Calvert Cliffs and granted by theNRC 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, asapplicable.

The components subject to the requirements of this CISI program are those which make up thecontainment structure, its leak tight barrier (including integral attachments) and those whichcontribute to its structural integrity, specifically, Class MC pressure-retaining components, andtheir 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 CISIportion of the Calvert Cliffs Ten-year Inservice Inspection (ISI) program.

The Second IntervalCISI 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 alldates associated with the Third Interval are postulated.

IWE(Class MC) Inspection Interval and PeriodsThe 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 endon September 9, 2018. This interval was shortened as a result of extending the first ten-yearCISI 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 bothUnits complies with IWL-2400 and is effective for IWL inspections conducted betweenSeptember 9, 2009 and September 9, 2018.Concrete examinations shall be conducted every five years (+/- one year) as described inIWL 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 thistime 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 Unitscomplies 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 betweenSeptember 9, 2018 and September 9, 2028.16 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEIWL (Class CC) Inspection Periods (Tendons)

For multiple-unit plant sites, such as Calvert Cliffs, the tendon examination frequency may beextended 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 testperformed within two years of one another, and experience similar environmental exposure.

The examinations required by IWL-2500 for unbonded post-tensioning systems can thenalternate between the two units every five years, as allowed by IWL-2421 (sites with multipleunits).For Calvert Cliffs Unit 1, all examinations required by IWL-2500 (Items L2.10 thru L2.50) shallbe 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) shallbe performed at 1, 5, and 15 years and every 10 years thereafter (25, 35, 45, 55 years). Onlythe 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 thesubsequent interval for both units and encompasses the first extended interval ILRT testing:Table 3.4-1, Calvert Cliffs IWL (Concrete)

Examination Periods and ScheduleUnit 1 and 2Period Specified Date Tolerance 10 year 8/14/2011

+/-1 year15 year 8/14/2016

+/- 1 year20 year 8/14/2021

+/- 1 year25 year 8/14/2026

+/- 1 year30 year 8/14/2031

+/- 1 yearTable 3.4-2, Calvert Cliffs IWL (Tendons)

Examination Periods and ScheduleUnit 1 Unit 2Period Specified Date Tolerance Period Specified Date Tolerance 35 Year 9/9/2011

+/- 1 Year 35 Year 9/9/2012

+/- 1 Year40 Year 9/9/2016

+/- 1 Year 40 Year 9/9/2017

+/- 1 Year45 Year 9/9/2021

+/- 1 Year 45 Year 9/9/2022

+/- 1 Year50 Year 9/9/2026

+/- 1 Year 50 Year 9/9/2027

+/- 1 Year55 Year 9/9/2031

+/- 1 Year 55 Year 9/9/2032

+/- 1 Year17 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGETable 3.4-3, Calvert Cliffs Unit I and 2 IWEIIWL Interval,

Periods, and OutagesUnit 1 Unit 2Period Period Refuel Period Period RefuelInspection Start End Refuel Outage Start End Refuel OutagePeriods Dates Dates Outage Year Dates Dates Outage Year1 July 1, June 30, RFO-20 2010 July 1, June 30, RFO-19 20112009 2012 RFO-21 2012 2009 20122 July 1, June 30, RFO-22 2014 July 1, June 30, RFO-20 20132012 2016 RFO-23 2016 2012 2016 RFO-21 20153 July 1, June 30, RFO-24 2018 July 1, June 30, RFO-22 20172016 2019 2016 2019 RFO-23 20191 July 1, June 30, RFO-25 2020 July 1, June 30, RFO-24 20212019 2022 RFO-26 2022 2019 20222 July 1, June 30, RFO-27 2024 July 1, June 30, RFO-25 20232022 2026 RFO-28 2026 2022 2026 RFO-26 20253 July 1, June 30, RFO-29 2028 July 1, June 30, RFO-27 20272026 2029 2026 2029 RFO-28 2029Adoption of Code CasesAll Code Cases adopted for ASME Section Xl activities for use during the second ten-yearcontainment ISI interval are listed below. The use of Code Cases is in accordance with ASMESection Xl, IWA-2440, 10 CFR 50.55a, and RG 1.147 (Reference 20). As permitted by ASMESection Xl and RG 1.14 7 or 10 CFR 50.55a, ASME Section Xl Code Cases may be adoptedand used as described below:Code Cases Adopted from RG 1.147N-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-3N-739 Alternative Qualification Requirements for Personnel Performing Class CC Concreteand Post-Tensioning System Visual Examinations.

This fulfills NRC concernsstated in 10 CFR 50.55a(b)(2)(ix)(F) regarding "owner-defined" personnel qualifications Relief RequestsTable 3.4-4 contains an index of Requests for Alternatives and Requests for Relief written inaccordance with 10 CFR 50.55a(a)(3) and (g)(5). The applicable submittal and NRC SERcorrespondence numbers are also included in Table 3.4-4 for each request for alternative andrequest for relief. Note that only Requests for Alternatives or Requests for Relief applicable tothe requirements for Class MC and CC components are addressed in Table 3.4-4.Table 3.4-4, Second Ten-year CISI Interval Relief RequestsRelief Code Case Relief Request Licensee NRC SERRequest 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) Case18 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEN-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," forthe near-distance acuity testing requirements.

Code Case N-753 provides an alternative to the visual acuity demonstration requirements ofIWA-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 forgeneral use by reference in RG 1.147, "Inservice Inspection Code Case Acceptability."

Afterthat time, if the licensee wishes to continue to use Code Case N-753, the licensee must followall conditions and limitations place on the use of the code case, if any, that are specified inRG 1.147.Component Exemptions, IWE and IWLThe basis for the selection of components at Calvert Cliffs which are determined to be within thescope of the required examinations was done in accordance with the requirements of IWE-1 220and IWL-1220 respectively.

Calvert Cliffs does have areas that are considered inaccessible which are therefore exemptfrom inspection and are described below:" IWE -The containment liner covering the containment foundation slab is inaccessible.

Thisarea is covered with the finished concrete floor and moisture barrier and accounts forapproximately 15% of the containment liner surface area." IWL -Portions of the concrete surface that are covered by the liner, foundation material orbackfill, or are otherwise obstructed by adjacent structures, components, parts orappurtenances are inaccessible.

The entire inside concrete surface of the Calvert Cliffscontainment buildings area covered in steel, which makes them inaccessible forexamination.

Examination Methods & Personnel Qualifications The examination methods used to perform Code examinations for the nonexempt Class MC andCC components are in accordance with 10 CFR 50.55a requirements and the applicable ASMECodes.Personnel performing IWE examinations shall be qualified in accordance with Exelon's writtenpractice, or approved vendor written practice for certification and qualification of nondestructive examination personnel.

Personnel performing IWL examinations shall be qualified in accordance with writtenprocedures prepared as required by IWL-2300, as modified by applicable Code Cases.19 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEInaccessible AreasFor Class MC applications, Calvert Cliffs shall evaluate the acceptability of inaccessible areaswhen conditions exist in accessible areas that could indicate the presence of or result indegradation to such inaccessible areas. For each inaccessible area identified, Calvert Cliffsshall provide the following in the Owners Activity Report-I, as required by 10 CFR50.55a(b)(2)(ix)(A):

" A description of the type and estimated extent of degradation, and the conditions that led tothe 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 anyinaccessible areas at this time. However, Exelon actively participates in various nuclear utilityowners groups and ASME Code committees to maintain cognizance of ongoing developments within the nuclear industry.

Industry operating experience is also continuously reviewed todetermine its applicability to Calvert Cliffs. Adjustments to inspection plans and availability ofnew, commercially available technologies for the examination of the inaccessible areas of thecontainment 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 ofClass 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 toaccelerated 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 theaugmented components are identified in Table 3.4-5.Table 3.4-5, Augmented Scope -Containment Interior Visible SurfacesUnit 1 Unit 2Item Summary Item SummaryNo. No. Description No. No. Description E4.11 A27601 Liner bulging -liner 176 E4.11 B118221 Liner indentation

-liner 221E4.11 A28901 Liner bulging -liner 189 E4.11 B130254 Bulging -liner 254E4.11 A29001 Liner bulging -liner 190 E4.11 B45153 Bulging -liner 153E4.11 A29201 Liner bulging -liner 192 E4.11 B45154 Bulging -liner 154E4.11 A30501 Liner bulging -liner 205 E4.11 B50000 Moisture barrierE4.12 A27602 Liner bulging -liner 176 E4.11 B69179 Liner bulging -liner 179E4.12 A28901 Liner bulging -liner 189 E4.11 B69187 Liner bulging -liner 187E4.12 A29002 Liner bulging -liner 190 E4.11 B69190 Liner bulging -liner 190E4.12 A29202 Liner bulging -liner 192 E4.11 B69195 Liner bulging -liner 195E4.12 A30502 Liner bulging -liner 205 E4.11 B69198 Liner bulging -liner 198Examination of each item number is required each period until the areas remain essentially unchanged for one period.20 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGE3.4.3 Supplemental Inspection Requirements With the implementation of the proposed change, TS 5.5.16 will be revised by replacing thereference to RG 1.163 with reference to NEI 94-01, Revision 3-A. This will require that ageneral visual examination of accessible interior and exterior surfaces of the containment forstructural deterioration that may affect the containment leak-tight integrity be conducted.

Thisinspection must be conducted prior to each Type A test and during at least three other outagesbefore the next Type A test if the interval for the Type A test has been extended to 15 years inaccordance 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 surfacesof the primary containment to identify evidence of structural deterioration, which could affecteither 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 theinterior and exterior surfaces of the containments.

Personnel performing STP-M-665-1 andSTP-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 asfollows:* 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 priorto each Type A test." Containment Concrete

-scheduled for inspection every 36 +/- 14 months and prior to everyType A test.Performance of these tests are also listed in Table 16-2 of Calvert Cliffs UFSAR as part of ourLicense 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 alsosatisfy the inspection requirements of NEI 94-01 Revision 3-A.3.4.4 Containment Leakage Rate Testing Program -Type B and Type C Testing ProgramCalvert 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 servicelife. The Types B and C testing program provides a means to protect the health and safety ofplant personnel and the public by maintaining leakage from these components belowappropriate limits. Per TS 5.5.16, the allowable maximum pathway total Types B and C leakageis 0.6 La where La equals approximately 276,800 sccm.21 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEAs discussed in NUREG-1493, Type B and Type C tests can identify the vast majority of allpotential containment leakage paths. Type B and Type C testing will continue to provide a highdegree 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 of2014 for Calvert Cliffs Unit 1 and from the Spring of 2003 through the Spring of 2013 has shownan exceptional amount of margin between the actual As-Found (AF) and As-left (AL) outagesummations and the regulatory requirements as described below:" The As-Found minimum pathway leak rate average for Calvert Cliffs Unit 1 shows anaverage 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 averageof 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 anaverage 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 averageof 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 CalvertCliffs 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 theTS 5.5.16 limit of 0.6 La (166,080 sccm) and demonstrates a history of successful tests. TheAs-Found (AF) minimum pathway summations represent the high quality of maintenance ofType 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 theprogram owner.Table 3.4-6, Unit 1 Type B and C LLRT Combined As-Found/As-Left Trend SummaryRFO 2008 2010 2012 2014AF Min Path (sccm) 11546.9 9949.8 4862.99(1) 14197.19Fraction of La 0.042 0.036 0.017 0.051AL Max Path (sccm) 9677.8 5737.49 17506.39 13234.94Fraction of La 0.035 0.021 0.063 0.048AL Min Path (sccm) 8003.7 4358.99 16510.19(1) 11885.74Fraction of La 0.029 0.016 0.060 0.043Table 3.4-7, Unit 2 Type B and C LLRT Combined As-Found/As-Left Trend SummaryRFO 2003 2005 2007 2009 2011 2013AF Mi Path (sccm) 10535.95(2) 14380.4 10689.8 15772.6 9660.7 20638.99Fraction of La 0.030 0.042 0.031 0.057 0.035 0.074AL Max Path (sccm) 12347.1 3848.9 13936.6 15668.4 11723.7 13810.99Fraction of La 0.036 0.011 0.050 0.057 0.042 0.050AL Min Path (sccm) 11091.6(2) 2784.9 9070.2 10727.50 9250.7 11797.69Fraction of La 0.032 0.008 0.026 0.039 0.033 0.04322 ATTACHMENT (1)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 withinadministrative limits. Following maintenance of 1CV632, subsequent AL testing resulted ina 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, againexceeding 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 themaximum limit of 10,000 sccm. The AL leakage was accepted-as-is.

AF/AL testingperformed in 2014 measured a leakage rate of 1,321 sccm, again exceeding the adminlimit. CR-2014-001688 was initiated.

Penetration 41 (1 MOV651, 1 MOV652) 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 themaximum 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 ameasured leakage rate of 14,790 sccm in 2010. Subsequent AF/AL testing performed in2012 resulted in a leakage rate of 4,808.2 sccm. This was below the administrative limit of8,000 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 thelargest contributor to the identified mismatch between AF and AL minimum pathwayleakage in 2012.Penetration 69 (Emergency Airlock) was AF/AL tested within administrative limits with ameasured leakage rate of 878.4 sccm in 2010. Subsequent AF/AL testing performed in2012 resulted in a leakage rate of 1,849.29 sccm. This was below the administrative limit of8,000 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 21SG), with details regarding these as follows:Penetration 2B (2CVC435)

-Unit 2 check valve CVC435 was AF tested withinadministrative leakage limits during the 2003 Refueling Outage (RFO). Subsequently, thevalve was replaced during the 2003 RFO with an equivalent approved Enertech nozzle typecheck valve under a Calvert Cliffs work order. Subsequent AL testing of the new checkvalve resulted in a leakage rate of 1,488 sccm. This exceeded the administrative leakagelimit of 296 sccm, but was not in excess of the maximum limit of 10,000 sccm.23 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEPenetration 21 SG (21 SG South Manway) -21 SG South Manway AF leakage rate wasmeasured 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 manwaycovers were reinstalled on the newly Steam Generator and this manway penetration wasAL 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, 2009RFOs for eddy current and visual inspections when necessary, and continued to remainwithin administrative leakage limits when tested each outage from 2005 through 2013. The2003 RFO AF versus AL leakage rate differences can be directly attributed to removal andreinstallation 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 theContainment 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 SealInspections Inspections on the containment liner are conducted in accordance with Examination CategoryE-A of the ASME Code Section Xl, 2004 Edition.

These inspections are performed such that100% of the accessible portion of the liner is inspected during each inspection period. Aspreviously mentioned the portion of the liner that covers the containment foundation slab isconsidered inaccessible.

Since this inaccessible area cannot be inspected, Calvert Cliffs musttherefore evaluate its acceptability whenever conditions exist in the accessible areas that couldindicate 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 eachinspection period.Inspection ResultsIn 1994 Calvert Cliffs discovered significant age related degradation of the Unit 1 moisturebarrier seal. As part of the corrective

actions, a decision was made to subsequently replace theUnit 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 thatexceeded 10% of the nominal wall thickness of W." The liner area of concern was the wall tofloor transition under the moisture barrier seal, between the wearing floor slab and thecontainment liner wall.24 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEAn evaluation was subsequently performed which determined that if the degradation was notstopped, but instead continued at its current rate, the pitted areas would degrade further andpose a concern in the future. While it was impossible to determine when the pitting began, itwas reasonable to assume that the degradation would be stopped or significantly slowed by thereplacement of the moisture barrier seal.As part of the evaluation, consideration was given as to whether additional areas needed to beexamined.

A determination was made that no additional examinations of other areas werenecessary.

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 withoutrepair of the degraded area since the area of degradation is non-structural in nature and has noeffect on the structural integrity of Containment.

The replacement of the moisture barrier seal involved use of a new seal material (high densitysilicone elastomer (HDSE)) that provides an effective seal against water, smoke, gas, andpressure.

Along with the installation of the new HDSE sealant, a modification to the design ofthe seal was done. The original base sealant was applied to a shallow depth at the top of thecompressible material in the joints and made flush with the nominal base slab. The new HDSEsealant was installed in such a manner to form a small curb above the joint, which would shedwater in addition to providing a seal. Also, to improve the seal, the HDSE was placed aminimum 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 toseparate 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 thereplacement of the moisture barrier seal has arrested the corrosion and pitting throughout theaffected area and has prevented any new areas of corrosion and pitting from occurring.

As aresult the liner continues to be acceptable to perform its safety function (i.e., act as a leak tightmembrane).

The Unit 2 moisture barrier continues to be subject to Augmented Inspections dueto the identification of a crack and subsequent repair of the seal during the 2013 refueling outage.3.5.2 Containment ConcreteConcrete 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 ResultsDuring 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.

25 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEThe 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. Theoccurrence of freeze-thaw cycles accelerates this process by breaking up the concretesurface.

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 minimizemoisture penetration.

Completion of these actions for Unit 1 and 2 Containments arescheduled 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 concreteto fall off. The proposed corrective action for this issue involves the removal of looseconcrete and the application of an epoxy-bonding compound to which low slump 5000 psiconcrete will be applied.

Completion of these actions for Unit 1 and 2 Containments arescheduled for July 1, 2015.An evaluation of these two issues determined the concrete in those areas is still capable ofmaintaining its structural integrity in the event of a design basis LOCA and that it will continue toperform this function beyond the completion date for the repairs.3.5.3 Containment TendonsContainment Tendon Inspections The containment tendons are inspected in accordance with Examination Category L-B of theASME 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 VerticalVertical (Replaced orUnit Dome Hoop (Original)

Restressed)

Total1 204 465 123 77 8711 0 3 2 0 5(Abandoned) 2 204 468 125 79 876The ASME required tendon lift-off test is conducted on a minimum of 25 tendons once everyten-years.

Per the ASME Code, a sample of each of the four types of active tendons must beexamined.

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 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEA common tendon is a tendon that is tested each time the test is performed.

A random tendonis a tendon selected for this test and is not selected again in subsequent tests. The replaced orre-stressed tendons do not have a common tendon because these tendons were recentlyreplaced as a result of the tendon issues discussed below.The ASME required wire removal tensile test is conducted once every ten years on a minimumof one tendon from each of the four tendon types.The ASME required visual examinations of the tendon anchorage area, the free water analysisand the analysis of the corrosion protection medium (grease analysis) are performed on aminimum of 25 tendons every five years. Each of these three tests is performed on tendonsthat are selected so as to have the same distribution between the four tendon types as for thetendon lift-off tests.Inspection ResultsIn 1997, during the performance of the 20-year (time from first tendon inspection) tendonsurveillance on Unit 1, conditions that did not meet the acceptance standards were found onsome of the Containment tendons.

Conditions that did not meet the acceptance standards werefound in all three containment tendon populations, i.e., hoop, dome, and vertical tendons.

Theabnormal conditions found on the hoop and dome tendons were considered minor enough thatthe acceptability of the concrete containment was not affected.

However the conditions foundon the vertical tendon population were more significant.

Several of the vertical tendons selectedfor the surveillance were found to contain broken and corroded wires at their top ends, justbelow the stressing washer. The discovery of broken wires in these tendons initiated anexpansion of the Unit 1 vertical tendon inspection scope to perform visual inspections and lift-offtesting on all Unit 1 vertical tendons.

Subsequently, broken and corroded wires were foundthroughout 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 wasconducted.

Although Unit 2 was only required to perform visual inspections, it was decided toalso perform lift-off testing of all the vertical tendons in order to facilitate inspection of the tendonwires 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 waswritten 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 the1997 surveillance on the Unit 1 and 2 Containments, an evaluation was conducted.

Theevaluation concluded that the tendon wire failures and corrosion problems resulted from acombination 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 actionswere 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 ventpath to help eliminate voids, re-greasing non-corroded vertical

tendons, and resealing aroundthe original tendon can all-thread penetrations with caulking.

Additional inspections wereperformed 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.27 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEThe goal of the long-term corrective action plan was to ensure that the Containments meet theirdesign basis requirements until plant end-of-life.

As one part of the long-term corrective actionplan, 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 thetendons with less severe corrosion that were not replaced were re-greased during 2001. Theremaining vertical tendon population (46 tendons per Unit) were replaced in 2001 and 2002, andhad new grease put in place at that time. At the end of these corrective

actions, all of thevertical 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 alsoreplaced with a new redesigned pressure-tight grease cap. The redesigned grease cap has aflange that is attached by studs and nuts to the tendon bearing plates by utilizing existing taps inthe plates.Enhanced Vertical Tendon Inspections To further confirm the effectiveness of the short- and long-term corrective

actions, an enhancedinspection program was initiated that consisted of a two-tiered approach.

The first tier involvedthe 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 verticaltendons 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 todetermine 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 enhancedinspections were conducted in 2007. No new issues were identified as a result of theseinspections.

Based on the satisfactory performance of the enhanced inspections, an assessment wasconducted 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 ResultsThe evaluation of the in-service inspection results for the 35th year, conducted in 2012 for Unit 1and 2013 for Unit 2 containment structures, concluded that no abnormal degradation of thepost-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 aremanaged through the conduct of various surveillance tests, in-service inspections, preventive maintenance activities, and maintenance procedures.

These documents will continue to bemodified as necessary to ensure they continue to provide reasonable assurance that the agingeffects will be adequately managed throughout the operating life of the units.28 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGE3.7 NRC SER Limitations and Conditions 3.7.1 Limitations and Conditions Applicable to NEI 94-01 Revision 2-AThe 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 intervalto 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 ResponseFor calculating the Type A leakage rate, the Calvert Cliffs will utilize the definition in NEIlicensee should use the definition in the NEI 94-01 Revision 3-A, Section 5.0. ThisTR94-01, Revision 2, in lieu of that in definition has remained unchanged fromANSI/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 thiscontainment inspections to be performed submittal.

prior to and between Type A tests. (Refer to In addition to the scheduled Containment 151SE Section 3.1.1.3).

inspections, general visual observations ofthe accessible interior and exterior surfacesof the containment structure shall continue tobe performed in accordance withSTP-M-665-1 and STP-M-665-2 "Containment Visual Inspection."

Containment Liner -scheduled forinspection during each refueling outage inaccordance with License RenewalCommitment and prior to each Type Atest.Containment Concrete

-scheduled forinspection every 36 +/- 14 months andprior to every Type A test.These are scheduled surveillance tests andare performed to meet the requirements ofTS 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 besatisfied.

The licensee addresses the areas of the Reference Sections 3.4.2 and 3.4.3 of thiscontainment 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 utilizedto perform visual inspection of the normallyaccessible internal and exterior surfaces ofthe primary containment to identify evidenceof structural deterioration, which could affect29 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGELimitation/Condition (From Section 4.0 of SE) Calvert Cliffs Responseeither structural integrity or leak tightness.

These are scheduled surveillance test andare 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, asapplicable.

(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 ofless than 15 years. If a licensee has to utilize NEI 94-01 Revision 3-A, Section 9.1. Thisthe provision of Section 9.1 of NEI TR 94-01, requirement has remained unchanged fromRevision 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 94-01demonstrate to the NRC staff that it is an Revision 2-A, SER Section 3.1.1.2, Calvertunforeseen emergent condition..

(Refer to SE Cliffs will also demonstrate to the NRC staffSection 3.1.1.2).

that an unforeseen emergent condition existsin the event an extension beyond the 15-yearinterval is required.

For plants licensed under 10 CFR Part 52, Not applicable.

Calvert Cliffs was not licensedapplications requesting a permanent under 10 CFR Part 52.extension of the ILRT surveillance interval to15 years should be deferred until after theconstruction and testing of containments forthat design have been completed andapplicants have confirmed the applicability ofNEI 94-01, Revision 2, and EPRI Report No.1009325, Revision 2, including the use ofpast containment ILRT data.3.7.2 Limitations and Conditions Applicable to NEI 94-01 Revision 3-AThe NRC staff found that the guidance in NEI TR 94-01, Revision 3, was acceptable forreferencing 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 onthe 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 INEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTsbe 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 LLRTsbe increased to 75 months with the requirement that a licensee's post-outage report include themargin between the Type B and Type C leakage rate summation and its regulatory limit. Inaddition, 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 applied30 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEto 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 heldto 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 1Condition 1 presents three (3) separate issues that are required to be addressed.

They are asfollows:" ISSUE 1 -The allowance of an extended interval for Type C LLRTs of 75 months carriesthe requirement that a licensee's post-outage report include the margin between the Type Band Type C leakage rate summation and its regulatory limit." ISSUE 2 -In addition, a corrective action plan shall be developed to restore the margin toan acceptable level." ISSUE 3 -Use of the allowed 9-month extension for eligible Type C valves is onlyauthorized for non-routine emergent conditions.

Response to Condition 1, Issue 1The post-outage report shall include the margin between the Type B and Type C MinimumPathway Leak Rate (MNPLR) summation value, as adjusted to include the estimate ofapplicable Type C leakage understatement, and its regulatory limit of 0.60 La.Response to Condition 1, Issue 2When the potential leakage understatement adjusted Type B & C MNPLR total is greater thanthe 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 preparedto restore the leakage summation margin to less than the Calvert Cliffs administrative leakagelimit. The corrective action plan shall focus on those components which have contributed themost 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 leakageperformance issues so as to maintain an acceptable level of margin.Response to Condition 1, Issue 3Calvert Cliffs will apply the 9-month grace period only to eligible Type C components and onlyfor non-routine emergent conditions.

Such occurrences will be documented in the record oftests.Topical Report Condition 2The basis for acceptability of extending the ILRT interval out to once per 15 years was theenhanced and robust primary containment inspection program and the local leakage rate testingof penetrations.

Most of the primary containment leakage experienced has been attributed topenetration leakage and penetrations are thought to be the most likely location of mostcontainment leakage at any time. The containment leakage condition monitoring regimeinvolves a portion of the penetrations being tested each refueling outage, nearly all LLRTs beingperformed during plant outages.

For the purposes of assessing and monitoring or trendingoverall containment leakage potential, the as-found minimum pathway leakage rates for the justtested penetrations are summed with the as-left minimum pathway leakage rates for31 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEpenetrations tested during the previous 1 or 2 or even 3 refueling outages.

Type C tests involvevalves, which in the aggregate, will show increasing leakage potential due to normal wear andtear, some predictable and some not so predictable.

Routine and appropriate maintenance mayextend this increasing leakage potential.

Allowing for longer intervals between LLRTs meansthat more leakage rate test results from farther back in time are summed with fewer just testedpenetrations and that total used to assess the current containment leakage potential.

This leadsto the possibility that the LLRT totals calculated understate the actual leakage potential of thepenetrations.

Given the required margin included with the performance criterion and theconsiderable 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, theprimary containment leakage rate testing program trending or monitoring must include anestimate of the amount of understatement in the Type B & C total, and must be included in alicensee's post-outage report. The report must include the reasoning and determination of theacceptability of the extension, demonstrating that the LLRT totals calculated represent theactual leakage potential of the penetrations.

Response to Condition 2Condition 2 presents two (2) separate issues that are required to be addressed.

They are asfollows:* ISSUE 1 -Extending the LLRT intervals beyond 5 years to a 75-month interval should besimilarly conservative provided an estimate is made of the potential understatement and itsacceptability 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 to75-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, andmust 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 totalscalculated represent the actual leakage potential of the penetrations.

Response to Condition 2, Issue 1The change in going from a 60 month extended test interval for Type C tested components to a75 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 leakageunderstatement adjustment factor of 1.25 to the As-Left leakage total for each Type Ccomponent currently on the 75 month extended test interval.

This will result in a combinedconservative Type C total for all 75 month LLRT being "carried forward" and will be includedwhenever the total leakage summation is required to be updated (either while on line orfollowing an outage).

When the potential leakage understatement adjusted leak rate total forthose Type C components being tested on a 75 month extended interval is summed with thenon-adjusted total of those Type C components being tested at less than the 75 month intervaland the total of the Type B tested components, if the Minimum pathway leakage rate is greaterthan the Calvert Cliffs administrative leakage summation limit of 0.50 La, but less than the32 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEregulatory limit of 0.6 La, then an analysis and corrective action plan shall be prepared to restorethe leakage summation value to less than the Calvert Cliffs administrative leakage limit. Thecorrective action plan shall focus on those components which have contributed the most to theincrease in the leakage summation value and what manner of timely corrective action, asdeemed appropriate, best focuses on the prevention of future component leakage performance issues.Response to Condition 2, Issue 2If the potential leakage understatement adjusted leak rate Minimum pathway leakage rate isless than the Calvert Cliffs administrative leakage summation limit of 0.50 La, then theacceptability of the 75 month LLRT extension for all affected Type C components has beenadequately demonstrated and that the calculated local leak rate total represents the actualleakage 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 inSection 12.1, Report Requirements:

A post-outage report shall be prepared presenting results of the previous cycle's Type B andType C tests, and Type A, Type B and Type C tests, if performed during that outage. Thetechnical contents of the report are generally described in ANSI/ANS-56.8-2002 and shall beavailable on-site for NRC review. The report shall show that the applicable performance criteriaare met, and serve as a record that continuing performance is acceptable.

The report shall alsoinclude the combined Type B and Type C leakage summation, and the margin between theType B and Type C leakage rate summation and its regulatory limit. Adverse trends in theType 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 leakageunderstatement adjusted Type B & C summation is identified, then an analysis anddetermination 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 whatmanner of timely corrective action, as deemed appropriate, best focuses on the prevention offuture 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 theestimate 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 forFloor Welds of Metal Containment Shell and Concrete Containment Metallic LinerThe NRC issued Information Notice (IN) 2014-07 (Reference

21) to inform addressees of issuesidentified by the NRC staff concerning degradation of floor weld leak-chase channel systems ofsteel containment shell and concrete containment metallic liner that could affect leak-tightness and aging management of containment structures.

The NRC expects that recipients will reviewthe information for applicability to their facilities and consider

actions, as appropriate, to avoidsimilar problems.

33 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEAt Calvert Cliffs, a test pipe was provided for each continuous segment of the bottom liner plateweld chase test channels (equivalent to containment weld channels).

The tops of the pipes arelocated above the cover slab and are sealed with caps. These pipes were initially used to testthe leak tightness of the bottom liner. During the performance of an ILRT, the caps for the linerplate 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 notapplicable to Calvert Cliffs. A more thorough analysis of this issue is in progress to determine ifthis 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 theperformance-based requirements of 10 CFR Part 50, Appendix J, Option B. It incorporated theregulatory 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 aperformance-based approach for determining Type A, Type B, and Type C containment leakagerate 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 thatextension of the containment ILRT interval from 10 to 15 years represents minimal risk toincreased leakage.

The risk is minimized by continued Type B and Type C testing performed inaccordance 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 ProgramThis experience is supplemented by risk analysis

studies, including the Calvert Cliffs 1 and 2risk analysis provided in Attachment

3. The findings of the risk assessment confirm the generalfindings of previous

studies, on a plant-specific basis, that extending the ILRT interval from tento 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 andrequirements continue to be met.10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to besubject to the requirements of Appendix J to 10 CFR Part 50, "Leakage Rate Testing ofContainment of Water Cooled Nuclear Power Plants."

Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity ofthe primary reactor containment and systems and components which penetrate thecontainment.

In addition, Appendix J discusses leakage rate acceptance

criteria, testmethodology, frequency of testing and reporting requirements for each type of test.34 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEThe 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 ratetesting is performed;

however, it did alter the frequency at which Type A, Type B, and Type Ccontainment leakage tests must be performed.

Under the performance-based option of 10 CFRPart 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 thatleakage limits will be maintained.

The change to the Type A test frequency did not directlyresult in an increase in containment leakage.

Similarly, the proposed change to the Type C testfrequency 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 areallowed by this guideline.

The Risk Impact Assessment of Extended Integrated Leak RateTesting Intervals, EPRI report 1018243 (Formerly TR-1009325, Revision

2) indicates that, ingeneral, the risk impact associated with ILRT interval extensions for intervals up to 15 years issmall. 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 CFRPart 50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals toup to 15 years and incorporates the regulatory positions stated in RG 1.163. The NRC stafffinds that the Type A testing methodology as described in ANSI/ANS-56.8-2002, and themodified testing frequencies recommended by NEI TR 94-01, Revision 2, serves to ensurecontinued leakage integrity of the containment structure.

Type B and Type C testing ensuresthat individual penetrations are essentially leak tight. In addition, aggregate Type Band Type Cleakage 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 leakagetightness of primary containment by minimizing potential leakage paths.For EPRI Report No. 1009325, Revision 2, a risk-informed methodology using plant-specific riskinsights and industry ILRT performance data to revise ILRT surveillance frequencies, the NRCstaff finds that the proposed methodology satisfies the key principles of risk-informed decisionmaking 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 toamend their TS in regards to containment leakage rate testing, subject to the limitations andconditions noted in Section 4.2 of the SER.The NRC staff reviewed NEI TR 94-01, Revision 3, and determined that it described anacceptable approach for implementing the optional performance-based requirements of OptionB to 10 CFR Part 50, Appendix J, as modified by the conditions and limitations summarized inSection 4.0 of the associated SE. This guidance included provisions for extending Type CLLRT 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 theleakage tightness of primary containment by minimizing potential leakage paths. The NRCstaff, 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 tocontainment leakage rate testing.

Any applicant may reference NEI TR 94-01, Revision 3, as35 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEmodified by the associated SER and approved by the NRC, in a licensing action to satisfy therequirements of Option B to 10 CFR Part 50, Appendix J.Based on the considerations above, (1) there is reasonable assurance that the health andsafety of the public will not be endangered by operation in the proposed manner, (2) suchactivities will continue to be conducted in accordance with the site licensing basis, and (3) theapproval of the proposed change will not be inimical to the common defense and security or tothe 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 TestFrequency 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."

Theproposed change to the TS would revise Calvert Cliffs TS 5.5.16, by replacing the reference toRegulatory Guide (RG) 1.163 with a reference to Nuclear Energy Institute (NEI) topical reportNEI 94-01 Revision 3-A (NRC-approved version specified in the 10 CFR Part 50, Appendix JProgram Plan) as the implementation document used by Calvert Cliffs to implement the Units 1and 2 performance-based leakage testing program in accordance with Option B of 10 CFRPart 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. Theproposed change also deletes exceptions previously granted to allow one time extensions of theILRT 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 withthe 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 probability or consequences of an accident previously evaluated?

Response:

No.The proposed amendment to the TS involves the extension of the Calvert Cliffs Unit 1 and 2Type A containment test interval to 15 years and the extension of the Type C test interval to75 months. The current Type A test interval of 120 months (10 years) would be extendedon a permanent basis to no longer than 15 years from the last Type A test. The currentType C test interval of 60 months for selected components would be extended on aperformance basis to no longer than 75 months. Extensions of up to nine months (totalmaximum 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 to36 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGEthe plant or a change in the manner in which the plant is operated or controlled.

Thecontainment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents.

As such, thecontainment 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 anaccident, and do not involve the prevention or identification of any precursors of anaccident.

Therefore, this proposed extension does not involve a significant increase in theprobability of an accident previously evaluated.

As documented in NUREG-1493, Type B and C tests have identified a very largepercentage of containment leakage paths, and the percentage of containment leakagepaths that are detected only by Type A testing is very small. The Calvert Cliffs Unit 1 and 2Type A test history supports this conclusion.

The integrity of the containment is subject to two types of failure mechanisms that can becategorized as (1) activity based and (2) time based. Activity based failure mechanisms aredefined 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 andconstruction requirements of the containment combined with the containment inspections performed in accordance with ASME Section Xl, the Maintenance Rule, and TSrequirements serve to provide a high degree of assurance that the containment would notdegrade in a manner that is detectable only by a Type A test. Based on the above, theproposed extension does not significantly increase the consequences of an accidentpreviously evaluated.

The proposed amendment also deletes exceptions previously granted to allow one timeextensions of the ILRT test frequency for both Units 1 and 2 and exceptions fromconducting post modification ILRT following replacement of the Units 1 and 2 SteamGenerators.

These exceptions were for things that have already taken place so theirdeletion is solely an administrative action that has no effect on any component and noimpact on how the units are operated.

Therefore, the proposed change does not result in a significant increase in the probability orconsequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident fromany accident previously evaluated?

Response:

No.The proposed amendment to the TS involves the extension of the Calvert Cliffs Unit 1 and 2Type A containment test interval to 15 years and the extension of the Type C test interval to75 months. The containment and the testing requirements to periodically demonstrate theintegrity 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 changedoes 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 CHANGEThe proposed amendment also deletes exceptions previously granted to allow one timeextensions 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 anadministrative 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 ofaccident 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 1and 2 Type A containment test interval to 15 years and the extension of the Type C testinterval to 75 months for selected components.

This amendment does not alter the mannerin which safety limits, limiting safety system set points, or limiting conditions for operation are determined.

The specific requirements and conditions of the TS Containment LeakRate Testing Program exist to ensure that the degree of containment structural integrity andleak-tightness that is considered in the plant safety analysis is maintained.

The overallcontainment leak rate limit specified by TS is maintained.

The proposed change involves only the extension of the interval between Type Acontainment leak rate tests and Type C tests for Calvert Cliffs Unit 1 and 2. The proposedsurveillance interval extension is bounded by the 15 year ILRT Interval and the 75 monthType C test interval currently authorized within NEI 94-01, Revision 3-A. Industryexperience supports the conclusion that Type B and C testing detects a large percentage ofcontainment leakage paths and that the percentage of containment leakage paths that aredetected only by Type A testing is small. The containment inspections performed inaccordance with ASME Section XI and TS serve to provide a high degree of assurance thatthe 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 safetyanalysis is maintained.

The design, operation, testing methods and acceptance criteria forType 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 notaffected by changes to the Type A and Type C test intervals.

The proposed amendment also deletes exceptions previously granted to allow one timeextensions of the ILRT test frequency for both Units 1 and 2 and exceptions fromconducting post modification ILRT following replacement of the Units 1 and 2 SteamGenerators.

These exceptions were for things that have already taken place so theirdeletion is an administrative action and does not change how the units are operated andmaintained, thus there is no reduction in any margin of safety.Therefore, the proposed change does not involve a significant reduction in a margin ofsafety.Based on the above, Calvert Cliffs concludes that the proposed amendment presents nosignificant 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 CHANGE4.4 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 proposedmanner, (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 securityor to the health and safety of the public.5.0 ENVIRONMENTAL CONSIDERATION A review has determined that the proposed amendment would change an inspection orsurveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in theamounts of any effluent that may be released

offsite, or (iii) a significant increase in individual orcumulative occupational radiation exposure.

Accordingly, the proposed amendment meets theeligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Therefore, pursuantto 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need beprepared in connection with the proposed amendment.

6.0 REFERENCES

1. Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program,September 19952. NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 20123. Regulatory Guide 1.174, Revision 2, An Approach For Using Probabilistic RiskAssessment In Risk-Informed Decisions On Plant-Specific Changes To The Licensing Basis, May 20114. Regulatory Guide 1.200, Revision 2, An Approach For Determining The Technical Adequacy Of Probabilistic Risk Assessment Results For Risk-Informed Activities, March20095. NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Optionof 10 CFR Part 50, Appendix J, July 19956. NUREG-1493, Performance-Based Containment Leak-Test

Program, January 19957. EPRI TR-104285, Risk Impact Assessment of Revised Containment Leak Rate TestingIntervals, August 19948. NEI 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, October 20089. Letter from M. J. Maxin (NRC) to J. C. Butler (NEI), dated June 25, 2008, Final SafetyEvaluation 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 RateTesting Intervals" (TAC No. MC9663)39 ATTACHMENT (1)EVALUATION OF THE PROPOSED CHANGE10. Letter from S. Bahadur (NRC) to B. Bradley (NEI), dated June 8, 2012, Final SafetyEvaluation 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 CliffsNuclear 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 CliffsNuclear 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, CalvertCliffs Nuclear Power Plant, Unit Nos. 1 and 2 -Amendment Re: Implementation ofAlternative 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-YearExtension 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, CalvertCliffs 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 LeakRate Testing Intervals, October 200819. Letter from G. H. Gellrich (CCNPP) to Document Control Desk (NRC), datedSeptember 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 201021. NRC Information Notice 2014-07, Degradation of Leak-Chase Channel Systems forFloor Welds of Metal Containment Shell and Concrete Containment Metallic Liner,May 5, 201422. Letter from R. V. Guzman (NRC) to S. L. Belcher (NMP), dated March 30, 2010, NineMile Point Nuclear Station, Unit No. 2 -Issuance of Amendment Re: Extension ofPrimary 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 CHANGE24. 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 PAGECalvert Cliffs Nuclear Power PlantSeptember 18, 2014 Programs and Manuals5.55.5 Progirams and Manuals5.5.16Containment Leakaqe Rate Testinq ProqramA program shall be established to implement the leakagetesting 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....

Based Cntainmnt Leak Test Pr.gram,"

dated September 1995,INSERT 1ineluding errata, as moiedifie Ay tche feliewilg exeeptions:

a. Nuclear Energy institcute (NEI) 94 01 1995, S..tion9.2.3:, The first Unit 1 Type A test perfermced a:ftthe Junc 15, 1992 Type A test shall be nolatter than Juno 14, 2007. The flirst Unitt 2 Type -Atest performold a~fter the May 2, 2001 Type A tostshall be performold ne later than May 1, 2016.b. Unit 1 is emecpted frefm pest medifioatien integrated leakage rate test infg reguirefmonts associated witsteamf gemerater replaeecmont-.

e. Unit 2 is emeepted fromA pest m~edifieatien integrated I"PA ^A .P -iA ... P h *PQ n"-R TPA ^ 4Q

• Q,~.A.*,-

4 T .A... ^* m *._1J_~t~a y~nrat rerpiaccecnt The peak calculated containment internal pressure for thedesign 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 be0.16 percent of containment air weight per day at Pa.Leakage rate acceptance criteria are:a. Containment leakage rate acceptance criterion is -5 1.0La. During the first unit startup following testing,in accordance with this program, the leakage rateacceptance criterion are -5 0.60 La for Types B and Ctests 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," datedJuly 2012CALVERT CLIFFS -UNIT 1CALVERT CLIFFS -UNIT 25.5-17Amendment No. 3G3Amendment No. 284 ATTACHMENT (3)EVALUATION OF RISK SIGNIFICANCE OF PERMANENT ILRTEXTENSION Calvert Cliffs Nuclear Power PlantSeptember 18, 2014

.IHUGHES0ASSOCIATES 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 PlantProject Number: 0054-0001-000 Project Title: Permanent ILRT Extension Revision:

3Name and DateDtg]Wty signed by Maln JohnsonPreparer Matthew Johnson Matt Johns am the author of thiso 2014 09.12 09:36:14-05'00'

.D.. y signed by NicholasReviewer:

Nicholas Lovelace Lol K._ t "' L aceDate 2014.09.12 14:06:09-05'00' Review Method Design Review Calculation DApproved by: Richard Anoba V C '7 al /I t/Revision 3 Page 1 ot93Revision 3Page I of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension REVISION RECORD SUMMARYRevision Revision Summary0 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 insection 2.0.Revision 3 Page 2of93Revision 3Page 2 of 93 IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension TABLE OF CONTENTS1 .0 P U R P O S E ......................................................................................................................

42 .0 S C O P E ...........................................................................................................................

43.0 REFERENCES

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64.0 ASSUMPTIONS AND LIMITATIONS

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85.0 METHODOLOGY and analysis

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85 .1 In p u ts ........................

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85.1.1 General Resources Available

.............................................................................

85.1.2 Plant Specific Inputs .......................................................................................

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

145.1.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage..........................................................................................................................

1 65 .2 A n a ly s is ......................................................................................................................

1 85.2.1 Step 1 -Quantify the Baseline Risk in Terms of Frequency per Reactor Year ..... 195.2.2 Step 2 -Develop Plant-Specific Person-Rem Dose (Population Dose) ...........

235.2.3 Step 3 -Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15Y e a rs .................................................................................................................

2 45.2.4 Step 4 -Determine the Change in Risk in Terms of LERF ...............................

275.2.5 Step 5 -Determine the Impact on the Conditional Containment Failure Probability (C C F P ) ........................................................................................................

..3 05 .3 S e n s itiv itie s .................................................................................................................

3 15.3.1 Potential Impact from External Events Contribution

........................................

315.3.1.1 Potential Impact from External Events Contribution Using IPEEE FireAnalysis

...............................................................................................

335.3.2 Potential Impact from Steel Liner Corrosion Likelihood

....................................

355.3.3 Expert Elicitation Sensitivity

............................................................................

365.3.4 Large Leak Probability Sensitivity Study .......................................................

386 .0 R E S U LT S ......................................................................................................................

4

07.0 CONCLUSION

S AND RECOMMENDATIONS

........................................................

42A .A tta ch m e nt 1 .................................................................................................................

4 3Revision 3 Page 3 of 93Revision 3Page 3 of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 1.0 PURPOSEThe purpose of this analysis is to provide a risk assessment of extending the currently allowedcontainment Type A Integrated Leak Rate Test (ILRT) to permanent fifteen years. Theextension 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 riskassessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1], themethodology 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 NRCregulatory 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 fora plant's licensing basis as outlined in Regulatory Guide (RG) 1.174 [Reference 4], themethodology 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 SCOPERevisions 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 tenyears 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 apartin which the calculated performance leakage rate was less than limiting containment leakagerate of 1La.The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision0, and established in 1995 during development of the performance-based Option B to AppendixJ. Section 11.0 of NEI 94-01 states that NUREG-1493, "Performance-Based Containment LeakTest Program,"

September 1995 [Reference 6], provides the technical basis to supportrulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. Thebasis consisted of qualitative and quantitative assessment of the risk impact (in terms ofincreased public dose) associated with a range of extended leakage rate test intervals.

Tosupplement the NRC's rulemaking basis, NEI undertook a similar study. The results of thatstudy 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 ofcontainment leakage on the health and safety of the public and the benefits realized from thecontainment leak rate testing.

In that analysis, it was determined that for a representative PWRplant (i.e., Surry), that containment isolation failures contribute less than 0.1 percent to the latentrisks from reactor accidents.

Consequently, it is desirable to show that extending the ILRTinterval will not lead to a substantial increase in risk from containment isolation failures forCCNPP.NEI 94-01 Revision 2-A contains a Safety Evaluation Report that supports using EPRI ReportNo. 1009325 Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate TestingIntervals, 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 theEPRI Risk Assessment methodology, EPRI TR-104285.

This methodology is followed todetermine the appropriate risk information for use in evaluating the impact of the proposed ILRTchanges.Revision 3Page 4 of 93 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 ofMechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MCpressure-retaining components and their integral attachments, and of metallic shell andpenetration liners of Class CC pressure-retaining components and their integral attachments inlight-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) requirelicensees to conduct visual inspections of the accessible areas of the interior of thecontainment.

The associated change to NEI 94-01 will require that visual examinations beconducted during at least three other outages, and in the outage during which the ILRT is beingconducted.

These requirements will not be changed as a result of the extended ILRT interval.

Inaddition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity ofcontainment penetration

bellows, airlocks, seals, and gaskets are also not affected by thechange 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 ofAppendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines asincreases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases inLarge Early Release Frequency (LERF) less than 10-7 per reactor year. Since the Type A testdoes not impact CDF, the relevant criterion is the change in LERF. RG 1.174 also defines smallchanges in LERF as below 10-6 per reactor year. RG 1.174 discusses defense-in-depth andencourages 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 ismaintained, is also calculated.

Regarding CCFP, changes of up to 1.1% have been accepted by the NRC for the one-timerequests for extension of ILRT intervals.

In context, it is noted that a CCFP of 1/10 (10%) hasbeen approved for application to evolutionary light water designs.

Given these perspectives, achange 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 todemonstrate the relative change in this parameter.

While no acceptance guidelines for theseadditional 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 ofincremental increases in population dose that have been accepted by the NRC. The range ofincremental population dose Increases is from <0.01 to 0.2 person-rem/year and/or 0.002% to0.46% 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 isdefined 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 impactassessment of the proposed extended ILRT interval.

For those plants that credit containment overpressure for the mitigation of design basisaccidents, a brief description of whether overpressure is required should be included in thissection.

In addition, if overpressure is included in the assessment, other risk metrics such asCDF should be described and reported.

Revision 3 Page 5 of 93Revision 3Page 5 of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

3.0 REFERENCES

The following references were used in this calculation:

1. Revision 3-A to Industry Guideline for Implementing Performance-Based Option of 10CFR Part 50, Appendix J, NEI 94-01, July 2012.2. Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, EPRI,Palo Alto, CA EPRI TR-104285, August 1994.3. Interim Guidance for Performing Risk Impact Assessments in Support of One-TimeExtensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Revision 4, developed for NEI by EPRI and Data Systems and Solutions, November2001.4. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions onPlant-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 UnitNo. 3 -Issuance of Amendment Re: Frequency of Performance-Based Leakage RateTesting (TAC No. MB0178),

April 17, 2001.10. Impact of Containment Building Leakage on LWR Accident Risk, Oak Ridge NationalLaboratory, 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 TestingIntervals Utilizing ORAMTM, 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, "PRAQuantification (QU) Notebook,"

July 2012.Revision 3 Page 6 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

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 forLevel 3 PRA Release Categories,"

August 2001.20. Anthony R. Pietrangelo, One-time extensions of containment integrated leak rate testinterval

-additional information, NEI letter to Administrative Points of Contact,November 30, 2001.21. Letter from J. A. Hutton (Exelon, Peach Bottom) to U. S. Nuclear Regulatory Commission, Docket No. 50-278, License No. DPR-56, LAR-01-00430, dated May 30,2001.22. Risk Assessment for Joseph M. Farley Nuclear Plant Regarding ILRT (Type A)Extension

Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, P0293010002-1929-030602, March 2002.23. Letter from D. E. Young (Florida Power, Crystal River) to U. S. Nuclear Regulatory Commission, 3F0401-11, dated April 25, 2001.24. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision2-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, "PRAQuantification (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,"

May2010.33. Landale, J., PRAER No. CO-2010-012, CENG, CO-2010-012, August 2010.34. Harrison, D., Generic Component Fragilities for the GE Advanced BWR SeismicAnalysis, 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.Revision 3 Page 7 of 93Revision 3Page 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 ofRegulatory Guide 1.200 as is relevant to this ILRT interval extension, as detailed inAttachment 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 forthe ILRT extension.

The IPEEE simplified seismic PRA [Reference 35] and the detailedFire PRA (model 6.1 M) are used for this sensitivity analysis.

It is reasonable to assumethat 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 beincluded in the calculations.

" Accident classes describing radionuclide release end states are defined consistent withEPRI methodology

[Reference 2]." The representative containment leakage for Class 1 sequences is 1 La. Class 3 accountsfor increased leakage due to Type A inspection failures.

" The representative containment leakage for Class 3a sequences is 10La based on thepreviously approved methodology performed for Indian Point Unit 3 [Reference 8,Reference 9]." The representative containment leakage for Class 3b sequences is 100La based on theguidance 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 theproposed ILRT extension, but is accounted for in the EPRI methodology as a separateentry for comparison purposes.

Since the containment bypass contribution to population dose is fixed, no changes on the conclusions from this analysis will result from thisseparate 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 ANALYSIS5.1 InputsThis section summarizes the general resources available as input (Section 5.1.1) and the plantspecific 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]Revision 3Page 8 of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension

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 usedin the Level 2 PRA for the size of containment leakage that is considered significant and is to beincluded in the model. The second study is applicable because it provides a basis of theprobability for significant pre-existing containment leakage at the time of a core damageaccident.

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 anassessment of the impact of different containment leakage rates on plant risk. The fifth studyprovides an assessment of the impact on shutdown risk from ILRT test interval extension.

Thesixth 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 ofextending ILRT and LLRT test intervals on at-power public risk. The eighth study provides anex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the basisfor the consequence analysis of the ILRT interval extension for CCNPP. The ninth studyincludes the NEI recommended methodology (promulgated in two letters) for evaluating the riskassociated 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, theeleventh study builds on the previous work and includes a recommended methodology andtemplate for evaluating the risk associated with a permanent 15-year extension of the ILRTinterval.

NUREG/CR-3539

[Reference 101Oak Ridge National Laboratory documented a study of the impact of containment leak rates onpublic 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 rateson LWR accident risks is relatively small.NUREG/CR-4220

[Reference 111NUREG/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 tocalculate the unavailability of containment due to leakage.NUREG-1273

[Reference 121A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of theNUREG/CR-4220 database.

This assessment noted that about one-third of the reported eventswere leakages that were immediately detected and corrected.

In addition, this study noted thatlocal leak rate tests can detect "essentially all potential degradations" of the containment isolation system.NUREG/CR-4330

[Reference 131NUREG/CR-4330 is a study that examined the risk impacts associated with increasing theallowable containment leakage rates. The details of this report have no direct impact on theRevision 3Page 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 leakagerate 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 byaccident sequences that result in failure or bypass of containment."

EPRI TR-105189

[Reference 141The 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. Thisstudy 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 onshutdown risk. The conclusion from the study is that a small, but measurable, safety benefit isrealized from extending the test intervals.

NUREG-1493

[Reference 61NUREG-1493 is the NRC's cost-benefit analysis for proposed alternatives to reducecontainment leakage testing intervals and/or relax allowable leakage rates. The NRCconclusions 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 pathsdetected solely by Type A testing, increasing the interval between integrated leak rate tests ispossible with minimal impact on public risk.EPRI TR-104285

[Reference 21Extending 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 andLLRT test intervals on at-power public risk. This study combined IPE Level 2 models withNUREG-1 150 Level 3 population dose models to perform the analysis.

The study also used theapproach of NUREG-1493 in calculating the increase in pre-existing leakage probability due toextending the ILRT and LLRT test intervals.

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

1. Containment intact and isolated2. Containment isolation failures dependent upon the core damage accident3. Type A (ILRT) related containment isolation failures4. Type B (LLRT) related containment isolation failures5. Type C (LLRT) related containment isolation failures6. Other penetration related containment isolation failures7. Containment failures due to core damage accident phenomena

8. Containment bypassConsistent with the other containment leakage risk assessment

studies, this study concluded:

"...the proposed CLRT (Containment Leak Rate Tests) frequency changes would have aminimal safety impact. The change in risk determined by the analyses is small in both absoluteand relative terms. For example, for the PWR analyzed, the change is about 0.04 person-rem per year...Revision 3Page 10 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension NUREG-1 150 [Reference 151 and NUREG/CR-4551

[Reference 71NUREG-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 theCCNPP Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it isconsidered adequate to represent CCNPP. (The meteorology and site differences other thanpopulation are assumed not to play a significant role in this evaluation.)

NEI Interim Guidance for Performinq Risk Impact Assessments In Support of One-TimeExtensions for Containment Integrated Leakage Rate Test Surveillance Intervals

[Reference 3,Reference 201The 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 LicenseAmendment for a One-Time Integrated Leakage Rate Test Extension

[Reference 51This submittal to the NRC describes a method for determining the change in likelihood, due toextending the ILRT, of detecting liner corrosion, and the corresponding change in risk. Themethodology 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 factoredinto the risk assessment for the ILRT one-time extension.

The Calvert Cliffs analysis wasperformed 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 LeakRate Testing Intervals

[Reference 241This report provides a generally applicable assessment of the risk involved in extension of ILRTtest intervals to permanent 15-year intervals.

Appendix H of this document provides guidancefor performing plant-specific supplemental risk impact assessments and builds on the previousEPRI risk impact assessment methodology

[Reference 2] and the NRC performance-based containment leakage test program [Reference 6], and considers approaches utilized in varioussubmittals, including Indian Point 3 (and associated NRC SER) and Crystal River.The approach included in this guidance document is used in the CCNPP assessment todetermine the estimated increase in risk associated with the ILRT extension.

This documentincludes the bases for the values assigned in determining the probability of leakage for the EPRIClass 3a and 3b scenarios in this analysis, as described in Section 5.2.5.1.2 Plant Specific InputsThe 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]Revision 3Page 11 of 93 1 RCA-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 ModelThe Level 1 Internal Events PRA Model that is used for CCNPP is characteristic of the as-builtplant. The current Level 1 model (CCNPP PRA Model Version 6.2a) [Reference 17] is a linkedfault 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 EventCDF (excluding seismic)

= 3.24E-5/year for Unit 1 and 3.71 E-5/year for Unit 2. Table 5-1provides 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 includedin CCNPP PRA Model Version 6.2a. The Fire PRA results come from Model Version 6.1 M. TheSeismic PRA results come from the IPEEE Seismic Analysis

[Reference 35].Table 5-1 -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-6Internal Floods 6.18E-6 1.06E-6Transients 3.40E-6 4.70E-6ISLOCA 1.97E-7 1.97E-7SGTR 4.71E-7 4.60E-7Total Internal Events CDF 1.61 E-5 1.41 E-5Total Internal Events CDF 1.34E-5(Excluding ISLOCA & SGTR)Table 5-2 -External Events CDFExternal Events Unit I Frequency (per year) Unit 2 Frequency (per year)Fire 3.15E-5 3.59E-5High Winds 9.19E-7 1.23E-6Seismic 1.07E-5 1.07E-5Total External Events CDF 4.31E-5 4.78E-5Note that the above Fire PRA values reflect the anticipated configuration of the plant upon fullimplementation of NFPA 805 and related plant modifications to resolve fire protection issues.Refer to Section 5.3.1.Level 2 ModelThe 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 themodel: INTACT, SERF (small early release frequency),

and LATE [Reference 31]. The currentLERF model (CCNPP PRA Model Version 6.2a) [Reference 17] is a linked fault tree model andwas quantified with the total Unit 1 Internal Events LERF = 1.39E-6/year and Unit 2 InternalRevision 3Page 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 5-3provides 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 includedin CCNPP PRA Model Version 6.2a. The Fire PRA results come from Model Version 6.1M. TheSeismic PRA results come from the IPEEE Seismic Analysis

[Reference 35].Table 5-3 -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-7Internal Floods 2.46E-7 2.17E-7Transients 1.50E-7 2.84E-7ISLOCA 1.97E-7 1.97E-7SGTR 4.71E-7 4.60E-7Total Internal Events LERF 1.39E-6 1.56E-6Table 5-4 -External Events LERFExternal Events Unit I Frequency (per year) Unit 2 Frequency (per year)Fire 2.97E-6 4.17E-6High Winds 2.21E-8 3.77E-8Seismic 1.41E-6 1.41E-6Total External Events CDF 4.40E-6 5.62E-6Note that the above Fire PRA values reflect the anticipated configuration of the plant upon fullimplementation 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 3bpopulation dose values are calculated from the Class 1 population dose and represented as1OLa and 1OOL a, respectively, as guidance in Reference 1 dictates.

Table 5-5 -Population DoseAccident Class Description Release (person-rem) 1 Containment Remains Intact 3.20E+042 Containment Isolation Failures 2.OOE+073a Independent or Random Isolation Failures SMALL 3.20E+0513b Independent or Random Isolation Failures LARGE 3.20E+062Revision 3Page 13 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of permanent ILRT Extension Table 5-5 -Population DoseAccident Class Description Release (person-rem)

Isolation Failure in which pre-existing leakage is not n/adependent on sequence progression.

Type B test Failures5 Isolation Failure in which pre-existing leakage is not n/adependent on sequence progression.

Type C test Failures6 Isolation Failure that can be verified by IST/IS or 7.01 E+06surveillance 7 Containment Failure induced by severe accident 5.61 E+078 Accidents in which containment is by-passed 2.25E+071. 10*La2. 100"* LRelease Cateaorv Definitions Table 5-6 defines the accident classes used in the ILRT extension evaluation, which isconsistent 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 testinterval, as described in Section 5.2 of this report.Table 5-6 -EPRI Containment Failure Classification

[Reference 2]Class Description Containment remains intact including accident sequences that do not lead to containment failure in the1 long term. The release of fission products (and attendant consequences) is determined by the maximumallowable 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 theType 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 Jtesting 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 certainbellows arrangements and failure of some sealing surfaces, which can lead to leakage.

Theproposed 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 3Revision 3Page 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 theEPRI Guidance

[Reference 24]. For Class 3a, the probability is based on the maximumlikelihood estimate of failure (arithmetic average) from the available data (i.e., 2 "small" failuresin 217 tests leads to "large" failures in 217 tests (i.e., 2/217 = 0.0092).

For Class 3b, theprobability 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 issuedadditional information concerning the potential that the calculated delta LERF values for severalplants 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 thequantitative 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 forsimplicity and to maintain conservatism.

However, some plant-specific accident classes leadingto core damage are likely to include individual sequences that either may already(independently) cause a LERF or could never cause a LERF, and are thus not associated with apostulated large Type A containment leakage path (LERF). These contributors can be removedfrom Class 3b in the evaluation of LERF by multiplying the Class 3b probability by only thatportion 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 toClass 3b. To be consistent, the same change is made to the Class 3a CDF, even thoughthese events are not considered LERF. Class 2 events refer to sequences with largepre-existing containment isolation failures; Class 8 events refer to sequences withcontainment bypass events. These sequences are already considered to contribute toLERF in the CCNPP Level 2 PRA analysis.

" A review of Class 1 accident sequences shows that several of these cases involvesuccessful operation of containment sprays. For calculation of the Class 3b and Class3a frequencies, the fraction of the Class 1 CDF associated with successful operation ofcontainment 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 inSection 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.

Forexample, the average time that a leak could go undetected with a three-year test interval is 1.5years (3 years / 2), and the average time that a leak could exist without detection for a ten-yearinterval is 5 years (10 years / 2). This change would lead to a non-detection probability that is afactor 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 afactor of 5 ((15/2)/1.5) increase in the non-detection probability of a leak.Revision 3 Page 15 of 93Revision 3Page 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 comparedto previous submittals (e.g., the IP3 request for a one-time ILRT extension that was approved bythe NRC [Reference 9]) because it does not factor in the possibility that the failures could bedetected by other tests (e.g., the Type B local leak rate tests that will still occur). Eliminating thispossibility 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 LeakageAn estimate of the likelihood and risk implications of corrosion-induced leakage of the steelliners occurring and going undetected during the extended test interval is evaluated using themethodology from the Calvert Cliffs liner corrosion analysis

[Reference 5]. The Calvert Cliffsanalysis was performed for a concrete cylinder and dome and a concrete base-mat, each with asteel liner.The following approach is used to determine the change in likelihood, due to extending theILRT, of detecting corrosion of the containment steel liner. This likelihood is then used todetermine 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 flawAssumptions

" Consistent with the Calvert Cliffs analysis, a half failure is assumed for base-matconcealed 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 Cliffsprevious analysis are assumed to still be applicable.

" Consistent with the Calvert Cliffs analysis, the estimated historical flaw probability is alsolimited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55astarted requiring visual inspection.

Additional success data was not used to limit theaging impact of this corrosion issue, even though inspections were being performed priorto this date (and have been performed since the time frame of the Calvert Cliffsanalysis),

and there is no evidence that additional corrosion issues were identified (SeeTable 5-7, Step 1)." Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumed todouble every five years. This is based solely on judgment and is included in this analysisto 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 tenyears and every two years." In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching theoutside atmosphere, given that a liner flaw exists, was estimated as 1.1% for the cylinderand dome, and 0.11% (10% of the cylinder failure probability) for the base-mat.

Thesevalues 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 thecylinder 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 crackRevision 3Page 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 failurelikelihood 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 (SeeTable 5-7, Step 5).Consistent with the Calvert Cliffs analysis, all non-detectable containment failures are assumedto result in early releases.

This approach avoids a detailed analysis of containment failure timingand operator recovery actions.Table 5-7 -Steel Liner Corrosion Base CaseStep Description Containment Cylinder and Containment BasematDome (85%) (15%)Historical liner flaw likelihood Events: 2 Events: 0Failure data: containment location (Brunswick 2 and North Anna 2) Assume a half failurespecific 2 / (70 x 5.5) = 5.19E-03 0.5 / (70 x 5.5) = 1.30E-03Success data: based on 70 steel-lined containments and 5.5 yearssince the 1OCFR 50.55arequirements of periodic visualinspections of containment surfacesYear Failure rate Year Failure rateAged adjusted liner flaw likelihood During the 15-year interval, assume 1 2.05E-03 1 5.13E-04failure 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-03average for the 5th to 10th year setto the historical failure rate. 15 year average = 6.44E-03 15 year average = 1.61 E-03Increase in flaw likelihood between3 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 flaw10%5% failure to identify visual flawsplus 5% likelihood that the flaw isVisual inspection detection failure not visible (not through-cylinder 100%likelihood but could be detected by ILRT) All Cannot be visually inspected events have been detectedthrough visual inspection.

5%visible failure detection is aconservative 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 3Page 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 ofStep 6 for the containment cylinder and dome, and the containment base-mat, as summarized below for CCNPP.Table 5-8 -Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for CCNPPDescription 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 alreadyindependently LERF or that could never result in LERF.5.2 AnalysisThe 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 riskassessment submittals on this subject [References 5, 8, 21, 22, and 23] have led to thefollowing results.

The results are displayed according to the eight accident classes defined inthe EPRI report, as described in Table 5-6.The analysis performed examined CCNPP-specific accident sequences in which thecontainment remains intact or the containment is impaired.

Specifically, the breakdown of thesevere accidents, contributing to risk, was considered in the following manner:" Core damage sequences in which the containment remains intact initially and in the longterm (EPRI TR-104285, Class 1 sequences

[Reference 2])." Core damage sequences in which containment integrity is impaired due to randomisolation failures of plant components other than those associated with Type B or Type Ctest components.

For example, liner breach or bellow leakage (EPRI TR-104285, Class3 sequences

[Reference 2])." Accident sequences involving containment bypassed (EPRI TR-1 04285, Class 8sequences

[Reference 2]), large containment isolation failures (EPRI TR-104285, Class2 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 thisevaluation as part of the baseline risk profile.

However, they are not affected by the ILRTfrequency 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 5-9 -EPRI Accident Class Definitions Accident Classes Description (Containment Release Type)1 No Containment Failure2 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 3Page 18 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-9 -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 theaccident classes presented in Table 5-9.Step 2 -Develop plant-specific person-rem dose (population dose) per reactor year for each ofthe eight accident classes.Step 3 -Evaluate risk impact of extending Type A test interval from 3 in 10 years to 1 in 15years 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) inaccordance 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 YearAs previously described, the extension of the Type A interval does not influence those accidentprogressions that involve large containment isolation

failures, Type B or Type C testing, orcontainment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks isincluded 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 theprobability of a liner breach or bellows failure (due to excessive leakage) at the time of coredamage. 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 forCCNPP by first determining the frequencies for Classes 1, 2, 6, 7, and 8. Table 5-10 provides acorrelation of the adjusted release category frequencies and the EPRI release classes in Table5-9. 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 categoryfor each sequence and the totals of each EPRI classification.

Table 5-13 provides a summary ofthe accident sequence frequencies that can lead to radionuclide release to the public and havebeen 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 ReportNo. 1009325, Revision 2-A [Reference 24]. Adjustments were made to the Class 3b and henceClass 1 frequencies to account for the impact of undetected corrosion of the steel liner per themethodology described in Section 5.1.4. Note: calculations were performed with more digitsthan shown in this section.

Therefore, minor differences may occur if the calculations in thissections are followed explicitly.

Revision 3Page 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 whicha pre-existing leakage in the containment structure (e.g., containment liner) exists that can onlybe detected by performing a Type A ILRT. The probability of leakage detectable by a Type AILRT is calculated to determine the impact of extending the testing interval.

The Class 3calculation 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 havebeen detected only during the performance of an ILRT and thus impact risk due to change inILRT 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:

2Pclass3a -- = 0.0092217Multiplying 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 Section5.1.3, additional consideration is made to not apply failure probabilities on those cases that arealready LERF scenarios (i.e., the Class 2 and Class 8 contributions).

Therefore, these LERFcontributions from CDF are removed.

Therefore, the frequency of a Class 3a failure iscalculated by the following equation:

Frequiclass 3a = Pclass3a * (CDFul -Class2u,

-Class8uj)

=2 *(1.61E-S

-5.01E-8 -6.77E-7)217Frequlcass3a

= 1.42E-7Frequ2class3a = Pclass3a * (CDFu2 -Class2u2 -Class8U2) 2 *(1.41E5

-4.29E8 -6.72E7)Frequ2class3a

= 1.23E-7In the database of 217 ILRTs, there are zero containment leakage events that could result in alarge early release.

Therefore, the Jeffreys Non-Informed Prior is used to estimate a failure rateand is illustrated in the following equations:

Probability

= Number of Failures

+ 1/2effreys Failure PrNumber of Tests + 10 + 1/2Pctass3b

-- 21 1- 0.0023217++/-1The frequency of a Class 3b failure is calculated by the following equation:

Frequlass3b

= Pclass3b

• (CDFul -Class2u,

-Class8uj)

= 's *(1.61E5

-5.01E8 -6.77E7)218Frequldcass3b

= 3.52E-8Frequ2c1ass3b Pclass3b

• (CDFu2 -Class2u2 -Class8u2) = *(1.41E5

-4.34E-8 -6.72E7)218Frequ2class3b

= 3.07E-8For this analysis, the associated containment leakage for Class 3a is 1 OLa and for Class 3b is100L,. 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 whichthe 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 5-12and then subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF), calculated below:Revision 3Page 20 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Frequiciass

ý= Frequiciassi

-(Frequlclass3a

-Frequlclass3b)

Frequ2ciassi = FreqU2classi

-(FreqU2class3a

-FreqU2class3b)

Class 2 Sequences.

This group consists of core damage accident progression bins with largecontainment isolation failures.

The frequency per year for these sequences is obtained from theEPRI Accident Class 2 frequency listed in Table 5-12.Class 4 Sequences.

This group consists of all core damage accident progression bins for whichcontainment isolation failure-to-seal of Type B test components occurs. Because these failuresare detected by Type B tests which are unaffected by the Type A ILRT, this group is notevaluated any further in the analysis, consistent with approved methodology.

Class 5 Sequences.

This group consists of all core damage accident progression bins for whicha containment isolation failure-to-seal of Type C test components.

Because the failures aredetected 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 binsfor 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 atest/maintenance evolution.

For CCNPP, this class is defined as the SERF category.

Thefrequency 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 whichcontainment failure induced by severe accident phenomena occurs (e.g., overpressure).

For thisanalysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table5-12.Class 8 Sequences.

This group consists of all core damage accident progression bins in whichcontainment bypass or SGTR occurs. For this analysis, the frequency is determined from theEPRI Accident Class 8 frequency listed in Table 5-12.Table 5-10 -Release Category Frequencies Release Category EPRI Category Unit I Frequency

(/yr) Unit 2 Frequency

(/yr)INTACT Class 1 6.76E-06 5.12E-06LERF Classes 2,7, 8 1.39E-06 1.56E-06SERF Class 6 1.87E-06 1.25E-06LATE Class 1 6.06E-06 6.17E-061Total (CDF) N/A 1.61E-05 1.41 E-051. Unit 2 LATE was quantified at 5E-12 truncation.

The other end states were quantified at 1 E-12.Revision 3 Page 21 of 93Revision 3Page 21 of 93 1 RCA-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 showsthis distribution.

Table 5-11 -Release Category Frequencies CCNPP LERF Description of Outcome EPRI Unit I Unit 2End State Category Frequency

(/yr) Frequency (lyr)LERF01 Containment failure following high- 7 3.66E-07 3.80E-07pressure (HP) vessel breach (VB)LERF02 Containment failure following HP VB 7 4.87E-10 5.45E-08LERF03 Containment failure following low pressure 7 4.25E-1 1 1.89E-07(LP) VBLERF04 Temperature induced (TI) SGTR 8 0.OOE+00 3.17E-09LERF05LERF06LERF07LERF08LERF09LERF10LERF11LERF12LERF13LERF14LERF15LERF16LERF17LERF18Containment failure following LP VBPressure induced (PI) SGTRContainment failure following LP VBLoss of isolation Containment bypassContainment failure following LP VBContainment failure following HP VBContainment failure following LP VBTI-SGTRContainment failure following LP VBPI-SGTRContainment failure following LP VBLoss of isolation Containment bvyass787287778787285.04E-080.00E+001.09E-083.34E-086.68E-071.37E-072.01 E-085.14E-088.75E-091.27E-080.OOE+000.OOE+003.05E-084.94E-109.61 E-080.OOE+009.68E-093.72E-086.56E-075.59E-081.40E-082.85E-081.21 E-087.63E-090.OOE+000.OOE+001.71 E-085.23E-10Contribution to EPRI Classification 2Contribution to EPRI Classification 7Contribution to EPRI Classification 8Total LERF6.39E-086.49E-076.77E-0711.39E-06 5.43E-088.35E-076.72E-071.56E-06Table 5-12 -Release Category Frequencies Release Category EPRI Category Unit I Frequency

(/yr) Unit 2 Frequency

(/yr)INTACT + LATE1 Class 1 1.28E-056 1.13E-056LERF2 Class 2 5.01E-086 4.34E-086SERF3 Class 6 1.87E-06 1.25E-06LERF4 Class 7 6.49E-07 8.35E-07LERF5 Class 8 6.77E-07 6.72E-07Total (CDF) 1.61E-5 1.41E-51. The EPRI Class 1 category consists of INTACT and LATE failures.

A LATE failure is classified as intact dueto 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 failuresas 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 withphenomenological failures as re-categorized in Table 5-11.5. The EPRI Class 8 category consists of the CCNPP assigned LERF contribution associated with bypass orSGTR failures as re-categorized in Table 5-11.6. The level 2 model contains a bounding contribution associated with pre-event containment liner failure.

Topreclude influencing the current detailed assessment, the contribution associated with this failure is adjustedby 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 valuesare the LERF contributions from events FAILLEAK and FAILLEAK_2 for Units 1 and 2, respectively.

Revision 3 Page 22 of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Class123a3b4Descriptic No containment Large containment isoSmall isolation failuresLarge isolation failuresSmall isolation failures

-failTable 5-13 -Baseline Risk Profilern Unit I Frequency

(/yr) Unit 2 Frequency (lyr)failure 1.27E-052 1.12E-052lation failures 5.01 E-08 4.34E-08(liner breach) 1.42E-07 1.23E-07(liner breach) 3.52E-08 3.07E-08ure to seal (type B) El E15 Small isolation failures

-failure to seal (type C) El El6 Containment isolation failures (dependent

failure, 1.87E-06 1.25E-06personnel errors)7 Severe accident phenomena induced failure (early 6.49E-07 8.35E-07and late)8 Containment bypass 6.77E-07 6.72E-07Total 1.61 E-05 1.41E-051.2.E represents a probabilistically insignificant value.The Class 3a and 3b frequencies are subtracted from Class 1to 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 thepopulation 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 CCNPPpopulation dose to EPRI Accident Class. Table 5-15 provides population dose for each EPRIaccident class.The population dose for EPRI Accident Classes 3a and 3b were calculated based on theguidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24] as follows:EPRI Class 3a Population Dose = 10

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

• 3.20E+4 = 3.20E+5Table 5-14 -Mapping of Population Dose to EPRI Accident ClassRelease EPRI Unit I Frequency Unit I Dose Unit 2 Frequency Unit 2 DoseCategory Category

(/yr) (person-rem)

(/yr) (person-rem)

INTACT + Class 1 1.28E-05 3.20E+04 1.13E-05 3.20E+04LATELERF Class 2 5.01 E-08 2.OOE+07 4.34E-08 2.OOE+07SERF Class 6 1.87E-06 7.01E+06 1.25E-06 7.01 E+06LERF Class 7 6.49E-07 5.61E+07 8.35E-07 5.61E+07LERF Class 8 6.77E-07 2.25E+07 6.72E-07 2.25E+07Revision 3 Page 23 of 93Revision 3Page 23 of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-15 -Baseline Population DosesClass Description Population Dose (person-rem) 1 No containment failure 3.20E+042 Large containment isolation failures 2.OOE+073a Small isolation failures (liner breach) 3.20E+0513b Large isolation failures (liner breach) 3.20E+0624 Small isolation failures

-failure to seal (type B) N/A5 Small isolation failures

-failure to seal (type C) N/A6 Containment isolation failures (dependent

failure, personnel errors) 7.01 E+067 Severe accident phenomena induced failure (early and late) 5.61 E+078 Containment bypass 2.25E+071. 10*La2. 100*La5.2.3 Step 3 -Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15YearsThe next step is to evaluate the risk impact of extending the test interval from its current 10-yearinterval 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 IntervalAs previously stated, Type A tests impact only Class 3 sequences.

For Class 3 sequences, therelease magnitude is not impacted by the change in test interval (a small or large breachremains 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 ischanged based on the NEI guidance as described in Section 5.1.3 by a factor of 10/3 comparedto the base case values. The Class 3a and 3b frequencies are calculated as follows:10 2 IFrequclass3alOyr 10 *2 * (CDFu1 -Class2u,

-Class8uj)

= *

• 1.54E-5 = 4.72E-73 217 3 21710 2110 22 = 10

• 2

• 1.34E-5 = 4.11E-7Frequ2class3alOyr

• * (CDFU2 -Class2U2 -C2ass8U2) =3217 3 21710 .5.Frequlclass3blOyr

= 10 0* * (CDFul -Class2u,

-Class8uj)

= *

• 1.54E5 = 1.17E73 218 3 218FreqU20ass~bl~r

= 10 __ * (CDFU2 -Class2U2

-Class8u2)

= 1*

• 1.34E5 = 1.02E-7F~qucas~l~r 3 218 3 218The results of the calculation for a 10-year interval for Units 1 and 2 interval are presented inTables 5-16 and 5-17, respectively.

Risk Impact Due to 15-Year Test IntervalThe risk contribution for a 15-year interval is calculated in a manner similar to the 10-yearinterval.

The difference is in the increase in probability of leakage in Classes 3a and 3b. For thiscase, the value used in the analysis is a factor of 5 compared to the 3-year interval value, asdescribed in Section 5.1.3. The Class 3a and 3b frequencies are calculated as follows:Frequ1c5 i 2 2 2F--ass3alSyr

= -* -* (CDFul -Class2u,

-Class8uj)

= 5 * -* 1.54E5 = 7.08E73 217 217Revision 3Page 24 of 93 IRCA-54001-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 15 2-- 2FreqU2Cass3alSyr is 2 * (CDFu2 -Class2u2 -Class8u2) = 5 *2

• 1.34E-5 = 6.17E-7F~q~lssa5r 3 217 21715 5 5Frequlcass3blyr

• * (CDFul -Class2u,

-Class8uj)

= 5 *

• 1.54E-5 = 1.76E-73 218 218s 1 5 5FreqU2Cass3blSyr 3 2' * * (CDFu2 -Class2u2 -Class8u2) = 5

• 2

• 1.34E-5 = 1.53E-73218 218The results of the calculation for a 15-year interval for Units 1 and 2 are presented in Table 5-18and 5-19.Table 5-16 -Unit I Risk Profile for Once in 10 Year ILRTClass Description Frequency Contribution

(%) Population Population

(/yr) Dose Dose Rate(person-(person-rem) rem/yr)1 No containment failure1 1.23E-05 76.17% 3.20E+04 3.92E-012 Large containment isolation failures 5.01E-08 0.31% 2.00E+07 1.00E+003a Small isolation failures (liner 4.72E-07 2.93% 3.20E+05 1.51 E-01breach)3b Large isolation failures (liner 1.17E-07 0.73% 3.20E+06 3.76E-01breach)4 Small isolation failures

-failure to E1 E El Eseal (type B)5 Small isolation failures

-failure to E1 E1 El Eseal (type C)Containment isolation failures6 (dependent

failure, personnel 1.87E-06 11.61% 7.01 E+06 1.31 E+01errors)7 Severe accident phenomena 6.49E-07 4.04% 5.61 E+07 3.64E+01induced failure (early and late)8 Containment bypass 6.77E-07 4.21% 2.25E+07 1.52E+01Total 1.61 E-05 6.67E+011. £ 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 5-17 -Unit 2 Risk Profile for Once in 10 Year ILRTClass 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-012 Large containment isolation failures 4.34E-08 0.31% 2.00E+07 8.67E-013a Small isolation failures (liner breach) 4.11 E-07 2.92% 3.20E+05 1.32E-013b Large isolation failures (liner breach) 1.02E-07 0.73% 3.20E+06 3.28E-01Small isolation failures

-failure toseal (type B)Revision 3Page 25 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-17 -Unit 2 Risk Profile for Once in 10 Year ILRTClass Description Frequency Contribution

(%) Population Population

(/yr) Dose Dose Rate(person-(person-rem) rem/yr)Small isolation failures

-failure toseal (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+01induced failure (early and late)8 Containment bypass 6.72E-07 4.76% 2.25E+07 1.51 E+01Total 1.41 E-05 7.24E+011. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.Table 5-18 -Unit I Risk Profile for Once in 15 Year ILRTClass 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-012 Large containment isolation 5.01E-08 0.31% 2.00E+07 1.OOE+00failures3a Small isolation failures (liner 7.08E-07 4.40% 3.20E+05 2.26E-01breach)3b Large isolation failures (liner 1.76E-07 1.09% 3.20E+06 5.64E-01breach)4 Small isolation failures

-failure E1 E1 El Eto seal (type B)5 Small isolation failures

-failure El El El Elto seal (type C)Containment isolation failures6 (dependent

failure, personnel 1.87E-06 11.61% 7.01 E+06 1.31 E+01errors)7 Severe accident phenomena 6.49E-07 4.04% 5.61 E+07 3.64E+01induced failure (early and late)8 Containment bypass 6.77E-07 4.21% 2.25E+07 1.52E+01Total 1.61 E-05 6.69E+011. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.Table 5-19 -Unit 2 Risk Profile for Once in 15 Year ILRTClass 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-01Revision 3 Page 26 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-19 -Unit 2 Risk Profile for Once in 15 Year ILRTClass 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-01failures3a Small isolation failures (liner 6.17E-07 4.37% 3.20E+05 1.97E-01breach)3b Large isolation failures (liner 1.54E-07 1.09% 3.20E+06 4.91E-01breach)4 Small isolation failures

-failure F1 El El Eto seal (type B)5 Small isolation failures

-failure El El El Elto seal (type C)Containment isolation failures6 (dependent

failure, personnel 1.25E-06 8.85% 7.01 E+06 8.75E+00errors)7 Severe accident phenomena 8.35E-07 5.92% 5.61E+07 4.69E+01induced failure (early and late)8 Containment bypass 6.72E-07 4.76% 2.25E+07 1.51 E+01Total 1.41 E-05 7.26E+011. 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 LERFThe risk increase associated with extending the ILRT interval involves the potential that a coredamage event that normally would result in only a small radioactive release from an intactcontainment could, in fact, result in a larger release due to the increase in probability of failure todetect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3bcontribution would be considered LERF.Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact ofplant-specific changes to the licensing basis. RG 1.174 [Reference 4] defines very smallchanges in risk as resulting in increases of CDF less than 10-6/year and increases in LERF lessthan 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 metricis 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 intervalextension (consistent with the EPRI guidance methodology).

Based on a 10-year test intervalfrom Tables 5-16 and 5-17, the Class 3b frequency is 1.17E-7/year for Unit 1 and 1.02E-7 forUnit 2; based on a 15-year test interval from Tables 5-18 and 5-19, the Class 3b frequency is1.76E-7 for Unit 1 and 1.54E-7 for Unit 2. Thus, the increase in the overall probability of LERFdue to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is1.41 E-7/year for Unit 1 and 1.23E-7 for Unit 2. Similarly, the increase due to increasing theinterval 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),

theRevision 3Page 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 criteriawhen compared to the original 3-year requirement.

Table 5-20 summarizes these results.Table 5-20 -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 2:15Interval 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-07LERF)ALERF (3 yearbaseli(ea

, 8.22E-08 1.41E-07 7.16E-08 1.23E-07baseline)

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

The increase in the overall probability of LERF due to Class 3b sequences being slightly greaterthan 1 E-7 is not unexpected.

Since the target is exceeded, some refinement is necessary.

Onemethod 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 debrisexpelled from the reactor remains covered with water. Therefore, if the accident sequencecontains containment spray success, the source term is not considered to lead to a large earlyrelease.

The methodology developed in Reference 33 is used for this containment spraysuccess sensitivity.

Excluding INTACT scenarios where containment spray is credited andtherefore scrubbing the source term release results in a frequency reduction.

Conservatisms are further reduced by analyzing the source term release times. Early releasetiming is defined by time short enough that ability to evacuate nearby population is impairedsuch 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 hourmark. 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.

Thesethree MAAP cases are matched with a corresponding plant damage state (PDS) in CCNPP'sLevel 2 notebook

[Reference 32]. Table 5-21 displays CCNPP's PDSs.Table 5-21 -Summary of CCNPP Plant Damage StatesPDS Containment RCS Pressure at Feedwater Pressurizer CHR? AC PowerBypass? Time of Core Availability?

PORV/SRV Status? Available?

Damage?1 No High Not Not stuck open Not Available available Available Not4 No Low Available Not stuck open Available Available Not5 No High Available Not stuck open Available Available 6 No Low Available Not stuck open Available Available Revision 3Page 28 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-21 -Summary of CCNPP Plant Damage StatesPDS Containment RCS Pressure at Feedwater Pressurizer CHR? AC PowerBypass? Time of Core Availability?

PORV/SRV Status? Available?

Damage?7 SGTR N/A N/A N/A N/A N/A8 ISLOCA N/A N/A N/A N/A N/ANot Not9 No High Available Not stuck open Available Available 10 NoHighNot 10 No High Available Not stuck open Available Available Not Not Not14 No High Available Stuck open Available Available ot NNot Not15 No High Available Not stuck open Available Available Not Not16 No High Available Stuck open Available Available Not Not17 No Low Available Not Stuck Open Available Available The HRIF MAAP case models a SBO that leads to a loss of main feedwater.

The analysisassumes a loss of containment heat removal and AC power. The reactor coolant system isisolated and the containment remains intact. Core damage occurs while the reactor coolantsystem is at high pressure.

Based on the information in Table 5-21, this case can be used torepresent PDS 15. Table 2-2 of Reference 32 contains the list of all the Level 1 core damageaccident sequences and how each is mapped to a PDS. Using the correlation of the HRIF caseand the SBO cases that contain a loss of feedwater (PDS 15), it is determined that thefrequency contribution can be removed from the LERF contribution because the release time ofthe 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 areSBO004, SBO005, SBOO10, SBO013, SBOO15, SBOO18, SBOO19, and SB0039.Another MAAP case evaluated is GIOY, which involves a Small LOCA inside containment withan equivalent break size of 0.005 ft2 and the containment isolated.

The reactor coolant system isat high pressure with auxiliary feedwater (AFW) and AC power available; containment aircooling (CAC) provides containment cooling and maintains containment pressure.

Theradionuclide 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 torepresent PDS 5. The following sequences are Small LOCA cases assigned to PDS 5 and areexcluded based on their late release:

SLOCA002, SLOCA003, and SLOCA012.

Another MAAP case evaluated is MRIF, which involves a Small LOCA inside containment withan equivalent break size of 0.02 ft2.Containment is isolated; the reactor coolant pressure ishigh; AFW and containment heat removal are not available; AC power is available.

Thesecharacteristics map to PDS 1. The following sequences are Small LOCA cases assigned toPDS 5 and are removed from the PDS 1 frequency:

TRAN003, TRAN004, TRAN005,TRAN007,

TRAN008, TRAN009,

SLOCA007, and SLOCA01 1 [Reference 18].Revision 3Page 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 adjustedrelease category frequencies after some conservatisms from containment spray success andrelease timing are excluded.

Table 5-22 -Adjusted Release Category Frequencies Release Category EPRI Category Unit I Frequency

(/yr) Unit 2 Frequency

(/yr)INTACT Class 1 3.28E-06 1.18E-06LERF Classes 2, 7, 8 9.51 E-07 1.08E-06SERF Class 6 1.54E-06 5.74E-07LATE Class 1 2.37E-06 2.18E-061Total (CDF) N/A 8.14E-06 5.O1E-061. 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 yieldsthe final results displayed in Table 5-23.Table 5-23 -Impact on LERF due to Extended Type A Testing Intervals with Adjusted CDFILRT Inspection Unit 1: 3 Unit 1:10 Unit 1:15 Unit 2: 3 Unit 2:10 Unit 2:15Interval 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-08LERF)ALERF (3 yearbaseline) 2.65E-08 4.54E-08 1.43E-08 2.46E-08ALERF (10 yearbaseline) 1.89E-08 1.02E-08The adjusted containment spray and PDS inputs allow the Unit 1 and 2 values to be much lessthan 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 thedefinition 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 inputinto 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 ofan accident.

This probability can be expressed using the following equation:

CCFP = 1 -f(ncf)CDFwhere f(ncf) is the frequency of those sequences that do not result in containment failure; thisfrequency is determined by summing the Class 1 and Class 3a resultsSince CCFP is only concerned with a containment failure and not whether the release is smallor large, the Class 1 results without refinement must be used to calculate the CCFP. Table 5-24Revision 3Page 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 testinterval and 15-year test interval is 0.88% for Unit 1 and 0.87% for Unit 2.Table 5-24 -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 2:15Interval 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-05f(ncf)/CDF 0.796 0.791 0.787 0.799 0.794 0.791CCFP 0.204 0.209 0.213 0.201 0.206 0.209ACCFP (3 yearbaseline) 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. Theincrease in the CCFP from the 3 in 10 year interval to 1 in 15 year interval is 0.876% for Unit 1and 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 thisinvestigation is the determination of the total LERF following an increase in the ILRT testinginterval from 3 in 10 years to 1 in 15 years.Calvert Cliffs is transitioning to NFPA 805 licensing basis for fire protection and submitted aLicense 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 inthe NFPA 805 LAR. Compensatory actions have been implemented to reduce the fire risk untilthe modifications are implemented.

The Unit 1 ILRT is scheduled for 2016, which is prior to thescheduled implementation of all the modifications by 2018. It is anticipated that many, but notall, of the NFPA 805 modifications will be completed by the Unit 1 refueling outage. Riskmitigation strategies will be in place for any open modification.

These strategies may be actionsto reduce fire initiating event probabilities, actions to improve suppression probability, and/oractions to recover or protect systems that mitigate core damage and large early releaseaccident sequences.

The Unit 2 ILRT is scheduled for 2023, so the NFPA 805 modifications willbe 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 reduceconservatism in the model, the plant damage state methodology described in Section 5.2.4 wasalso applied to the CDF portion of the Fire PRA model. The following shows the calculation forClass 3b for Units 1 and 2:0.5Frequlciass 3b =ciass3b

• (CDFl -PDSCDFl)

= *- (3.18E-05

-2.20E-5)

= 2.25E-8Freulcassb Plas~b (CF1 D~c~z)218 0.5... cass3b * (CDF2 -PDSCDF2) = * (3.62E-05

-2.26E-5)

= 3.12E-8Revision 3 Page 31 of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Freqciass 3blOyr 10

• Pclassi * (CD F1 -PDSCDF-)

= 1 * * (3.18E-05

-2.20E-5)

= 7.49E-83 3 21810 10 0.5Freqclass 3bloyr 10

• Pclass3b

• (CDF2 -PDScDF2) = 10

• 0- * (3.62E-05

-2.26E-5)

= 1.04E-073 3 218150.Freqclass 3blSyr is

• Pclass3b

• (CDF1 -PDScDFl)

= 5 *0. * (3.18E-05

-2.20E-5)

= 1.12E-073 218Freqcass~b~yr

=150.Freqcls3blSyr is

• Pclass3b

• (CDF2 -PDScDF2) = 5

• 0- * (3.62E-05

-2.26E-5)

= 1.56E-073 218Seismic events were addressed through a simplified seismic PRA in Section 3 of the IPEEE forCCNPP [Reference 35]. The Seismic PRA method screened all the components that met a highconfidence low probability of failure (HCLPF) for the review level seismic event occurring with amagnitude of 0.3g. The remaining components were grouped together as a proxy component.

Itwas assumed that if this proxy component failed it would result in core damage. This method isconsidered conservative.

Table 5-25 shows data from Table 3-6 of the IPEEE Seismic Analysis

[Reference 35].Table 5-25 -Seismic Contribution to Frequencies of Containment Failure Categories Containment Failure Category Associated Seismic CDF (/yr)I. Intact Containment 4.62E-07I1. Late Containment Failure 8.63E-06I1l. Early Small Containment Failure 1.70E-07 to 1.27E-06IV. Early Large Containment Failure 3.13E-07 to 1.41 E-06V. Small Containment Bypass 0VI. Large Containment Bypass 0Total 1.07E-05Note: The Seismic contribution to Containment failure categories III and IV is shown as a range of values. A range is shownbecause the contribution of a certain PDS will be apportioned between the small and large early containment

failures, butthe 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 theIPEEE 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-5

-3.13E-7 -0) = 2.38E-810 10 0.5Freqclass 3bloyr 1" Pclass3b

• (CDF -CatIV -CatVI) = 10

• 0 *(1.07E-5

-3.13E-6-0)=7.92E-8 15Freqclass 3blsyr = is

• PcLass3b

• (CDF -CatIV -CatVI) = 5

• 52° *(1.07E-5

-3.13E-6-0)=

1.19E-73 218CNNPP topographical location presents the opportunity for high wind events. These eventsinclude tornadoes, thunderstorms, freezing precipitation, and hurricanes.

Hurricanes poseapproximately one threat per year and one significant threat per 10 years (Reference 24). Thesenatural disasters are modeled in the internal events model. As shown in Table 5-2 and 5-4 showthat high wind risk is approximately two orders of magnitude lower than fire risk. Since high windrisk is already included in the internal events PRA, no further analysis is necessary to include itscontribution to Class 3b frequency.

The seismic and fire contributions to Class 3b frequencies are then combined to obtain the totalexternal event contribution to Class 3b frequencies.

The change in LERF is calculated for the 1Revision 3Page 32 of 93 1 RCA-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 5-26and 5-27 for Units 1 and 2, respectively.

Table 5-26 -CCNPP Unit I External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from3 per 10 years to 13 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-07Internal Events 1.14E-08 3.78E-08 5.68E-08 4.54E-08Combined 5.76E-08 1.92E-07 2.88E-07 2.30E-07Table 5-27 -CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from3 per 10 years to 13 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-07Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08Combined 6.11E-08 2.04E-07 3.05E-07 2.44E-07The internal event results are also provided to allow a composite value to be defined.

Whenboth the internal and external event contributions are combined the total change in Unit 1 and 2LERF meet the guidance for small change in risk, as it exceeds the 1.OE-7/yr and remains lessthan 1.OE-6 change in LERF for both units. For this change in LERF to be acceptable, totalLERF must be less than 1.OE-5.Conservatively using the highest seismic LERF value and not crediting containment spraysuccess or plant damage state adjustments for the Internal Events or Fire PRA, the total LERFvalues 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 bebetween 1.OE-7 and 1.OE-6.5.3.1.1 Potential Impact from External Events Contribution Using IPEEE Fire AnalysisAn assessment of the impact of external events is also performed using fire risk analysis fromthe 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; asgiven in Section 4.6.8.3 of the IPEEE, the estimated Unit 2 fire CDF is 1.1E-5/yr.

The Level 2results are shown here in table 5-28.Revision 3 Page 33 of 93Revision 3Page 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 5-28 -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-06I1. Late Containment Failure 56.5% 4.07E-05 6.22E-06Ill. Early Small Containment Failure 1.7% 1.21 E-06 1.87E-07IV. Early Large Containment Failure 6.5% 4.67E-06 7.15E-07V. Small Containment Bypass 0.0% 0.00E+00 0.00E+00VI. Large Containment Bypass 0.0% 0.00E+00 0.00E+00Total 7.32E-05 1.10E-05Using 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-5

-4.67E-6 -0)= 1.57E-721810Unit 1: Frequlclass3bloyr

= 3

• Pclass3b

• (CDF -CatIV -CatVI)10 0.5-- * -0 * (7.32E-5

-4.67E-6 -0) 5.24E-73 218Unit 1: Frequlclass3blSyr is

• Pclass3b

• (CDF -CatIV -CatVI)=5 * * (7.32E-5

-4.67E-6 -0)= 7.86E-7218Unit 2: Frequ2cjass3b = Pctass3b

• (CDF -CatIV -CatVI) = .* (1.1E-5 -7.15E-7 -0)= 2.36E-821810E-5 -7.1 E- -0)t.6 -Unit 2: Frequ2class3blOyr 3

• Pciass3b

• (CDF -Cat/V -CatVI)10, * -L- * (1.10E-5

-7.15E-7 -0) = 7.86E-83 21815Unit 2: FreqU2class3blSyr

= 1

• Pclass3b

• (CDF -CatIV -CatVl)= 0 * (1.10E-5

-7.15E-7 -0)= 1.18E-7218As done in Section 5.3.1, the IPEEE seismic and fire contributions to Class 3b frequencies arethen combined to obtain the total external event contribution to Class 3b frequencies.

Thechange in LERF is calculated for the 1 in 10 year and 1 in 15 year cases and the changedefined for the external events in Tables 5-29 and 5-30 for Units 1 and 2, respectively.

Table 5-29 -CCNPP Unit I External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from3 per 10 years to 13 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-07Internal Events 1.14E-08 3.78E-08 5.68E-08 4.54E-08Combined 1.92E-07 6.41 E-07 9.61 E-07 7.69E-07Revision 3 Page 34 of 93Revision 3Page 34 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-30 -CCNPP Unit 2 External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from3 per 10 years to 13 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-07Internal Events 6.14E-09 2.05E-08 3.07E-08 2.46E-08Combined 5.35E-08 1.78E-07 2.67E-07 2.14E-07The internal event results are also provided to allow a composite value to be defined.

Whenboth the internal and external event contributions are combined the total change in Unit 1 and 2LERF meet the guidance for small change in risk, as it exceeds the 1.OE-7/yr and remains lessthan 1.OE-6 change in LERF for both units. For this change in LERF to be acceptable, totalLERF must be less than 1.OE-5.Conservatively using the highest seismic LERF value and not crediting containment spraysuccess or plant damage state adjustments for the Internal Events PRA, the total LERF valuesare 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 bebetween 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 wasperformed for the risk impact assessment for extended ILRT intervals.

As a sensitivity run, theinternal event CDF was used to calculate the Class 3b frequency.

The impact on the Class 3bfrequency due to increases in the ILRT surveillance interval was calculated for steel linercorrosion likelihood using the relationships described in Section 5.1.4. The EPRI Category 3bfrequencies for the 3 per 10-year, 10-year and 15-year ILRT intervals were quantified using theinternal events CDF. The change in the LERF risk measure due to extending the ILRT intervalfrom 3 in 10 years to 1 in 10 years, orto 1 in 15 years is provided in Tables 5-31 and 5-32. Thesteel liner corrosion likelihood was increased by a factor of 1000, 10000, and 100000. Exceptfor extreme factors of 10000 and 100000, the corrosion likelihood is relatively insensitive to theresults.Revision 3 Page 35 of 93Revision 3Page 35 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-31 -Unit 1 Steel Liner Corrosion Sensitivity Cases3b 3b 3b LERF LERF LERFFrequency 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 toyear ILRT) year ILRT) year ILRT) I-per-10) 1-per-15) 1-per-15)

InternalEvent 3B 1.14E-08 3.78E-08 5.68E-08 2.65E-08 4.54E-08 1.89E-08Contribution Corrosion Likelihood 1.15E-08 3.98E-08 6.36E-08 2.84E-08 5.22E-08 2.38E-08X 1000Corrosion Likelihood 1.24E-08 5.76E-08 1.25E-07 4.52E-08 1.13E-07 6.77E-08X 10000Corrosion Likelihood 2.17E-08 2.35E-07 7.42E-07 2.14E-07 7.20E-07 5.06E-07X 100000Table 5-32 -Unit 2 Steel Liner Corrosion Sensitivity Cases3b 3b 3b LERF LERF LERFFrequency 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 toyear ILRT) year ILRT) year ILRT) 1-per-10) 1-per-15) 1-per-15)

InternalEvent 3B 6.14E-09 2.05E-08 3.07E-08 1.43E-08 2.46E-08 1.02E-08Contribution Corrosion Likelihood 6.19E-09 2.15E-08 3.44E-08 1.53E-08 2.82E-08 1.29E-08X 1000Corrosion Likelihood 6.70E-09 3.11E-08 6.77E-08 2.44E-08 6.1OE-08 3.66E-08X 10000Corrosion Likelihood 1.17E-08 1.27E-07 4.01E-07 1.16E-07 3.89E-07 2.74E-07X 1000005.3.3 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 asdescribed in Reference

24. In this sensitivity case, an expert elicitation was conducted todevelop probabilities for pre-existing containment defects that would be detected by the ILRTonly based on the historical testing data.Using the expert knowledge, this information was extrapolated into a probability versusmagnitude relationship for pre-existing containment defects.

The failure mechanism analysisalso 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 expertelicitation 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 tothe potential for large magnitude leakage events.The expert elicitation results are used to develop sensitivity cases for the risk impactassessment.

Employing the results requires the application of the ILRT interval methodology Revision 3Page 36 of 93 1 RCA-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 thecontainment.

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 largeearly release frequency, can be reflected.

For the purposes of this sensitivity, the same leakagemagnitudes 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 theJefferys non-informative prior and the expert elicitation use in the base methodology and thissensitivity case.Table 5-33 -CCNPP Summary of ILRT Extension Using Expert Elicitation Values (from Reference 24)Leakage Size (La)Jefferys Non-Informative PriorExpert Elicitation Mean Probability of Occurrence Percent Reduction 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 5-16 19 for the expertelicitation yields the results in Tables 5-34 and 5-35 for Units 1 and 2, respectively, aredeveloped.

Table 5-34 -CCNPP Unit I Summary of ILRT Extension Using Expert Elicitation ValuesAccident ILRT IntervalClass3 per 10 Years 1 per 10 Years I per 15 YearsBase Adjusted Dose Dose Frequency Dose Frequency DoseFrequency Base (person-Rate Rate RateFrequency 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-042 5.01E-08 5.01E-08 2.OOE+07 1.OOE+00 5.01E-08 1.OOE+00 5.01E-08 1.OOE+003a N/A 5.98E-08 3.40E+03 2.03E-04 1.99E-07 6.78E-04 2.99E-07 1.02E-033b N/A 1.52E-08 3.40E+04 5.17E-04 5.06E-08 1.72E-03 7.60E-08 2.58E-036 1.87E-06 1.87E-06 7.01E+06 1.31E+01 1.87E-06 1.31E+01 1.87E-06 1.31E+017 6.49E-07 6.49E-07 5.61 E+07 3.64E+01 6.49E-07 3.64E+01 6.49E-07 3.64E+018 6.77E-07 6.77E-07 2.25E+07 1.52E+01 6.77E-07 1.52E+01 6.77E-07 1.52E+01Totals 1.61E-05 1.61E-05 1.06E+08 6.58E+01 1.61E-05 6.58E+01 1.61E-05 6.58E+01ALERF(3 per 10 N/A 3.55E-08 6.08E-08yrs base)ALERF(1 per 10 N/A N/A 2.53E-08yrs base)CCFP 20.26% 20.48% 20.64%Revision 3 Page 37 of 93Revision 3Page 37 of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-35 -CCNPP Unit 2 Summary of ILRT Extension Using Expert Elicitation ValuesAccident ILRT IntervalClass 3 per 10 Years I per 10 Years I per 15 YearsBase Adjusted Dose Dose Rate Frequency Dose Rate Frequency Dose RateFrequency 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-032 4.34E-08 4.34E-08 2.OOE+07 8.67E-01 4.34E-08 8.67E-01 4.34E-08 8.67E-013a N/A 5.21 E-08 3.40E+03 1.77E-04 1.74E-07 5.90E-04 2.60E-07 8.86E-043b N/A 5.21 E-08 3.40E+04 1.77E-03 1.74E-07 5.90E-03 2.60E-07 8.86E-036 1.25E-06 1.25E-06 7.01E+06 8.75E+00 1.25E-06 8.75E+00 1.25E-06 8.75E+007 8.35E-07 8.35E-07 5.61E+07 4.69E+01 8.35E-07 4.69E+01 8.35E-07 4.69E+018 6.72E-07 6.72E-07 2.25E+07 1.51E+01 6.72E-07 1.51E+01 6.72E-07 1.51E+01Totals 1.41E-05 1.41E-05 1.06E+08 7.16E+01 1.41E-05 7.16E+01 1.41E-05 7.16E+01ALERF(3 per 10 N/A 1.22E-07 2.09E-07yrs base)ALERF(1 per 10 N/A N/A 8.69E-08yrs base)CCFP 20.21% 21.08% 21.69%The results illustrate how the expert elicitation reduces the overall change in LERF and theoverall results are more favorable with regard to the change in risk.5.3.4 Large Leak Probability Sensitivity StudyThe large leak probability is a vital portion of determining the Class 3b frequency.

CCNPP hadpreviously calculated the large leak probability using the WCAP method. Table 5-36 presentsthe 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 5-36 -CCNPP Large Leak Probabilities Using the WCAP MethodTest Interval WCAP Large Leak EPRI Accident EPRI Accident Class 3bProbability Class 3b Frequency:

Unit 2Frequency:

Unit 13 per 10 years 2.47E-4 1.38E-09 8.21 E-1010 years 7.41E-4 4.05E-09 2.41 E-0915 years 1.11E-3 5.96E-09 3.55E-09Using the same EPRI approach, but with an updated Class 3b frequency calculated from theWCAP large leak probability data, Table 5-37 contains the final results for both units.Revision 3 Page 38 of 93Revision 3Page 38 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 5-37 -Impact on LERF due to Extended Type A Testing Intervals with WCAP CDFILRT Inspection Unit 1:3 Unit 1:10 Unit 1: 15 Unit 2:3 Unit 2: 10 Unit 2:15Interval 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-09LERF)ALERF (3 year 2.67E-09 4.57E-09 1.59E-09 2.73E-09baseline)

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

___These results demonstrate that the EPRI methodology is conservative when used to calculate alarge leak probability as compared to the WCAP method.Revision 3 Page 39 of 93Revision 3Page 39 of 93 1 RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension 6.0 RESULTSThe results from this ILRT extension risk assessment for CCNPP are summarized in Table 6-1for Unit 1 and Table 6-2 for Unit 2.Table 6-1 -Unit I ILRT Extension SummaryClass Dose Base Case Extend to Extend to(person-3 in 10 Years I in 10 Years I in 15 Yearsrem)CDF/Year Person- CDF/Year Person- CDF/Year Person-Rem/Year Rem/Year Rem/Year1 3.20E+04 5.59E-06 1.79E-01 5.46E-06 1.75E-01 5.37E-06 1.72E-012 2.00E+07 3.33E-08 6.66E-01 3.33E-08 6.66E-01 3.33E-08 6.66E-013a 3.20E+05 4.56E-08 1.46E-02 1.52E-07 4.87E-02 2.28E-07 7.30E-023b 3.20E+06 1.14E-08 3.63E-02 3.78E-08 1.21E-01 5.68E-08 1.82E-016 7.01E+06 1.54E-06 1.08E+01 1.54E-06 1.08E+01 1.54E-06 1.08E+017 5.61E+07 2.49E-07 1.40E+01 2.49E-07 1.40E+01 2.49E-07 1.40E+018 2.25E+07 6.68E-07 1.50E+01 6.68E-07 1.50E+01 6.68E-07 1.50E+01Total 8.14E-06 4.07E+01 8.14E-06 4.08E+01 8.14E-06 4.09E+01ILRT Dose Rate from 3a and 3bFrom 3 N/A 1.15E-01 1.96E-01ATotal YearsDose Rate From 10 N/AYearsN/A 8.18E-02YearsFrom 3 N/A 0.282% 0.483%%ADose YearsRate From 10 N/AYearsN/A 0.201%Years3b Frequency (LERF)From 3 N/A 2.65E-08 4.54E-08YearsALERFFrom 10 N/A N/A 1.89E-08YearsCCFP %From 3 N/A 0.326% 0.558%YearsACCFP%From 10 N/A N/A 0.233%YearsRevision 3 Page 40 of 93Revision 3Page 40 of 93 I RCA-54001

-000-CALC-001 Evaluation of Risk Significance of Permanent ILRT Extension Table 6-2 -Unit 2 ILRT Extension SummaryClass Dose Base Case Extend to Extend to(person-3 in 10 Years 1 in 10 Years I in 15 Yearsrem)CDF/Year Person- CDF/Year Person- CDF/Year Person-Rem/Year Rem/Year Rem/Year1 3.20E+04 3.32E-06 1.06E-01 3.25E-06 1.04E-01 3.20E-06 1.02E-012 2.OOE+07 1.84E-08 3.69E-01 1.84E-08 3.69E-01 1.84E-08 3.69E-013a 3.20E+05 2.47E-08 7.89E-03 8.22E-08 2.63E-02 1.23E-07 3.95E-023b 3.20E+06 6.14E-09 1.96E-02 2.05E-08 6.55E-02 3.07E-08 9.82E-026 7.01E+06 5.74E-07 4.02E+00 5.74E-07 4.02E+00 5.74E-07 4.02E+007 5.61 E+07 4.01 E-07 2.25E+01 4.01 E-07 2.25E+01 4.01E-07 2.25E+018 2.25E+07 6.60E-07 1.49E+01 6.60E-07 1.49E+01 6.60E-07 1.49E+01Total 5.01E-06 4.19E+01 5.01E-06 4.19E+01 5.01E-06 4.20E+01ILRT Dose Rate from 3a and 3bFrom 3 N/A 6.19E-02 1.06E-01ATotal YearsDose Rate From 10 N/AYearsN/A 4.42 E-02YearsFrom 3 N/A 0.148% 0.254%%ADose YearsRate From 10 N/AYearsN/A 0.106%Years3b Frequency (LERF)From 3 N/A 1.43E-08 2.46E-08YearsALERFFrom 10 N/A N/A 1.02E-08YearsCCFP %From 3 N/A 0.286% 0.490%YearsACCFP%From 10 N/A N/A 0.204%YearsRevision 3Page 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 withextending the Type A ILRT test frequency to 15 years:Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impactof plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines verysmall changes in risk as resulting in increases of CDF less than 1.OE-06/year andincreases in LERF less than 1.OE-07/year.

Since the ILRT does not impact CDF, therelevant criterion is LERF. The increase in LERF resulting from a change in the Type AILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 4.54E-8/year forUnit 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 avery 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 impactassessment of the extended ILRT intervals.

This results of this calculation meet thesecriteria for both units. Moreover, the risk impact for the ILRT extension when comparedto other severe accident risks is negligible.

" The increase in the conditional containment failure from the 3 in 10 year interval to 1 in15 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 itrepresents 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 yearswas found to lead to an imperceptible increase in risk. The estimated increase in risk isvery small because ILRTs identify only a few potential containment leakage paths thatcannot be identified by Type B or Type C testing, and the leaks that have been found byType A tests have been only marginally above existing requirements.

" Given the insensitivity of risk to containment leakage rate and the small fraction ofleakage paths detected solely by Type A testing, increasing the interval betweenintegrated leakage-rate tests is possible with minimal impact on public risk. The impactof relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated.

Beyondtesting the performance of containment penetrations, ILRTs also test integrity of thecontainment structure.

The findings for CCNPP confirm these general findings on a plant-specific basis considering thesevere accidents evaluated for CCNPP, the CCNPP containment failure modes, and the localpopulation surrounding CCNPP.Revision 3 Page 42 of 93 I RCA-54001

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

An independent PRA peer review was conducted under the auspices of the Pressurized WaterReactor 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 notsignificant using a sensitivity study) have been captured in the PRA Configuration RiskManagement Program (CRMP) database.

On an on-going basis, other potential PRA model anddocumentation 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 PRAmodel. PRA screening is required for all design and procedure changes." New engineering calculations and revisions to existing calculations are reviewed for theirimpact 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 theMaintenance 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 fromthe Individual Plant Examination for Generic Letter 88-20, "Individual Plant Examination forSevere Accident Vulnerabilities."

The model is maintained and updated in accordance withCalvert Cliffs procedures, and has been updated to meet the ASME PRA Standard andRevision 2 of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy ofProbabilistic Risk Assessment Results for Risk-Informed Activities."

The Calvert Cliffs internal events PRA model was peer reviewed in June 2010. All findingswhich had significant impact on this analysis have been addressed.

This assessment isprovided as Table 1. The ILRT application was determined to be an application requiring aCapability Category II PRA model per the Regulatory Guide 1.200 criteria, Revision

2. This isbased on the requirement for numerical results for CDF and LERF to determine the risk impactof the requested change and the fact that this change is risk-informed, not risk-based.

Table 1includes discussion of all findings from the industry peer review along with the assessment andevaluation of the finding that shows that they have either been addressed or have no materialimpact on the ILRT interval extension request.The peer review found that 97% of the SR's 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. Asnoted in the peer review report, the majority of the findings were documentation related.

Of the11 SRs which did not meet Category II or better, seven were related to conservatisms ordocumentation in LERF and two were related to internal floods. There were 39 findings.

Allfindings which could be relevant to the ILRT extension evaluation were updated in the internalevents model used to quantify the Level 2 release states. Thus, with the exception of minorRevision 3Page 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 causesconservative 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 methodsapplied, 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) andRegulatory 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 ofrecord) 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 andobservations (F&Os). The findings and their dispositions are provided in Table 2. All findings arebeing provided and have been dispositioned.

All F&Os that were defined as suggestions havebeen dispositioned and will be available for NRC review. The Fire PRA is adequate to supportthe ILRT extension.

The Calvert Cliffs seismic PRA model is relatively conservative and, other than the highmagnitude acceleration event, is not a dominant contributor.

The Calvert Cliffs high winds PRAmodel is very conservative in the tornado area in that all tornados are grouped into the mostconservative event. PRA risk for tornadoes and high winds are based upon IPEEEvalues. Calvert Cliffs has maintained and updated a high wind PRA model in order to performrisk assessment of tornado missile impacts and hurricane force winds. Although this model hasnot been peer reviewed in compliance with the ASME/ANS RA-Sa-2009

standard, the model isbased upon accepted methodology and utilizes the ASME/ANS RA-Sa-2009 compliant internalevents model. High winds updates are not expected to cause a significant increase in CDF orLERF. A more detailed assessment would be expected to cause a decrease in CDF.Revision 3 Page 44 of 93Revision 3Page 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 ILRTExtension 1-16AS-B3 SystemsSY-B6 AnalysisBased on Sections 2.4 and 2.10 ofthe System Analysis Introduction Notebook (CO-SY-00, Rev. 0) thisSR appears to be met. However,there is a potential issue related tothis SR. Did not find reference toany engineering analysis neededto support Containment Air Cooleroperation when this system isassumed to be available duringLOSP when the containment heatsup prior to electrical recovery.

(This F&O originated from SR SY-B6)Complete The PRA Internal EventsAccident SequenceNotebook, CO-AS-001, Section 3.3, has beenupdated with an engineering analysis of this issue. Theanalysis identifies thatduring the Loss of OffsitePower sequences, theContainment Air Coolers arecredited for SBO conditions where the containment heats up, and then, afterpower recovery, the aircoolers are credited forcontainment pressure andtemperature control.

Forthese accident sequences, offsite power is restored inone hour, and thecontainment pressure andcorresponding saturation temperature remain wellbelow containment designparameters that wouldchallenge the CACs.Furthermore, failure of CACsis not risk significant, due tothe potential availability ofcontainment spray.No impact onILRT analysis.

Subsequent analysis hasfound this issueto be non-significant:

1) thetemperature riseis not likely tochallenge thecontainment aircoolers, and 2)the importance ofthe air coolers issignificantly reduced by theredundant function providedby containment sprayREFERENCE CO-AS-001 Revision 3 Page 45 of 93Revision 3Page 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 ILRTExtension 1-17 IFSO-Al Internal Examined Internal Flooding Complete An engineering analysis has Due to theQU-E3 Flooding Notebook (CO-IF-001, Rev. 1) subsequently been relatively lowSections 3.0 and 3.1. Part of the performed for AFW contribution toInternal Flood analysis may not be discharge piping flooding.

CDF, this floodcomplete for assessing the Aux The fraction of at-power time has no impact onFeedwater 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 beIFSO-A1) screened due to their lowimpact on CDF (<1 E-9).REFERENCE CO-IF-001 1-18 IFSO-A4 Internal Examined Internal Flooding Open Human-induced impacts on No impact onIFEV-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 aof 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 controlon the flood frequency, Section 5.3 database (CRMP), but notof the IF report states that they yet addressed.

were included, but their inclusion should be better documented orreferenced from IF (e.g., a samplecalculation showing humancontribution would be helpful)(This F&O originated from SRIFSO-A4)1-19 IFEV-B3 Internal While some items are included in Open In the Internal Flood No impact onIFPP-B3 Flooding Section 7.0 of the IF report, many notebook, the discussions ILRT analysis.

IFQU-B3 other instances of uncertainties on uncertainties and This is aRevision 3Page 46 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 ILRTExtension 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 PRASection 7.0 nor are the configuration controlimplications of these other database (CRMP), but notuncertainties and assumptions are yet addressed.

discussed.

(This F&O originated from SRIFPP-B3)1-25 DA-C7 Data For the most part actual plant- Complete The ESFAS logic train The low riskspecific data is used as a basis for testing has a very low risk significance ofthe number of demands significance and generally ESFAS logic trainassociated actual plant does not take the logic OOS. testing isexperiences (See basis for DA- The train does go to 2-out- considered toC6), which includes both actual of-3 logic. Occurrences have no impactplanned and unplanned activities, where the train is in 2-out-of-on ILRT analysis.

However, there are a few ESFAS 3 logic is incorporated intotesting and/or other logic channel the PRA Data Analysistesting that are not tracked via the Notebook, CO-DA-plant computer.

001, Section 2.6 and 3.5.For the logic relays there isCreated this F&O on non- a RAW of <1.04 anddocumentation of ESFAS/logic Birnbaum on the order oftrain testing, which needs to 4E-07. Any logic relayinclude actual practice.

unavailability that does notcause the ESFAS channel to(This F&O originated from SR DA- be OOS and bypassed, isC7) therefore of low significance.

REFERENCE CO-DA-001 Revision 3 Page 47 of 93Revision 3Page 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 ILRTExtension 2-7 IFPP-A5 Internal Flood Section 2.3 provides a discussion Complete A walkdown was performed No impact onthat walkdowns used to confirm to assess the susceptibility ILRT analysis.

plant arrangement.

The following to jet impingement or spray This is annote is contained in section 2.3: in rooms 105A and 203. All Internal Floodequipment is considered documentation Unfortunately, the walk-down failed by spray or issue.documentation from the original impingement for floodflooding analysis no longer exists. sources originating in theA plant walk-down was performed room. Notebook CO-IF-001 as a part of this analysis to provide was updated with thisfamiliarity with the plant design as additional documentation.

well as confirm findings from theoriginal walk-down.

This walk- REFERENCE down is documented in a set of CO-IF-001 notes and photographs included inAppendix B.Walkdown photos for room 105Aand 203 show equipment andpotential flood propagation paths.However, there is not enoughspatial information to developspecific targets for floodimpingement or spray.(This F&O originated from SRIFPP-A5)2-9 DA-D4 Data Evidence of meeting this SR at Complete Table 2-6 of the Data No impact onCC-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 aSections 2.1 and 2.7. Found update results for the documentation inconsistencies in the value of total SACMs. However, the issue. Thenumber components of different correct values were used in Internal Eventstypes (for both units) in Table 2-5 the models for peer review.Revision 3Page 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 ILRTExtension of the PRA Data Notebook with the model includesactual 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 updateData 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 beencorrected in CO-DA-001.

Other minor typographical errors were identified andcorrected 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 onAnalysis notebook states that marked up drawings were generated for ILRT analysis.

plant system drawings are each system notebook.

This is anprovided as supplements to the Where Unit 1 and Unit 2 are Internal Eventssystem notebook, which depicts similar, just the Unit 1 documentation the boundary of the system in boundary is depicted.

In issue that hasterms of PRA modeling.

The addition, the system been addressed.

drawings are not in the notebooks.

notebooks include drawingsnippets,

sketches, and(This F&O originated from SR SY- descriptive text that alsoC2) depict the system boundary.

REFERENCES CO-SY-[AII]

Revision 3Page 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 ILRTExtension 3-5 SY-Al 1 Systems The fault tree does not include Complete A bounding sensitivity case This finding doesSY-A6 Analysis potential failures of the AFW was run to include failure of not impact theaccumulator 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 theCDF. Short-term failure of accumulators isthe AFW operation is two orders ofdominated by failure of magnitude lowerelectrical support systems than activeand failure of active hardware failureshardware (i.e. valves and that support theinstrumentation).

The same systemapplicable 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 onSY-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 anthe documentation is complete and Operation review, and these Internal Eventsthorough.

In most cases it clearly were added to the system documentation follows the RG 1.200 SRs. models and system issue that hasIn some places, assumptions were notebooks.

Furthermore, a been addressed.

imbedded in the documentation comprehensive review ofwithout sufficient reference or PRA mechanical systemsjustification.

Examples include:

notebooks and drawingswas performed to identifySI 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 toRevision 3Page 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 ILRTExtension is assumed that only one-fourth Sections 3.4.d of thepump discharge is spilled via the applicable systembreak. For a hot leg break, the notebooks.

entire pump discharge reaches thecore.'SI notebook page 12, 2nd bullet'The maximum time assumed foroperation for the safety injection pumps is 30 seconds following SIAS initiation.'

CO-SY-000 states that eachsystem notebook addresses flowdiversions (where applicable) insection 3.4.d. Although flowdiversions appear to be addressed (for example, the SW notebooktalks about flow diversion),

there isno consistent discussion in eachsystem 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 onthe grouping of components for updated to add additional ILRT analysis.

plant specific failure data. Many of component types and failure The model usedthe groupings appear to take into modes to better reflect for the ILRTaccount differences in such things service conditions.

Service analysis includesas size, type, mission type (e.g., Water and Salt Water the updated dataFW TDP run vs. AFW TDP pumps were broken out. and failurestandby).

However, in some cases, AFW pumps and Safety modes.it is not clear what the basis for the Injection pumps were brokengrouping is. For example, SW out. This resulted in changesMDP RUN and SRW MDP RUN to the associated failureare grouped together even though rates. The change has beenRevision 3Page 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 ILRTExtension they are of different service reflected in the Dataconditions (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 thoughthe two SI pumps are pumpingborated water, while the AFWpump is pumping condensate grade water. No documentation ofthe appropriateness of thesegroupings 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 oneach system are described in the mutual exclusive modeling ILRT analysis.

system notebook section 3.4.e. was performed.

Each The PRA modelSeveral SY notebooks were system notebook and each that was updatedreviewed to determine system model was reviewed as part of thisappropriateness of the mutually to validate the review was usedexclusive cutsets.

All appeared appropriateness of the as the model forreasonable.

A review was modeling and reconcile any the ILRTperformed of the MUTEX gate differences, and to verify analysis.

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

While the majority ofthese are intuitively obvious (e.g.,Revision 3Page 52 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 ILRTExtension 11 Steam Generator Tube Rupture REFERENCE occurs as an IE AND 12 Steam SY-CO-[ALL]

Generator Tube Rupture occurs asan IE), these should be included inan 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 oneach event tree was performed.

reviews was presented to ILRT analysis.

The utility stated that during this the peer review team; The originalreview, cutsets were reviewed to although, the documentation internal eventsdetermine if any mutually exclusive was separate from the cutset reviewevents were contained within formal QU notebook notes have nowcutsets, if any flag settings were package.

A note was added been archived.

inappropriate or if any recoveries to the QU notebook directing were overlooked or added the reader to the location ofinappropriately.

A review of a the cutest review notes andsampling of cutsets did not indicate spreadsheets.

The PRAany inappropriate results.

configuration controlHowever, the QU notebook does procedure, CNG-CM-1.01-not include a discussion of this 3003, requires a review ofreview. cutsets for PRA changes.

In(This F&O originated from SR QU- practice, the top CDF andD3) LERF cutsets are examinedfor 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 onSY-Al0 Initiating The only mention in CO-SC-00l of Diesel Generator modeling ILRT analysis.

Events shared systems between the unitsRevision 3Page 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 ILRTExtension IE-C3 is the SBO EDG, noted in Section was updated as described in The finding hasSC-A4 4.1.2. It states 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 Internalunit. 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, ispreference of 11, then 24, then 12, with redundant safety used in the ILRTthen 23.* This is noted in the EDG equipment out-of-service, analysis.

system notebook but no basis is with a goal to restore at leastprovided.

The procedures do not one 4KV bus. Since 4KVactually have a preference, which Buses 11 and 24 supportyields a potentially non- AFW, those busses wouldconservative analysis.

For have a preference overexample, if there is a LOOP, the Busses 14 and 21, all elseU2 diesels fail to start and the Ul being equal. No unitdiesels 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.

IfSBO diesel would then be aligned there is a conflict in theto U2, and it is non-conservative to order-of-preference, forgive the U1 bus 11 full credit. If example, both 4KV Bus 11such non-conservatism is and 4KV Bus 24 are notnegligible, some analysis should powered, then a 50-50be performed to demonstrate this. probability is assumed as tothe preferred bus.(This F&O originated from SR IE-Al 0) REFERENCE C0-SY-023-024

• Note: Peer review finding was notprecise.

It should have stated buspreference for Unit 1 is 11, then24, 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 onReliability appendix (ESFOHFCISZEFG) was added to the model. A ILRT analysis.

not included in the fault tree sensitivity run with the basic The missingmodels. CCNPP staff noted that it event included in the current basic event hasRevision 3Page 54 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 ILRTExtension had previously been modeled, but model showed no increase been added toinadvertently deleted in an update. in risk. The system notebook the internalCO-SY-048 was updated.

events model(This F&O originated from SR HR- used in the ILRTCl) REFERENCE analysis.

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

gather plant specific design configuration control The review ofinformation, operating practices, database (CRMP), but has condition reportsetc. that could potentially affect the not yet been addressed.

did not identifygeneric flooding frequencies.

In any designresponse to an NRC RAI on the issues orCCNPP ISI program plan, CCNPP operating mentioned a review of Condition practices thatReports that did not find any items would affect thethat would increase the flooding generic floodingfrequency.

frequencies.

The CR review meets part of therequirement, but the SR also callsfor reviews of plant design,operating practices, etc. thatshould be considered.

Theevaluation should be documented in the PRA.(This F&O originated from SRIFEV-A6)Revision 3 Page 55 of 93Revision 3Page 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 ILRTExtension 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 ILRTLE-G4 notebook and quantified in Table uncertainties that could be analysis.

The5-2 of the QU notebook.

However, addressed.

Two changes dominant LERFno discussion of the uncertainties have been implemented to contributors wereor insights from them is provided, address significant reviewed andFor example, Sensitivity 1 shows a uncertainties and reduced model changes74% reduction in LERF, but this LERF. First, a reverse-flow implemented.

large reduction is not investigated, check valve in the CVCS The Calvert CliffsLetdown line was credited LERFAlso, conservatisms in the ISLOCA as a potential ISLOCA contribution isanalyses were discussed in the AS recovery.

Second, a new now similar toreview. SGTR was treated in an human action was added other PWRs.overly conservative manner by with realistic timing forcategorizing all SGTR as LERF. Steam Generator isolation and RCS depressurization (This F&O originated from SR LE- on a SGTR. These and lessF3) significant model updatesresulted in a LERF-to-CDF ratio change fromapproximately 17% toapproximately 10%. Thisnewer ratio is in the typicalrange 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 onLE-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 internalnot by accident progression are noted as such in the eventssequence, phenomena, Level 2 Notebook.

Accident documentation containment challenges or progression sequences are issue.containment failure mode. located in Section 4.2.3 andRevision 3Page 56 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 ILRTExtension Appendix C. The Level 2(This F&O originated from SR LE- notebook has been updatedG3) to point to additional phenomena andcontainment challenges andfailure mode Tables/Figures in the QU NotebookREFERENCE 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 internalRelease limitations in the LE analysis that 001, Revision 2 -added events findingcould impact applications are discussion of results of does not impactdocumented in the QU notebook, impact on application of the the ILRTbut it is not. Given the conservative Unit 2 ILRT extension analysis.

modeling of SGTR and ISLOCA, request.the impact on applications shouldinclude 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 ILRTLE-F2 (per SR LE-F2). The QU notebook as described in the SRs. analysis.

TheLE-C3 discusses the top 20 LERF cutsets After conservatisms were dominant LERF(which total 73% of the total addressed (see discussion contributors wereLERF). It notes conservatism in for F&0 4-19 above), no reviewed andthe cutsets and says it will be significant issues were model changesevaluated in Section 5.2, but is not. identified.

implemented.

Revision 3Page 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 ILRTExtension Section 4.3.6 of the QU notebook The Calvert Cliffscompares the total LERF of REFERENCE LERFCCNPP to St. Lucie, but does not CO-LE-001 contribution iseven break the results down by now similar tocontributor (e.g., SGTR, ISLOCA, other PWRs.etc.).Also, the ASME PRA StandardSRs C-3, C-10 and C-13 require areview of the LERF results forconservatism in the following areas:1. Engineering analyses to supportcontinued equipment operation oroperator actions during severeaccident progression that couldreduce the LERF2. Engineering analyses to supportcontinued equipment operation oroperations after containment failure.3. Potential credit for repair ofequipment.

No such review has beenperformed, despite the largeconservatism noted in thecontainment 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 toRelease more containment penetrations to considered of adding an ILRT analysis.

isolate on CIAS, a feasible operator action in order Modeling of anoperator action is to manually close containment operator action toclose the failed valves from the penetration from the Main manually closeMain Control Room. Control Room to recover failed valves fromRevision 3Page 58 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 ILRTExtension from a containment isolation the main control(This F&O originated from SR LE- failure.

A review of cutsets room would notD7) shows that a recovery is not significantly feasible for top LERF reduce LERF, assequences, because the such an action issequence includes either 1) not feasible fora loss of CR indication,

2) the significant includes a station black-out sequences wherecondition, or 3) includes containment non-recoverable pipe isolation hasbreaks. failed.REFERENCE CO-LE-001 Attachment S5-17 IE-C1 Initiating Bayesian updates of non-time-Complete CENG understands the No impact onIE-C13 Events based LOCA data were improper.

general concern on ILRT analysis.

IE-C4 The small and medium LOCA Bayesian updating of rare The approachfrequencies were obtained from events. However, the used for LOCAdraft NUREG 1829 then Bayesian method used was based on frequencies hasupdated (in App E) with CCNPP a white paper developed by been validated byexperience from 2004 to 2008. The industry experts regarding industry expertsVery Small LOCA prior having LOCA frequencies.

These and is the samealpha = 0.4, Mean = 1.57E-03; was experts included INL, NRC approach as wasBayesian updated to a Posterior and Industry experts.

In used for thehaving a mean value of 7.02E-04.

addition, the approach used NRC's SPARThis represents an excessive drop for the Calvert PRA was the model.associated with CCNPP same as used for the NRCexperience of 4 to 5 years. SPAR model. This issue isSimilarly, the Small and Medium captured in the PRALOCAs were Bayesian updated configuration controlwith the whole industry experience database (CRMP).rcy data. The draft NUREG 1829LOCA frequencies were obtainedRevision 3Page 59 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 ILRTExtension from expert elicitations (not time- REFERENCE based) that included crack CO-IE-001 propagation analysis.

TheBayesian update for VSLOCAused the Alpha parameter and themean value to justify that the priormean was based on 255 rcy. Thismay not have been the basis forthe expert elicitations in NUREG1829.Also, the Medium LOCA frequency may be classified as extremely rare event. It would require noBayesian updating.

The currentCCNPP SLOCA and MLOCAfrequencies are very close eventhough the source data in NUREG1829 indicates a negativeexponential drop in thesefrequencies.

(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 onIE-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 analysisjustify 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 3Page 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 ILRTExtension 5-23 HR-A2 Human The Pre-Initiator HRAs did not Complete It is agreed that the No impact onReliability include the miscalibration of SIT miscalibration of SIT ILRT analysis.

pressure.

For example, in the pressure could have a Given theevent where SIT pressure is negative impact on various pressure of themiscalibrated high, various accident scenarios involving CCNPP SITsaccident scenarios requiring SI are LLOCA and VLLOCA they are onlynegatively impacted.

Add SIT initiators.

However, this required andpressure miscalibrated high or, instrumentation is not providejustify no impact on CDF / LERF. modeled explicitly and is significant benefittherefore deemed included on Large LOCAs.(This F&O originated from SR HR- within the component The frequency ofA2) boundary for the SIT. As a Large LOCAsuch the miscalibration times the pre-probability would be initiator included in the SIT frequency isunavailability, negligible.

REFERENCE CO-HR-001 5-25 SC-C1 Success Simplify the traceability of Tsw. In Complete Where applicable, the Tsw No impact onHR-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 internalprovided as a positive re-statement Criteria notebook (CS-SC- eventsof 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 wasadding Tsw to the success criteria also updated.and linking that to the PCTrancase where Tsw was developed.

REFERENCE Also, in the SC report (Table B-3), CO-HR-001 consider adding the actual time tocore uncovery (or core damage)Revision 3Page 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 ILRTExtension instead of providing a "Yes" entryin 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 onLE-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 CliffsTable 6.11-2 of the Level 2 WCAP percentage of cladding level 2 analysisshowed a potentially significant oxidation is the main factor for internalimpact from Hydrogen burn. used to develop a maximum events initiators.

Provide an estimate of the impact H2 concentration in theof Hydrogen burn on containment containment, and, in turn, apressure.

Use an accident containment pressure isscenario that is likely to produce calculated if the H2larger amounts of H2 with failed completely burns. These arecontainment spray. The optimal then mapped to site-specific time to estimate the impact of containment failureHydrogen 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 theEOF and TSC personnel have A simplifying assumption isconvened and 'are ready to guide made that "no pre-burning ofthe Main Control Room into hydrogen generated in theperiodic Hydrogen burns before core melt progression isthe formation of explosive considered."

Calvert Cliffs'mixtures.

severe action management procedures do include(This F&O originated from SR LE- actions to reduce H2D1) concentration in thecontainment, but theseRevision 3Page 62 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 ILRTExtension actions are not credited inthe PRA model. Also,Containment Spray is notquestioned for the LERFaccident sequences.

Containment Spray is afactor in LATE containment failure accident sequences.

REFERENCES CO-LE-001 5-31DA-D4 DataThe summary table for Bayesianupdated parameters (on Page 53of the PRA Data Notebook, CO-DA-001, Rev. 1) shows the CS-MDP was Bayesian updated withplant experience containing 1failure and Zero run-hours.

TheCCNPP PRA staff responded tothis issue as an isolated case.There is an actual FTR > 1 hrComplete The aforementioned footnote was incorporated into Table 2-6 of CO-DA-001.

REFERENCE CO-DA-001 No impact on theILRT analysis forthis minorinternal eventsdocumentation issue and nochanges wererequired for theCS-MDP failurerate.(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 existingCriteria the sole basis for any success break size analysis is analysis meetscriteria.

However, upon inspection discussed in the Success the intent of theof the PCTran run tables in the SC Criteria notebook.

SR and therefore report appendices, many instances Furthermore, a review was there is noof surrogate or inferred results conducted of this issue; in impact on thewere found. Instead of running addition, TH analyses were ILRT analysis.

specific PCTran calculations to completed to verify theRevision 3Page 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 ILRTExtension cover the whole SLOCA break size break-size ranges. It wasspectrum, intermediate break sizes found that the computerhave been calculated simulations adequately supplemented with expert represented the variousjudgment 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 onAnalysis 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 wasthe PRA model. However, a further event was removed.

All removed from thereview of these items is required remaining AFW equipment model. A reviewfor completeness.

unavailability events in the did not discovermodel 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 thesystem notebook.

NotebookC0-SY-036 was updated.

Areview for concurrent maintenance was previously performed and documented in the Data Notebook.

REFERENCE C0-SY-036 Revision 3Page 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 ILRTExtension 6-8 HR-H2 Human Some recovery actions included in Complete For each screening HRA, No impact onReliability the model (thus credited) are set to the internal events analysis ILRT analysis.

screening values. In the HEP was updated to include a Theevaluation (appendices of the HR specific reference to the documentation report) there are no indications that earlier HRA analysis.

for internalprocedures,

training, or other Included are the applicable events HRAsshaping factors are available on a success criteria for each was updated toplant-specific basis. recovery.

Refer to CO-HR- address this001, Internal Events Human finding.(This F&O originated from SR HR- Reliability

Analysis, and theH2) associated HRA Calculator file.For Fire PRA development, the internal events HRAswith screening values wereanalyzed to assure that theywere sufficiently conservative for firescenarios.

Refer to Section4.1 of CO-HRA-001, FirePRA Human Reliability Analysis.

Documentation forfire HRA actions are similarRevision 3 Page 65 of 93Revision 3Page 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 ILRTExtension to that done for the updatedinternal 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 onReliability in general and will facilitate the ncebooksin Fire PRA. This isupgrdes hoever soe bsicthe reference section soupgrades,

however, some basic HRA designator names and a documentation event names are not consistent d finding.

HRAbetween the HR report and the the HR Calculator, HtR names in thesystem notebooks.

notebook, CAFTA Model model and6.0. 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 onIFSN-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 aand was well documented.

More describing the screening of documentation detail is needed as to why the the containment building finding for thecontainment buildings were from flooding analysis.

Internal Floodscreened from the analysis.

Essentially, the containment notebook.

is designed for LOCAcondition, which screensRevision 3Page 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 ILRTExtension (This F&O originated from SR reactor coolant system andIFPP-A2) related piping system. Otherpiping systems have limitedinventory, are normallyisolated, or have a low flowrate. Reference CO-IF-001.

REFERENCE CO-IF-001 6-14 IFSO-B1 Internal While the flooding calculations Open This is a documentation No impact onIFSN-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 adone, additional documentation of been captured in the PRA documentation data would enhance the IF report. configuration control issue.It appears that the input reports database (CRMP), but notand references are based on yet closed-out.

poorly documented or non-officially revisioned reports and information sources.(This F&O originated from SRIFSN-A9)6-16 IFQU- Internal Walkdowns have been conducted Open This is an internal floods No impact onAl1 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 aIF 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 3Page 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 ILRTExtension (This F&O originated from SRIFQU-A1 1)6-17 IFQU- Internal By including the flooding events Open The level of modeling detail No impact onA10 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 aaccounted 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 faultdocumentation is found related to trees so that the equivalent the IF analysis in the LE report, general transient initiator although the IF report states that (e.g loss of CCW) isthe LERF impacts due to flooding appropriately defined in theare documented and analyzed in transient fault tree. Inthe LE report. addition, cutset reviewshave not revealed the(This F&O originated from SR current modeling to beIFQU-A10) deficient in this regard.This documentation of theabove basis is captured inthe PRA configuration control database (CRMP),but not yet addressed.

6-18 HR-H2 Human The system time window Tsw for Complete It was determined that the No impact onReliability post initiator HRAs was frequently text in Section 3.1.5.7 was ILRT analysis.

Asassociated with 'core damage'.

incorrect and does not described in thisPost initiator HRAs that appear in capture how stress is F&O for internalthe top cutsets may require actually applied in the EPRI events, the stresssuccess criteria linked to beginning HRA Calculator.

CO-HR-001, levels in theof core uncovery (about 20 Internal Events PRA Human model areminutes before 'core damage').

Or, Reliability

Analysis, has appropriate, butthe operator actions that may fall been updated to show the updates to theinto that final 20-minute time stress level applied to each documentation Revision 3Page 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 ILRTExtension period should be overridden to HFE and the justification for are required.

Theassume a high stress level. While stress selection.

Also internal eventssection 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 hasbeen provided for inclusion (This F&O originated from SR HR- in a future update of theH2) HRA notebook.

For the Fire PRAdevelopment, the internalevents HRA stress levelswere carried forward.

Asdescribed in CO-HRA-001, Fire PRA Human Reliability

Analysis, additional stresseswere evaluated andincorporated due to the fireinitiator.

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 onReliability 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 isValves SI-4142, 4143'. This does not impact station operable withoutmanual action was not modeled in operability.

The Safety the manualthe PRA. The CCNPP PRA staff Injection Pumps and action to shut theprovided reasonable response to Containment Spray Pumps RWT outletthis issue. Based on CR-2009-will not fail if the RWT valves. There is005581, there is no impact on isolation valves do not close no impact onRevision 3Page 69 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 ILRTExtension pump operability.

Also, the staff with a RAS signal. A design internal eventswill continue to track the CR. If margin issue has been CDF. The issuethere are any changes to the identified.

This issue has was added to thedisposition of pump operability, been added to the plant's plant's marginthen a new HRA may be added to margin management management the PRA model (if warranted).

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, wouldcapture any design changesconcerning 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 onReliability combinations, it assumes that CVCOHFBHEOTA-B-8HRS ILRTactions occur in the order of the and AFW0HF-CC-SGDEC-analysis.

Thetime delay (Td). However, the time 8HR were added to model new HRA eventsdelay is not the same for all Td variances where CST are notsequences, and care must be depletion occurs early and significant.

taken to make the combinations when it occurs later. Thisappropriate for the sequences in accounts for appropriate which they occur. Page 88 of the sequencing of events.HRA notebook indicates this wasconsidered, since the Td was This specific issue with timemodified for events occurring prior delay and CST depletion to reactor trip, and also for OTCC has been addressed andafter SG overfill.

However, not all incorporated into the PRAoccurrences have been model. An updatedaddressed.

The combination dependency analysis hasRevision 3Page 70 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 ILRTExtension examined by the review team is been performed, whichCombination 770 (OTCC after CST includes these new HRAdepletion).

In this event the CST events. The dependency depletion should come first. analysis shows that thesenew HRA actions are not(This F&O originated from SR HR- significant for CDF orG7) LERF. A PRA configuration control database (CRMP)item has been initiated toformally incorporate theupdated 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 ondocumentation and result files. incorrectly flagged as being ILRT analysis forSB0037 and SB0038 sequences SBO sequence 37 (offsite this internalappear 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 />) events1, 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 tablesA2) B-2, C-1, and D 1 in CO-QU-001. Spot-check wasperformed to identify othererrors. In CO-QU-002, fixedsequence 12 table 4.2-5,which incorrectly showedsequence 37 instead of 38.Revision 3 Page 71 of 93Revision 3Page 71 of 93 I RCA-54001

-000-CALC-001 F&O ID SREvaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review -Facts and Observations Topic Finding/Observation Status Disposition REFERENCE CO-QU-001 CO-QU-002 Impact to ILRTExtension Revision 3 Page 72 of 93Revision 3Page 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

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT AnalysisPP- PP-B3 Plant Complete The containment is partitioned CO-PP-001, Calvert Cliffs Fire PRA Plant No impact toB3- 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 affectsthis was accounted for in the partitioning of the containment into two the FPRA plantPRA 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 inPAUs. There is no justification this buffer zone were identified as qualified given for the 20 foot cables that were verified to be encasedassumption.

The turbine deck within marinate covered raceways.

Theis continuous from unit 1 to unit covers prevent the cables from becoming2. This area is divided into 2 potential combustibles and therefore arePAUs, TURB1 and TURB2, but not considered intervening combustibles.

there is no discussion for thebasis of the partitioning.

The unit 1 and unit 2 Turbine Deck wasFinding level of significance is walked down to assess for the acceptability baseparaon wredithino r site of the Appendix R partitioning into distinctseparation with no requisite PAUs. The boundary was assessed tojustification.

have at least a 20-foot separation betweenMaintain 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 deckthe fire modeling.

Document volume assessed for damaging hot gasthe spatial separation and no layer development.

The partitioning wasintervening combustibles for found acceptable and consistent withthe turbine deck. NUREG/CR-6850, Section 1.5.2, wheremain turbine decks are typical applications where spatial separation has beencredited.

PP- PP-B5 Plant Complete The water curtain in the CCW The Component Cooling Water room water No impact to35- 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 affectsReiin3Pae7 f9Revision 3Page 73 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

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysiswas that the water curtain was issued by the NRC in response to Calvert the FPRA plantpart 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 adirect reference to theirAppendix R program as thebasis for the acceptability forthis or provide a design basisjustification for the watercurtain and document that inthe PP notebook if theAppendix R program reference cannot be found.PP- PP-B7 Plant Complete

1. The walk down A table was created to correlate the No impact toB7- 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 affectsQLS- Screening page 561 of the walkdown documentation, to the plant analysis unit the FPRA plantdocumentation 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 areasinaccessible such as: #23 The facilities and rooms that were notCharging 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 3Page 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

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT AnalysisCorridor.

These areas appear were generated.

For areas that were notto 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 accesswere inaccessible and did not otherwise denied),

The reason forhave a confirmatory walkdown.

inaccessibility was added to Table 17.Finding level assessed due tothe incompleteness of thewalkdown documentation.

1. Prepare a table thatcorrelates the PAUs from thePP notebook with the areanomenclature used in thewalkdown documentation.

2. Complete the walkdowns, particularly for areas screenedin the QLS task.CS- CS-B1 Fire PRA Complete Current Breaker coordination The breaker coordination study has been No impact toB1- 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 affectsand order to receive a category II all PRA common power supplies are the FPRA plantLocation met for CS-BI. assumed to meet -or will meet -the Cablecoordination requirements of NFPA 805, Selection Complete the breaker except as noted in CO-CS-001, Fire PRA analysis andcoordination study. Cable Selection Notebook.

As described in the item hasthe cable selection

notebook, two 120VAC beenlighting panels are not validated as completed.

coordinated, and these panels areassumed to fail for all Fire PRA scenarios.

Also, as described in the PRA notebook abreaker for 480V motor control centerRevision 3Page 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

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT AnalysisMCC101BT has not been validated ascoordinated.

This breaker, 52-10150, ismodeled so that a fire-induced electrical fault on the breaker's power cabling will failMCC101BT.

Finally, the notebook identifies that selected 120V power panels havecoordination issues, but that these will beaddressed by design changes andreferenced in Attachment S -Modifications and Implementation Items.PRM-B3-01PRM-B3PRM-B4PRM-B5FirePRA/Plant ResponseModelComplete The FPRA model did notaddress events involving lossof both HVAC trains to theMCR, long term heatup ofMCR and need for operatoractions outside the MCR tocompensate for the loss ofelectronic controls in the MCR,which was assumed as aCCDP of 1.0 for the plant. Thebasis for excluding thispotential Core Damagesequence was addressed inquestions to the Calvert CliffsPRA team. This sequence is anew sequence outside thecurrent FPRA model logictrees.Consider using a combination of MCR heatup calculations todefine the time when operators would leave MCR and considerLoss of Control Room HVAC can affect theoperability and availability of equipment inthe control room and cable spreading room. As described in Calvert PRA SystemAnalysis Notebooks CO-SY-002, CO-SY-017, and C0-SY-030, loss of HVAC ismodeled to have the effect of increasing the failure rate of 120VAC and 125VDCinstruments and controls in the cablespreading room. For the control room,degradation of the 125VDC system is usedas a conservative surrogate for controlroom I&C degradation.

Loss of Control Room HVAC andsubsequent temperature increases mayadversely affect operator responses.

Themodel reflects degradation of humanactions by the degradation of the 125VDCsystem used for instruments and controls.

Loss of Control Room HVAC is notexpected to cause abandonment byNo impact tothe ILRTanalysis, as theloss of MCRHVACmodeling hasbeenimplemented inthe modelsused in theILRT analysisRevision 3 Page 76 of 93Revision 3Page 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

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisa recovery action for restoring cooling the MCR.operations staff of the control room due tohigh temperatures.

On complete loss ofHVAC with no mitigation, such as no use ofemergency fans, calculation CA02725shows a CR temperature of 123 deg F at24-hours.

While this is a challenging environment, this temperature is assessedas insufficient to solely drive a completeCR abandonment scenario.

NUREG/CR-6738 describes operational experience where operators will continue to occupy thecontrol room even under severeenvironments.

Operations staff says that in consideration of high temperatures in the control room,that Operations would do what was neededto keep the cores safe and covered.

Thesite safety director says that for atemperature of 123 deg F, the site wouldimplement a mitigation strategy whichwould include stay-times, assessment ofindividuals for heat-related conditions, useof ice vests, and call-in of additional qualified operations staff to rotate into thecontrol room.The above discussion was included in CO-SY-030, Control Room HVAC PRA SystemNotebook.

Revision 3 Page 77 of 93Revision 3Page 77 of 93 1 RCA-54001

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-OOO-CALC-OO1 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review -Facts and Observations

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT AnalysisFSS- FSS-A5- A501FireScenarioSelection andAnalysisComplete A range of ignition source /target set combinations hasbeen represented forunscreened PAUs. Thesecombinations are identified inrelevant calculation sheets forunscreened PAUs. In somePAUs, sub-PAUs are definedand damage from a potential fire within the sub-PAU isaddressed.

However, it is notclear how or why damagewould be limited to thespecified sub-PAU becausethere are no physical barriersbetween specified sub-PAUs.

The documentation is suchthat it cannot be determined ifthe selected fire scenarios provide reasonable assurance that the risk contribution ofeach unscreened PAU can becharacterized.

Another issuethat influences the potential forfire propagation across sub-PAU boundaries is that thetemperature measurement locations specified in thedetailed FDS fire modelingevaluations do not generally coincide with locations wheremaximum temperature areFDS modeling was used for fire scenarioevaluations in the Cable Spreading Roomsand Switchgear Rooms. In both cases,thermocouple location was adjusted asidentified in F&O FSS-D3-02.

For the CSR,consequences were divided into scenarios based on mitigation potential.

First, if thescenario was suppressed by the Halonsystem then the limit of damage was basedon what was predicted by FDS in terms oftemperature and energy. If it wasunsuppressed it went to total room bum,which assumes failure of all targets in theroom, regardless of the initial scenarioboundary.

For the Switchgear Room FDSanalysis, the analysis was updated to addclarity to the analysis.

A discussion of theapplication of sub-PAUs has been added toAddendum 1 to CO-FSS-004, Fire PRADetailed Fire Modeling Notebook.

Damagewas not limited to specified sub-PAUs.

Specific examples of the treatment of firegrowth and the application of sub-PAUshave been provided.

As described in CO-FSS-004, the sub-PAUanalysis included spatial information fromwalkdown, along with engineering

judgment, to determine if fire sources couldfail additional components, cables, or othercombustibles, potentially leading to moredamage to surrounding equipment orcables. For scenarios that leveraged FDTNo impact tothe ILRTanalysis, asthis affects theFPRA modeland the item iscomplete.

Revision 3 Page 78 of 93Revision 3Page 78 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

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisexpected (e.g., within the fire modeling, the issue related to whether theplume). analysis had correctly addressed theimpact of transients along the edge of aAs a consequence, for some boundary interface for a sub-PAU.

Afire csecenarios dame tom comparable consideration was also relatedfire scenarios damage to to secondary combustion and oil fires.targets is not predicted when it Resolution involved selection of severalshoulfed b mased oither. representative PAUs for a sensitivity studyspecified damage criteria, that expanded the existing sub-PAUs andSomthe bascios artepercreed examined secondary ignition potential.

on the basis of temperature measurements that do notrepresent conditions at targetswithin the fire plume. (SeeF&O FSS-D3-02)

This couldhave a significant impact onthe potential for firepropagation across sub-PAUboundaries and needs to bediscussed more thoroughly.

FSS- FSS- Fire Complete There were indications that The PAUs were considered representative No impact toA5- A5 Scenario Calvert Cliffs had the tools and of the work performed based on several the ILRT01 Selection information in place to properly criteria.

The analysis indicated that the analysis, asand evaluate the propagation of methods mentioned were indeed this affects theAnalysis fires across the sub-PAU appropriate.

Sub-PAU impacts did not FPRA modelboundaries given no physical change from the expanded assessment and the item isbarriers but there were no and that secondary ignition was bounded complete.

examples showing that this by the existing analysis and wasevaluation was performed or appropriately addressed.

The analysis wasany explicit descriptions of how incorporated into the documentation forthey were performed in CO-FSS-004.

general.

The concern here isthat without an explicitRevision 3Page 79 of 93 I RCA-54001

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-OOO-CALC-OO1 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review -Facts and Observations

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisdescription of the process forevaluating the spread of firesacross sub-PAU boundaries with no physical barriers anddetailed

examples, there is thepotential that in the future, newpeople updating the PRA maynot know that they have toevaluate this.Calvert Cliffs needs to describetheir process for evaluating firegrowth and propagation between sub-PAUs and asapplicable, between PAUs.Specific examples of the sub-PAU fire growth need to beprovided.

If fire propagation from sub-PAU to sub-PAU wasnot treated, Calvert Cliffsneeds to evaluate all sub-PAUs to determine if there isany potential for fire spreadand then model the potential for spreading fires and fordamage occurring across sub-PAU boundaries.

FSS- FSS- Fire Complete Where used, the FDS model FDS modeling was used for fire scenario No impact toD2- D2 Scenario was generally used with a level evaluations in the Cable Spreading Rooms the ILRT01 Selection of grid resolution that was and Switchgear Rooms. analysis, asand below the level of grid this affects theAnalysis resolution documented in the FPRA modelRevision 3Page 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

-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT AnalysisNUREG-1824 Verification and For the Cable Spreading Room FDS fire and the item isValidation study for the FDS scenarios, a grid study was performed on complete.

model. A validation study was the updated FDS model. The studynot conducted to support the recommended a grid size that was withinuse of this lower level of grid the range in NUREG/CR-1824.

That gridresolution.

Grid resolution has size was used for CSR FDS scenarioa bearing on the results of FDS evaluations.

The study and results werecalculations.

Grid resolutions incorporated into CO-FSS-004, Fire PRAoutside the validation range in Detailed Fire Modeling Notebook.

NUREG-1824 should bejustified and validated.

The Unit 1 27' and 45' Switchgear RoomsIncrease the level of grid were updated to increase the level of gridInreaselutin the ll o U g id e resolution to a value that is within theresolution in the FDS PAU Fire vldto ag ouetdiEvaluations (C0-FSS-004 R1) validation range documented inEvthaluationsidCreFolutiRi)

NUREG/CR-1824.

Results calculated inso that the grid resolution is the Unit 1 FDS models were applied to Unitdocumented in NUREG-1824.

2. Results of the updated model areincorporated into C0-FSS-004 asAddendum 1.FSS- FSS- Fire Complete This SR is not met because FDS modeling was used for fire scenario No impact toD3- D3 Scenario detailed FDS fire modeling evaluations in the Cable Spreading Rooms the ILRT01 FSS- Selection evaluations of PAUs 302, 306, and Switchgear Rooms. analysis, asB2 and 311, 317, 407 and 430 assume this affects theFSS- Analysis that material surfaces are For the Cable Spreading Room FDS fire FPRA model"inert."

As noted on p. 44 of and the item isD4 C0-FSS-004 R1, this scenarios, the Unit 1 CSR was modified to complete.

include actual material properties andassumption was made "... so sensitivity analysis.

Actual materialthe PAU structure (walls, floor, properties were used in the updatedor ceiling) itself would absorb U 1CSR FDS model rather than the priorany ceiling)mitself vaoulbsofrb use of "inert" material conditions.

Adiabatic any heat from the various fire conditions were used for any items withscenarios, producing a more material properties that are unknown or ofRevision 3Page 81 of 93 I RCA-54001

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisconservative or worst caseresult for all fire scenarios' impacts to the components and cables within the PAUmodel. As such, no detailedmaterial properties wererequired to be defined in FDSfor the scenarios to functioncorrectly."

However,specification of materialsurfaces as "inert" in FDS doesnot prevent heat absorption into material surfaces.

On thecontrary, this specification maintains material surfaces atambient temperature in FDS,which tends to maximize heatabsorption into these surfaces.

To prevent heat absorption intomaterial

surfaces, they shouldhave been specified as"adiabatic" rather than as"inert."

The "inert" parameter inFOS maximizes heat transferto surfaces rather thanminimize it. This can result ina high uncertainty to bound the analysisand prevent heat transfer into thoseobjects.

The CSR FDS model wasexecuted and the results compared to thebaseline results.

This study was thendocumented in FSS-004.

The results wereapplied to Unit 2 CSR. This study was thendocumented in FSS-004, Fire PRADetailed Fire Modeling Notebook.

The Unit 1 27' and 45' Switchgear Roomswere updated to specify representative material properties as referenced byNUREG 1805. This adjustment enabled theanalysis to obtain more realistic estimates of environmental conditions for these firescenarios.

Results calculated in the Unit 1FDS models were applied to Unit 2.Results of the updated model areincorporated into CO-FSS-004 asAddendum 1.lower calculated gastemperatures.

Specify materials surfaces as"adiabatic" rather than as"inert" in EDS to prevent themfrom absorbing heat in order toachieve the stated goal ofRevision 3Page 82 of 93 I RCA-54001

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisproducing a more conservative or worst case result. This mayprove to be overlyconservative, in which casespecification of realistic material properties could beused to achieve more realistic estimates of environmental conditions for these firescenarios.

FSS-D3-02FSS-D3FSS-A5FireScenarioSelection andAnalysisComplete Temperature measurement locations specified in thedetailed FDS fire modelingevaluations do not generally coincide with locations wheremaximum temperature areexpected (e.g., within the fireplume). As a consequence, forsome fire scenarios damage totargets is not predicted when itshould be based on thespecified damage criteria.

Some scenarios are screenedon the basis of temperature measurements that do notrepresent conditions at targetswithin the fire plume.Re-run FDS simulations withtemperature measurement probes located within the fireFDS modeling was used for fire scenarioevaluations in the Cable Spreading Roomsand Switchgear Rooms.For the Cable Spreading Room FDS firescenarios, new measurement devices wereincluded in the updated U1CSR FDSmodel. The thermocouples were placeddirectly above the fire source in theupdated FDS model and the scenarios re-evaluated.

The results were applied to Unit2 CSR. This study and the results werethen documented in FSS-004, Fire PRADetailed Fire Modeling Notebook.

The Unit 1 27' and 45' SWGR rooms wereupdated to alter the location of thethermocouples such that the centerline plume temperature was recorded and usedto determine target impacts.

ResultsNo impact tothe ILRTanalysis, asthis affects theFPRA modeland the item iscomplete.

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisplume or use other fire calculated in the Unit 1 FDS models weremodeling tools such as FDTs applied to Unit 2. Results of the updatedto 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 toD8- D8 Scenario evaluated for detailed fire evaluations in the Cable Spreading Rooms the ILRT01 Selection modeling cases that use FDS. and Switchgear Rooms. analysis, asand This fire detection timing is For the updated Cable Spreading Room this affects theAnalysis then used to estimate FDS fire scenarios, cable tray obstructions FPRA modelautomatic fire suppression were placed in the ceiling area of the and the item istiming 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 modeldetection timing is based on under the new cable tray obstructions inmodeling that does not include the vicinity of the fire source. The scenarios obstructions located beneath were re-evaluated.

The results werethe ceiling that could have an applied to Unit 2. A sensitivity study wasimpact on fire detector also performed.

The study and newresponse.

The fire detection scenario results were incorporated into CO-timing is also based on an FSS-004, Fire PRA Detailed Fire Modelingunjustified assumption Notebook.

regarding the type of smokedetectors installed in theaffected PAUs. Obstructions to The Unit 1 27' and 45' SWGR rooms werethe flow of fire gases can have also updated to include significant an impact on smoke obstructions such as cable trays and beamconcentrations and velocities, pockets within the switchgear rooms.which in turn influence smoke Results calculated in the Unit 1 FDSdetector response.

Without models were applied to Unit 2. Results andincluding such obstructions in details of this analysis are documented infire modeling simulations, their CO-FSS-004 as Addendum 1.Revision 3Page 84 of 93 I RCA-54001

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisimpact on fire detection timesis not evaluated.

Include obstructions locatedbeneath the ceiling for theaffected fire scenarios in orderto evaluate their impact on firedetection timing. Providejustification for the selection ofthe type of smoke detectorspecified in the FDSsimulations for these firescenarios.

FSS- FSS- FireF3-01 F3 ScenarioSelection andAnalysisComplete To achieve CC Il/111 for this SR,a quantitative assessment ofthe risk of the selected firescenarios involving a) exposedstructural steel and b) thepresence of a high-hazard firesources must be completed consistent with the FQrequirements including thecollapse of the exposedstructural steel and anyattendant damage. Such anassessment has not beendone or was not documented in a readily discernible manner.This has a potential impact onfire risk quantification.

The Turbine Building was reviewed forpotential fire scenarios where structural steel can be adversely affected.

From thescenarios

examined, those that candamage structural steel were selected forfurther analysis.

The frequency, severityfactor and non-suppression probability ofeach scenario were developed andincluded in the Structural Failure AnalysisNotebook.

These impacts were then addedto FRANX database and quantified as partof the final Fire PRA risk quantification inFire Quantification Notebooks CO-FRQ-001 and CO-FRQ-002.

No impact tothe ILRTanalysis, asthis affects theFPRA modeland the item iscomplete.

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT AnalysisComplete a quantitative assessment of the risk of theselected 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 toG4- G4 Scenario effectiveness, reliability and credited passive fire barrier features in the the ILRT01 Selection availability of credited passive multi-compartment analysis.

At Calvert analysis, asand fire barrier features has not Cliffs, the fire barriers are verified to be this affects theAnalysis been documented in the multi- effective through test procedures.

An FPRA modelcompartment analysis.

To unreliability value was applied to all and the item isachieve a CC II capability normally closed doors that represents the complete.

assessment, the effectiveness, probability of the door being propped openreliability and availability of given a fire in the exposing compartment.

credited passive fire barrier The probability of finding a failed sealedfeatures must be assessed.

wall penetration is assumed to be verysmall to warrant propagation scenarios.

Adiscussion of the effectiveness, reliability, Assess the effectiveness, and availability of fire barriers was added toreliability and availability of CO-FSS-008, Calvert Fire PRA Multi-credited passive fire barrier Compartment Analysis.

features and document thisassessment.

FSS- FSS- Fire Complete The effectiveness, reliability Active fire barriers were evaluated as No impact toG5- G5 Scenario and availability of credited effective in studies used to support the ILRT01 Selection active fire barrier features have Appendix R analysis.

An unreliability value analysis, asand not been quantified in the has been applied to all normally open, self this affects theAnalysis multi-compartment analysis.

closing dampers and doors; A discussion FPRA modelTo achieve a CC II capability of the effectiveness of credited active fire and the item isassessment, the effectiveness, barriers was added to CO-FSS-008, Calvert complete.

reliability and availability of Fire PRA Multi-Compartment Analysis.

credited active fire barrierRevision 3Page 86 of 93 I RCA-54001

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisfeatures must be quantified.

Quantify the effectiveness, reliability and availability ofcredited active fire barrierfeatures and document thisassessment.

HRA- HRA- Human Complete Improve documentation of the CO-HRA-001, Fire Human Reliability No impact toB2- B2 Reliability adverse operator actions notebook, was updated to detail the the ILRT01 Analysis needed to address the impact adverse operator actions added to the analysis, asof grounded or shorted model following the fire AOP review this affects theelectrical buses that might process.

Table 3 was added to Section 2.2 FPRA modelhave an impact on other plant detailing each basic event, set to true (1.0) documentation buses if not isolated and re used in the model to annotate the adverse and the item isenergized 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 theactions are modeled as inputs HFEs added to the model as part of theto FRANX. AOP review, including actions to restoreAC power to busses lost due to fire failureProvide new tables listing the sequences.

actions considered orreferences to specificlocations.

HRA- HRA- Human Complete Documentation for what was CO-HRA-001, Fire Human Reliability No impact toEl- El Reliability done was very good, however,

Notebook, was updated detailing the Alarm the ILRT01 Analysis the details for not selecting any Response Procedure review process.

analysis, asspurious alarms is not clear. Table 12 was expanded to show the ARP this affects theThe documentation of the review of alarm impact and operator FPRA modeladverse actions put into the interview notes for CR annunciators that andRevision 3Page 87 of 93 I RCA-54001

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysismodel as "true" are not in theHRA report, actions identified in the cutset reviews are notclearly identified, rational fornot using specific HFEs in theRCP trip actions, for identifying actions from procedures andthe process for assigning uncertainty range for thecombos. Doesn't permitverification of the rational forjudgments made in decidingwhat is in and out of the FireHRA. Also, from thecalculation viewpoint the needto know the use of allmanpower requirements duringearly time after fire initiator fordependency analysis.

Enhance documentation of thespecific issues needed toreproduce the assumptions and calculations used in theHRA.could result in a manual reactor trip. Noannunciators were identified that wouldcause the operator to terminate a systemsor components operation based solely onthe alarm itself, but several were identified that could potentially result in the operatortripping the Unit unnecessarily.

CO-HRA-001 was also updated to detail theadverse operator actions added to themodel following the fire AOP reviewprocess.

Table 3 was added to Section 2.2detailing each basic event, set to true (1.0)used in the model to annotate the adverseoperator actions in the model. Theseinclude actions to de-energize electrical busses to isolate them from potential shorts and grounds.

Table 2 shows theHFEs added to the model as part of theAOP review, including actions to restoreAC power to busses lost due to fire failuresequences.

New HFEs added as part of the cutsetreview process are identified in Table 1 ofCO-HRA-001, Fire Human Action Reliability notebook.

These are annotated with"identified during the development of thePRM Notebook."

The cutset reviews aredescribed in CO-QNS-001, Fire PRAQuantitative Screening Notebook.

A newdependency analysis was performed afterthe new HFEs were added to the model,documentation and the item iscomplete.

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisensuring new dependency combinations are considered.

Additional information was added to Table1 of the Human Reliability AnalysisNotebook, CO-HRA-001, detailing whyeach HFE was either retained or removed.For example, event FGAFWOSGTRISOL, Operator Feeds Affected SG with SGTR toAssure Heat Removal, was "Not retainedfor fire scenarios, because these actionsare SGTR specific.

Modeling was notnecessary to ensure these actions did notappear in the cutsets, because the SGTRinitiator is not being used for firescenarios."

Combination event multipliers are used incutsets of multiple HEP actions to accountfor dependencies between HEP actions.

Toaccount for the uncertainty in HEP actions,an uncertainty parameter is added to theHEP action. When performing uncertainty

analysis, the uncertainty parameters forcombination events is increased proportionally when they are multiplied bythe combination event multipliers.

Based on interviews, there are sufficient non-control room personnel for firerecovery actions.

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisoperators assigned to the fire brigade.There were no identified staffing issues orinterferences between operators performing fire recovery actions andmembers of the fire brigade.FQ- FQ-A1 Fire Risk Complete Treatment of 0 CCDPs The fire risk quantification process has No impact toAl- Quantificati scenarios is not clear and been updated in notebooks CO-FRQ-001 the ILRT01 on appears to result in an and C0-FRQ-002 to address the issue with analysis, asunderestimate of total risk (the FRANX fire scenarios having a zero this affects theunderestimate appears to be conditional probability for CDF and LERF. FPRA modelsmall based on the sensitivity and the item isevaluations performed):

1. When documented analysis shows that complete.

1 -with respect to opposite unit selected fire scenarios for one unit arequantification, use CCDP for screened from impact for the opposite unitreactor trip initiator unless (typically, no trip would be initiated),

thenconfirmation of no trip is that scenario may be excluded from thedocumented; 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 roomabandonment

2. F&O PRM-B3-01 identifies the concern3 -for scenarios with limited with loss of Control Room HVAC withimpact 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 revealedon reactor trip initiator that loss of CR HVAC is not expected tocause abandonment by the operations staffMore than 50% of the of the control room due to highscenarios have a 0 CCDP but temperatures.

Loss of CR HVAC andno clear discussion of the subsequent temperature increases mayadversely affect operator responses, andRevision 3Page 90 of 93 I RCA-54001

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisbasis for the 0 CCDP is the model reflects degradation of humanprovided, actions with loss of CR HVAC. CO-SY-030, Control Room HVAC PRA SystemTreatment of 0 CCDPs Notebook, was updated to include thisTrenatmtof 0discussion.

scenarios:

1 -with respect to opposite unitquantification, use CCDP for 3. The new quantification processreactor trip initiator unless described in the FRQ notebooks is toconfirmation of no trip is assure a nominal conditional value isdocumented; 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 y HVAC losvalue to assure resolution in the scenarioscenarios, 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 the3 -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 2cutsets below truncation limit,apply a baseline CCDP basedon reactor trip initiator FQ- FQ-B1 Fire Risk Complete We observed zero CCDPs for The fire risk quantification process has No impact toB1- Quantificati some PAU CDF and LERF been updated in notebooks C0-FRQ-001 the ILRT01 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 theeliminated loss of HVAC to the conditional probability for CDF and LERF. FPRA modelMCR as a potential MCR and the item isabandonment sequence.

1. When documented analysis shows that complete.

Treatment of 0 CCDPs selected fire scenarios for one unit arescenarios:

screened from impact for the opposite unit1 -with respect to opposite unit (typically, no trip would be initiated),

thenquantification, use CCDP for that scenario may be excluded from theRevision 3Page 91 of 93 I RCA-54001

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisreactor trip initiator unlessconfirmation of no trip isdocumented; 2 -address use of 0 CCDP forcontrol room HVAC lossscenarios, apply CCDPconsistent with control roomabandonment (F&O FQ-Al-01(F))3 -for scenarios with limitedimpact with a 0 CCDP, due tocutsets below truncation limit,apply a baseline CCDP basedon reactor trip initiator Allowing zero CCDPs allowsscenarios in the fire model toquantify with no contribution tothe CDF or LERF value andthis under represents thosefrequencies especially whenconsidering delta riskevaluations.

Replace the zero entries withthe lowest CCPD for a planttrip with only random failures ofthe safety equipment as in theinternal events model. Wediscussed this with the CalvertCliffs PRA team and some ofthe zeros are due to fire areasin one unit potentially contributing to the CCDP of theopposite unit's fire risk quantification.

Otherwise, a nominal conditional probability, as described in item 3 below,would apply.2. F&O PRM-B3-01 identifies the concernwith loss of Control Room HVAC withcontrol room abandonment.

As discussed in more detail with the resolution to PRM-B3-01, subsequent investigation revealedthat loss of CR HVAC is not expected tocause abandonment by the operations staffof the control room due to hightemperatures.

Loss of CR HVAC andsubsequent temperature increases mayadversely affect operator responses, andthe model reflects degradation of humanactions with loss of CR HVAC. CO-SY-030, Control Room HVAC PRA SystemNotebook, was updated to include thisdiscussion.

3. The new quantification processdescribed in the FRQ notebooks is toassure a nominal conditional value iscalculated for these low significant scenarios by 1) recalculating the zero-conditional scenarios at a lower truncation value to assure resolution in the scenariocutset file and conditional probabilities, and/or to 2) use a baseline conditional probability for CDF and LERF for theRevision 3 Page 92 of 93Revision 3Page 92 of 93 I RCA-54001

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-FindingsF&O SR Topic Status Finding Disposition Impact onID ILRT Analysisopposite unit. With the internal events reactor trip initiating vent -exception of these cases a IEOPT for Unit 1 or IEOPT-2 for Unit 2method for handling the zerosneeded to be developed andapplied in the frequency quantifications.

Revision 3 Page 93 of 93Revision 3Page 93 of 93 ATTACHMENT (4)REGULATORY COMMITMENT Calvert Cliffs Nuclear Power PlantSeptember 18, 2014 ATTACHMENT (4)REGULATORY COMMITMENT The table below lists the action committed to in this submittal.

Any other statements in thissubmittal are provided for information purposes and are not considered to be regulatory commitments.

Regulatory Commitment DateComplete repairs to address the effects of concrete weathering to July 1, 2015Unit 1 and Unit 2 containment structure dome area.Complete repairs to address concrete delamination around the sloped July 1, 2015surface above the equipment hatch on Unit 1 and Unit 2.1