ML070530296

Technical Specification Change Request - Type a Test Extension

ML070530296
Person / Time
Site:	Limerick
Issue date:	02/20/2007
From:	Cowan P Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, NRC/NRR/ADRO
References
Download: ML070530296 (90)

Text

Exelon Nuclear www.exeloncorp.com 200 Exelon Way Cl Kennett Square, PA 19348 10 CFR 50.90 February 20,2007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Limerick Generating Station, Units 1 and 2 Facility Operating License Nos. NPF-39 and NPF-85 NRC Docket Nos. 50-352 and 50-353

Subject:

Technical Specifications Change Request -Type A Test Extension Pursuant to 10 CFR 50.90, Exelon Generation Company, LLC (EGC) hereby requests an amendment to Appendix A, Technical Specifications, of Facility Operating License Nos. NPF-39 and NPF-85. The proposed change modifies Technical Specifications (TS) 6.8.4.g, Primary Containment Leakage Rate Testing Program. Specifically, the proposed change will revise TS 6.8.4.g to reflect a one-time extension of the containment Type A Integrated Leak Rate Test (ILRT) from 10 to 15 years. This one-time extension will require the Type A ILRT to be performed no later than May 15, 2013, (Unit 1) and May 21,2014, (Unit 2).

EGC requests approval of the proposed changes by February 20,2008. Once approved, the amendment shall be implemented within 60 days. The proposed changes have been reviewed by the Plant Operations Review Committee and approved by the Nuclear Safety Review Board. No new regulatory commitments are established by this submittal.

Technical Specifications Change Request -Type A Test Extension February 20,2007 Page 1 We are notifying the Commonwealth of Pennsylvania of this application for changes to the Technical Specifications by transmitting a copy of this letter and its attachments to the designated State Official.

If any additional information is needed, please contact Tom Loomis at (610) 765-5510.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 20fhof February 2007.

Respectfully, Pamela B. cowan Director, Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: (1) Evaluation of Proposed Change (2) Markup of Proposed Technical Specification Page Change (3) Retyped Page for Technical Specification Change (4) Risk Assessment for LGS, Units 1 and 2 To Support ILRT (Type A)

Interval Extension Request cc: R. R. Janati, Commonwealth of Pennsylvania S. J. Collins, Administrator, Region 1, USNRC S. Hansell, USNRC Senior Resident Inspector J. Shea. Proiect Manaaer. USNRC

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGE

ATTACHMENT 1 Evaluation of Proposed Change CONTENTS

SUBJECT:

Type A Test Extension 1.o INTRODUCTION

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Regulatory Requirements/Criteria

6.0 ENVIRONMENTAL CONSIDERATION

7.0 PRECENDENT

8.0 REFERENCES

ATTACHMENT 1 Evaluation of Proposed Change 1.O INTRODUCTION This letter is a request to amend Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LGS), Units 1 and 2. The proposed change modifies Technical Specifications (TS) 6.8.4.g, Primary Containment Leakage Rate Testing Program.

Specifically, the proposed change will revise TS 6.8.4.g to reflect a one-time extension of the containment Type A Integrated Leak Rate Test (ILRT) from 10 to 15 years. This one-time extension will require the Type A ILRT to be performed no later than May 15, 2013 (Unit 1) and May 21,2014 (Unit 2).

Exelon Generation Company, LLC (EGC) requests approval of the proposed change by February 20,2008. Once approved, the amendment shall be implemented within 60 days.

2.0 PROPOSED CHANGE

The proposed change would revise TS 6.8.4.9 (Primary Containment Leakage Rate Testing Program) of the LGS, Unit 1 Technical Specifications to add the following statement:

I, as modified by the following exception to NEI 94-01, Rev. 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J:

a. Section 9.2.3: The first Type A test performed after May 15, 1998 shall be performed no later than May 15, 2013.

Additionally, the proposed change would revise TS 6.8.4.g (Primary Containment Leakage Rate Testing Program) of the LGS, Unit 2 Technical Specifications to add the following statement:

, as modified by the following exception to NEI 94-01, Rev. 0, Industry Guideline for Implementing Performance-BasedOption of 10 CFR 50, Appendix J:

a. Section 9.2.3: The first Type A test performed after May 21, 1999 shall be performed no later than May 21, 2014.

3.0 BACKGROUND

The proposed change involves a one-time extension to the ten (10) year frequency of the performance-basedleakage rate testing program for Type A tests as required by Nuclear Energy Institute (NEI) 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (Reference 1). The most recent containment Type A Integrated Leak Rate Tests (ILRTs) for LGS, Units 1 and 2 were performed on May 15, 1998 and May 21, 1999, respectively, and would need to be performed no later than Refueling Outage 1R12 (Unit 1) in 2008, and 2R10 (Unit 2) in 2009. The proposed exception would allow the next Type A ILRTs to be performed within fifteen (15) years (i.e., May 15, 2013 (Unit 1) and May 21, 2014 (Unit 2)) from the most recent Type A ILRT as opposed to the current ten (10) year frequency.

This one-time extension will result in the following:

0 Perform a Type A ILRT no later than May 15,2013 (Unit 1) and May 21,2014 (Unit 2).

ATTACHMENT 1 Evaluation of Proposed Change Page 2 of 11 0 A substantial cost savings will be realized by deferring the Type A test for an additional five (5) years. Cost savings have been estimated for each outage at approximately $1.1 million, which includes labor, equipment and critical path outage time needed to perform the test.

4.0 TECHNICAL ANALYSIS

4.1 10CFR 50, Appendix J, Option B The testing requirements of 10 CFR 50, Appendix J provide assurance that leakage through the containment, including systems and components that penetrate the containment, does not exceed allowable leakage rate values specified in the TS and Bases. The allowable leakage rate is limited such that the leakage assumptions in the safety analyses are not exceeded. The limitation of containment leakage provides assurance that the containment would perform its design function following an accident, up to and including the design basis accident.

10 CFR 50, Appendix J was revised, effective October 26, 1995, to allow licensees to choose containment leakage testing under Option A, Prescriptive Requirements, or Option B. Amendment Nos. 118 and 81 for LGS, Units 1 and 2 permit implementation of 10 CFR 50, Appendix J, Option B (Reference 2). TS 6.8.4.g currently requires the establishment of a leakage testing program in accordance with 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program implements the guidelines contained in RG 1.163 which specifies a method acceptable to the NRC for complying with Option B by approving the use of NEI 94-01, subject to several regulatory positions stated in RG 1.163.

10 CFR 50, Appendix J, Option B,Section V.B.3 specifies that RG 1.163, or other implementing documents used to develop a performance-based leakage testing program must be included, by general reference, in the plants TS. Additionally, deviations from guidelines endorsed in a regulatory guide are to be submitted as a revision to the plants TS. Therefore, this application does not require an exemption from 10 CFR 50, Appendix J, Option B.

The adoption of the Option B performance-basedcontainment leakage rate testing program did not alter the basic method by which Appendix J leakage rate testing is performed or its acceptance criteria, but it did alter the test frequency of containment leakage testing in Type A, B, and C tests. The required testing frequency is based upon an evaluation which utilizes the as-found leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained.

The allowable frequency for the Type A ILRT is based, in part, upon a generic evaluation documented in NUREG-1493, Performance-Based Containment Leak-Test Program (Reference 3).

NUREG-1493 made the following observations with regard to changing the test frequency:

Reducing the Type A ILRT frequency to once per twenty years was found to lead to an imperceptible increase in risk. The estimated increase in risk is small because

ATTACHMENT 1 Evaluation of Proposed Change Page 3 of 11 Type A ILRTs identify only a few potential leakage paths that cannot be identified by Type B and C testing, and the leaks that have been found by Type A ILRTs have only been marginally above the existing requirements. Given the insensitivity of risk to containment leakage rate, and the small fraction of leakage detected solely by Type A ILRTs, increasing the interval between Type A ILRTs has minimal impact on public risk.

While Type B and C tests identify the vast majority (i.e., greater than 95%) of all potential leakage paths, performance-based alternatives are feasible without significant risk impacts. Since leakage contributes less than 0.1 percent of overall risk under existing requirements, the overall effect is very small.

The required surveillance frequency for Type A ILRTs in NEI 94-01 is at least once per ten years based on an acceptable performance history (i.e., two consecutive periodic Type A ILRTs at least 24 months apart or refueling cycles where the calculated performance leakage rate was less than 1.O La). Based on the ILRT history discussed below, the current test interval is 10 years.

4.2 Integrated Leak Rate History Type A ILRT testing is performed to verify the integrity of the containment structure.

Industry test experience has demonstrated that Type B and C tests detect a large percentage of containment leakage and that the percentage of containment leakage detected only by integrated containment leakage testing is very small. Results of previous LGS, Units 1 and 2 Type A ILRTs demonstrate that the LGS, Units 1 and 2 containment structure remains essentially a leak tight barrier and represents minimal risk to increased leakage. These plant specific results support the conclusions of NUREG-1493. The specific results for the LGS, Units 1 and 2 Type A ILRTs are as follows:

a. LGS, Unit 1 ILRT Results Year O/owt/daV 1984 0.1642 I 1987 I 0.1469 I I 1990 I 0.2870 I I 1998 I 0.3070 I

b. LGS, Unit 2 ILRT Results J O/owt/dav 1989 0.2310 1993 0.1 836 1999 0.3272 The LGS, Units 1 and 2 TS 6.8.4.g limits the maximum allowable primary containment leakage rate to 0.5% Wday at Pa (44.0 psig).

ATTACHMENT 1 Evaluation of Proposed Change Page 4 of 11 4.3 Plant Design and Operational Performance The containment system, designed by Bechtel Power Corporation, limits the release of radioactive materials to the environs subsequent to the occurrence of a postulated LOCA so that the offsite doses are below the "reference values" stated in 10 CFR 100.

The design employs the drywelVpressure-suppressionfeatures of the BWR/Mark II containment concept. The containment consists of a dual barrier: the primary containment, and the secondary containment. The primary containment is a steel-lined reinforced concrete pressure-suppression system of the over-and-under configuration.

The secondary containment is the enclosure that encloses the reactor, and its primary containment, and fuel storage areas.

The primary containment is a seismic Class I structure and is designed to withstand the jet forces resulting from a rupture of a reactor coolant system pipe.

The primary containment has provisions for rendering the containment atmosphere non-flammable by reducing and maintaining the oxygen content to less than 4 percent during normal and accident conditions.

The Concrete Containment consists of an 8 ft thick reinforced concrete basemat, a 6 ft-2 in. thick reinforced concrete cylindrical Suppression Chamber wall, and a 6 ft-2 in. thick reinforced concrete conical Drywell wall, which provide containment, structural support, and radiation shielding functions.

The liner plate is 1/4 inch thick steel. It consists of several sections: the cylinder, dome, and floor. These sections are connected by horizontal channels and angles in order to provide a leak tight barrier in the Containment.

Penetrations are located in the Containment so that systems can pass through the pressure boundary while the Containment function is fulfilled. Each penetration consists of a pipe sleeve with an annular ring welded to it, and embedded in the concrete to resist normal operating and accident loads. The pipe sleeve is also welded to the liner plate as a seal to prevent leakage.

The Containment is equipped with a 12 ft.-2 in. diameter equipment hatch in the Drywell wall, a 12 ft.-2 in. diameter equipment hatch/personnel lock in the Drywell wall, two 4-ft.

4-in. diameter access hatches in the Suppression Chamber wall and a 3 foot dian eter CRD removal hatch in the Drywell wall.

The LGS Updated Final Safety Analysis Report (UFSAR) Section 6.2.1 describes functional requirements and capabilities of the containment design including the ir ternal design pressure of 55 psig.

4.4 Containment Inspections As approved in the Reference 4 NRC Safety Evaluation Report, LGS aligned the lnservice Inspection (ISI) and Containment Inservice Inspection Intervals (CISI). As a result, the next LGS, Units 1 and 2 IS1 and CIS1 intervals began on February 1, 200'7.

Additionally, LGS, Units 1 and 2 comply with the 2001 Edition through the 2003 Addenda for both the IS1 and the CIS1 program.

ATTACHMENT 1 Evaluation of Proposed Change Page 5 of 11 IWL examinations were performed in the first CIS1 interval in accordance with the 1992 Edition, 1992 Addenda of the ASME Section XI Code. These exams were performed in accordance with the five (5) year frequency as defined in IWL-2400.

The results of the most recent Unit 1 IWL inspections of concrete revealed no reportable indications. These inspections were completed in 1RIO (2004). The next IWL concrete containment inspections are scheduled to be completed prior to March 2009, in accordance with the requirements of the 2001 Edition, 2003 Addenda, of ASME Section XI, as modified by 10CFR50.55a.

The results of the most recent Unit 2 IWL inspections of concrete revealed no reportable indications. These inspections were completed in 2R08 (2005). The next IWL concrete containment inspections are scheduled to be completed prior to March 2010, in accordance with the requirements of the 2001 Edition, 2003 Addenda, of ASME Section Xi, as modified by 10CFR50.55a.

The results of the most recent Unit 1 IWE examinations have been completed per the code requirements. One (1) recordable indication of a pit in the suppression pool steel liner was isolated from further corrosion by performing a qualified coating repair. The remaining wall thickness under this pit was greater than the required design minimum wall thickness. Previously, one other less severe pit was similarly isolated.

The results of the most recent Unit 2 IWE examinations have not identified any recordable indications; however, the suppression pool has not yet been inspected as part of the first interval IWE inspections. This inspection is scheduled to be completed in 2R10 (2009).

There are no IWE augmented inspections required for either Unit 1 or Unit 2.

There are no relief requests being developed for this interval that will impact containment inspections.

NRC Information Notice 92-20 (Inadequate Local Leak Rate Testing) addresses the inability to obtain valid local leak rate test results on penetrations which are designed with a stainless steel, two-ply bellows. There are no bellows of similar design within the LGS, Units 1 and 2 Appendix J scope.

LGS implements a safety-related coatings program that ensures qualified coating systems are used inside primary containment. The program assures that safety-related coatings are selected, procured, applied and inspected in a manner that conforms to the applicable 10 CFR 50 Appendix 6 criteria. Unqualified coatings are controlled and tracked to ensure that emergency core cooling systems (ECCS) will not be adversely affected by coating debris following an accident, and to assure coatings will not cause any adverse effects on Systems, Structures, and Components (SSCs) safety functions.

The program objective is to conform to licensee commitments made in response to Generic Letter 98-04. Coatings are also monitored in accordance with a formal Maintenance Rule (10 CFR 50.65) condition monitoring program. Engineering reviews and evaluates the results of coating condition examinations performed by examiners qualified in accordance with ASTM D 4537, 1991 Edition.

ATTACHMENT 1 Evaluation of Proposed Change Page 6 of 11 Based on the above discussion, the ASME Section XI containment inspections and the safety-related coatings program are intended to provide a high degree of assurance that any degradation of the containment structure is identified and corrected before a containment leakage path is introduced.

4.5 Risk Analysis As discussed in Attachment 4, the Probabilistic Risk Assessment results demonstrate a very small impact in risk associated with the one-time extension of the containment Type A ILRT from 10 to 15 years. The risk assessment follows the guidelines from NEI 94-01 (Reference I), the methodology used in EPRl TR-104285 (Reference 5), the NEI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals (References 6 and 7), the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a request for a plants licensing basis as outlined in Regulatory Guide 1.I 74 (Reference 8), and the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-inducedleakage of steel liners going undetected during the extended test interval. The format of this document is consistent with the intent of the Risk Impact Assessment Template for evaluating extended integrated leak rate testing intervals provided in the December 2005 EPRl final report (Reference 9).

The following is a brief summary of some of the key aspects of the Type A ILRT interval extension risk analysis from 10 to 15 years:

o Regulatory Guide 1.I 74 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.I 74 defines very small changes in risk as resulting in increases of CDF (Core Damage Frequency) below 10-6/yr and increases in LERF (Large Early Release Frequency) below 10-7/yr.

Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in internal events LERF resulting from a change in the Type A ILRT test interval from three in ten years to one in fifteen years is estimated as 4.31 E-8/yr using the NEI guidance as written, and at 1.43E-8/yr using the EPRl Expert Elicitation methodology. In either case, the estimated change in LERF is determined to be very small using the acceptance guidelines of Regulatory Guide I.174.

o The change in Type A test frequency to once-per-fifteen-years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.066 person-rem/yr using the NEI guidance, and drops to 0.013 person-rem/yr using the EPRI Expert Elicitation methodology. Therefore, in either case, the risk impact when compared to other severe accident risks is negligible.

o The increase in the Conditional Containment Failure Frequency from the three in ten year interval to one in fifteen year interval is about 1.2% using the NEI guidance, and drops to about 0.4% using the EPRI Expert Elicitation methodology. Although no official acceptance criteria exist for this risk metric, it is judged to be very small.

o Since the increase in LERF falls well below the small change category using the acceptance guidelines of Regulatory Guide 1.I 74, a detailed examination of the external events impact is not required, nor would it change the conclusions from this assessment.

ATTACHMENT 1 Evaluation of Proposed Change Page 7 of 11 Therefore, increasing the Type A ILRT interval to 15 years is considered to be insignificant since it represents a very small change to the Limerick Generating Station risk profile.

4.6 Primary Containment Pressure Suppression Testing In the Reference 2 NRC Safety Evaluation Report which approved the use of 10 CFR 50, Appendix J, Option B for LGS, Units 1 and 2, the NRC evaluated the revisions to the scheduling of the drywell-to-suppressionchamber bypass leakage test and the drywell-to-suppression chamber vacuum breaker leakage test. The amendment evaluated the proposal to extend the drywell-to-suppressionchamber bypass leakage test to a ten-year frequency. This change also included conducting the drywell-to-suppression chamber vacuum breaker leakage tests during those refueling outages when the drywell-to-suppressionchamber bypass leakage test is not performed (24-month frequency). This requirement is contained in TS 4.6.2.1 .e, which requires that the drywell-to-suppressionchamber bypass leak tests be conducted to coincide with the Type A test (ILRT).

This proposed change will extend the drywell-to-suppressionchamber bypass leakage test frequency to once in 15 years. A review of the past test history for the drywell-to-suppression chamber bypass leakage test has identified no failures. Therefore, extending this test to a 15 year frequency is acceptable. The following are the test results:

Unit 1 (Acceptance 0.005 sq. ft.1 -

Unit 2 (Acceptance 0.005 sq. ft.1 1984 - 0.00026 1989 - 0.000069 1987 - 0.000051 1993 - 0.000076 1990 - 0.000278 1999 - 0.000012 1998 - 0.000075 No frequency change is required for the drywell-to-suppressionchamber vacuum breaker leakage tests, because these tests are conducted independently of the Type A ILRT.

Additionally, the proposed changes do not modify the acceptance criteria of either of these tests.

5.0 REGULATORY ANALYSIS

5.1 NO SIGNIFICANT HAZARDS CONSIDERATION Exelon Generation Company, LLC (EGC) has evaluated the proposed change to the Technical Specifications (TS) for Limerick Generating Station (LGS), Units 1 and 2 and has determined that the proposed changes do not involve a significant hazards consideration and is providing the following information to support a finding of no significant hazards consideration.

ATTACHMENT 1 Evaluation of Proposed Change Page 8 of 11 Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed change will revise TS 6.8.4.9 (Primary Containment Leakage Rate Testing Program) of the LGS, Units 1 and 2 TS to reflect a one-time extension to the Type A Integrated Leak Rate Test (ILRT) as currently specified in the Technical Specifications.

This change will extend the requirement to perform the Type A ILRT from the current requirement of 10 to 15 years, which is no later than May 15, 2013 for LGS, Unit 1 and is no later than May 21 , 2 0 1 4 for Unit 2.

The function of the containment is to isolate and contain fission products released from the reactor coolant system following a design basis Loss of Coolant Accident (LOCA) and to confine the postulated release of radioactive material to within limits. The test interval associated with Type A ILRTs is not a precursor of any accident previously evaluated.

Type A ILRTs provide assurance that the LGS, Units 1 and 2 containments will not exceed allowable leakage rate values specified in the TS and will continue to perform their design function following an accident. The risk assessment of the proposed change has concluded that there is an insignificant increase in Large Early Release Frequency, Person-Rem, and Conditional Containment Failure Frequency. Additionally, containment inspections have also been performed which demonstrate the continued structural integrity of the primary containment and will be performed in the future as required by the ASME Code.

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

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No The proposed change for a one-time extension of the Type A ILRTs for LGS, Units 1 and 2 will not affect the control parameters governing unit operation or the response of plant equipment to transient and accident conditions. The proposed change does not introduce any new equipment, modes of system operation or failure mechanisms.

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

Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No The integrity of the containment penetrations and isolation valves is verified through Type B and Type C local leak rate tests (LLRTs) and the overall leak tight integrity of the containment is verified by a Type A ILRT, as required by 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors. These tests are performed to verify the essentially leak tight characteristics of the containment at the

ATTACHMENT 1 Evaluation of Proposed Change Page 9 of 11 design basis accident pressure. The proposed change for a one-time extension of the Type A ILRT does not affect the method for Type A, B or C testing or the test acceptance criteria.

EGC has conducted a risk assessment to determine the impact of a change to the LGS, Units 1 and 2 Type A ILRT from 10 to 15 years. This risk assessment measured the impact to the Large Early Release Frequency, Person-Rem, and Conditional Containment Failure Frequency. This assessment indicated that the proposed LGS, Units 1 and 2 Type A ILRT interval extension has a very small change in risk to the public and is an acceptable plant change from a risk perspective.

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

Based upon the above, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

5.2 REGULATORY REQUIREMENTS/CRITERIA 10 CFR 50.36, Technical specifications, provides the regulatory requirements for the content required in a plants Technical Specifications (TS). 10 CFR 50.36(~)(5),

Administrative cont roIs , requires provisions reIating to organizat ion and management ,

procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner be included in a plants TS.

Additionally, 10 CFR 50, Appendix J, Option B,Section V.B.3, Implementation, specifies that the regulatory guide or other implementing documents used to develop a performance-based leakage testing program must be included, by general reference, in the plants TS. Additionally, deviations from guidelines endorsed in a regulatory guide are to be submitted as a revision to the plants TS.

The proposed change will revise TS 6.8.4.g to reflect a one-time extension to the LGS, Units 1 and 2 Type A ILRT as currently specified in the Technical Specifications. The one-time extension deviates from the guidelines contained in Regulatory Guide (RG) 1 .I 63.

The proposed change is consistent with the requirements of 10 CFR 50.36(~)(5)and 10 CFR 50, Appendix J, Option B,Section V.B.3.

Additionally, in accordance with 10 CFR 50, Appendix J, Option B,Section V.B, the proposed change to the LGS, Units 1 and 2 TS does not require a supporting request for an exemption to Option B of Appendix J, in accordance with 10 CFR 50.12, Specific exempt ions,

6.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

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

ATTACHMENT 1 Evaluation of Proposed Change Page 10 of 11 cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(~)(9).

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

PRECEDENT The proposed amendment incorporates into the LGS, Units 1 and 2 TS a change that is similar to changes (i-e., extend Type A ILRT from 10 to 15 years) approved by the NRC for Peach Bottom Atomic Power Station, Unit 3 (Reference lo), Three Mile Island, Unit 1 (Reference 1I ) , Vermont Yankee Nuclear Power Station (Reference 12), and Cooper Nuclear Station (Reference 13).

REFERENCES Nuclear Energy Institute (NEI) 94-01, Revision 0, Industry Guideline for Implementing Performance-BasedOption of 10 CFR 50, Appendix J, dated July 26, 1995 U. S. Nuclear Regulatory Commission letter dated January 24, 1997, Limerick Generating Station, Units 1 and 2 (TAC NOS. M96117 and M96118)

NUREG-I493, Performance-Based Containment Leak-Test Program, dated July 1995 U. S. Nuclear Regulatory Commission letter dated January 24, 2007, Limerick Generation Station, Units 1 and 2 - Relief Requests 13R-01 For Alignment of lnservice Inspection and Containment lnservice Inspection (TAC NOS. MD2727 AND MD 2728)

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

Letter from A. Pietrangelo (NEI) to NEI Administrative Points of Contact, Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leak Rate Test Surveillance Intervals, November 13, 2001.

Letter from A. Pietrangelo (NEI) to NEl Administrative Points of Contact, One-Time Extension of Containment Integrated Leak Rate Test Interval - Additional Information, November 30,2001.

Regulatory Guide 1.I 74, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, July 1998.

Electric Power Research Institute, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, 1009325, Revision 1, December 2005.

U. S. Nuclear Regulatory Commission letter dated October 4, 2001 Peach Bottom

)

Atomic Power Station, Unit 3 - Issuance of Amendment RE: Extension of the Containment Integrated Leak Rate Test (TAC NO. MB2094)

ATTACHMENT 1 Evaluation of Proposed Change Page 11 of 11 (11) U. S. Nuclear Regulatory Commission letter dated August 13, 2003, Three Mile Island Nuclear Station, Unit 1 (TMI-1), RE: Deferral of Containment Integrated Leakage Rate Test (TAC NO. MB6487)

(12) U. S. Nuclear Regulatory Commission letter dated August 31, 2005, Vermont Yankee Nuclear Power Station - Issuance of Amendment RE: One-Time Extension of Integrated Leak Rate Test Interval (TAC NO. MC4662)

(13) U. S. Nuclear Regulatory Commission letter dated October 3, 2006, Cooper Nuclear Station - Issuance of Amendment RE: Additional Extension of Appendix J, Type A, Integrated Leak Rate Test Interval (TAC NO. MC9732)

ATTACHMENT 2 MARKUP OF PROPOSED TECHNICAL SPECIFICATION PAGE CHANGE Revised TS Paaes 6-14c (Units 1 and 2)

TlVF C

~ AND ~~~ (Continued) r i m a r v C o n t a i m n t J eabge Rate Testina Proaram

( I 9.

A program s h a l l be established t o implement the leakage r a t e t e s t i n g of the containment as r e uired by 10 CFR 50.54 (01 and 10 CFR 50, Ap endix 3, 4

Option 8, as modi i e d by approved exemptions. .This program s all be i n accordance w i t h the guidelines contained i n Regulatory Guide 1.163 t:

ontainment Leakage Test program," dated September containment internal pressure for the design basis loss of i s 44.0 psig.

The maximum alldwable primary containment leakage rate, L,, a t Pa, s h a l l be 0.5% o f primary containment a i r weight per day.

Leakage rate acceptance c r i t e r i a are:

a. P r i m a r y Containment leakage rate. acceptance c r i t e r i o n i s l e s s than or equal to 1.0

k. During t h e f i r s t u n i t startup following t e s t i n g i n accordance w i t t h i s program, the leakage r a t e acceptance c r i t e r i a are l e s s than or equal t o 0.60 L, f o r t h e Type 8 and Type C t e s t s and less than o r equal t o 0.75 l, f o r Type A tests; be ' A i r lock t e s t i n g acceptance c r i t e r i a are:

1) .' Overall a i r l o c k leakage r a t e i s less than or equal t o 0.05 L, when tested a t g r e a t e r - t h a n or equal t o Pam

2) Seal leakage rate i s less than or equal to 5 s c f per hour when the gap between t h e door seals i s pressurized t o 10 psig.

The provisions o f S p e c i f i c a t i o n 4.0.2 do n o t apply t o the t e s t frequencies specified in the P r i m a r y Contai nment Leakage Rate Testing Program.

The revisions of Specification 4.0.3 are applicable t o t h e t e s t s described j n t i e Primary Containment Leakage Rate Testing Program.

I b .

h. Technical -1ftcatfons U S ) RasW Control Pr0ara.1~

This program provides a means for processing changes t o the Bases of these Technical Speci f icat4 ons.

a. Changes t o ' t h e Bases o f t h e TS s h a l l be made under appropriate administrative controls and reviews.

be Licensees may make changes t o Bases without p r i o r NRC approval I

provided the .changes do n o t r e q u i r e e i t h e r o f the following:

A change i n the TS incorporated i n t h e license; or I

A change t o the UFSAR or Bases t h a t requires NRC approval pursuant t o 10 CFR 50.59.

C. The Bases Control Program s h a l l contain provisions t o ensure t h a t the Bases are maintained c o n s i s t e n t w i t h the U f S A R e I

de Proposed changes t h a t meet t h e c r i t e r i a o f b e above shall be reviewed and approved by t h e NRC p r i o r t o implementation. Changes t o the Bases implemented without p r i o r NRC approval shall be provided t o the NRC 0n.a frequency consistent w i t h 10 CFR S0.71(e).

LIMERICK - UNIT 1 6- 1 4 ~ Amendment No. M, 162

INSERT A

,as modified by the following exception to NEI 94-01, Rev. 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J:

a. Section 9.2.3: The fist Type A test performed after May 15, 1998 shall be performed no later than May 15,2013.

A program s h a l l be established t o implement t h e leakage r a t e t e s t i n g o f t h e containment as r e u i r e d by 10 CFR 50.54 (01 and 10 CFR 50, Ap endix J ,

9 Option 8, as modi i e d by approved exemptions. This program s a l l be i n accordance w i t h the gui del ines contained i n Regul atory Guide 1 163 1.

tainment Leakage Test program, " dated September ntainment i n t e r n a l pressure f o r the design basis loss of The maximum allowable primary containment leakage rate, L, a t P, s h a l l be 0.5% of primary containment a i r weight per day.

Leakage r a t e acceptance c r i t e r i a are:

a. Primary Containment leakage r a t e acce tance c r i t e r i o n i s l e s s than o r equal t o L O Lh. P During the f i r s t un t startup f o l l o w i n g t e s t i n g i n accordance w i t t h i s program, the leqkage r a t e acceptance c r i t e r i a are l e s s than o r equal t o 0.60 L, f o r t h e Type B and Type C t e s t s and less than or equal t o 0.75 L, f o r Type A tests;

b. A i r lock t e s t i n g acceptance c r i t e r i a are:

1) Overa77. a i r l o c k leakage r a t e i s l e s s than o r equal t o 0.05 L, when tested a t greater than o r equal t o P.,

2) Seal leakage r a t e i s less than o r equal t o 5 s c f per hour when the gap between the door seals i s pressurized t o 10 psig.

The provisions o f Specification 4.0.2 do not apply t o the t e s t frequencies specified i n the P r i m a r y Containment Leakage Rate Testing Program.

The r o v i s i o n s o f Specification 4.0.3 are applicable t o the t e s t s described i n t l e Primary Containment leakage Rate Testing Program.

h. Technical Sp=ifications (TS) Bases Control Proarm This program provides a means for processing changes t o t h e 8ases o f these Technical Specifications.

a. Changes t o the Bases of the TS s h a l l be made under appropriate admi n i s t r a t i ve contro 1 s and reviews .

b. Licensees may make changes t o Bases without r i o r NRC approval provided t h e changes do not r e q u i r e e i t h e r o f the fol owing:

A change i n the TS incorporated i n the license; or A change t o the UFSAR or Bases that requires NRC approval pursuant t o 10 CFR I

50.59.

C. The Bases Control Program s h a l l contain provisions t o ensure t h a t the Bases are maintained consistent w i t h t h e UFSAR.

d. Proposed changes t h a t meet the c r i t e r i a o f b. above s h a l l be reviewed and approved by the NRC p r i o r t o implementation. Changes t o the Bases implemented without p r i o r NRC approval sha77 be provided t o t h e NRC on a frequency consistent w i t h 10 CFR 50.71Ce).

LIMERICK - UNIT 2 6-14~ Amendment No. 84, 124

INSERT B

,as modified by the following exception to NEI 94-01, Rev. 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J:

a. Section 9.2.3: The first Type A test performed after May 21, 1999 shall be performed no later than May 2 1,2014.

ATTACHMENT 3 RETYPED PAGE FOR TECHNICAL SPECIFICATION CHANGE Retvped TS Paaes 6-14c (Units 1 and 2)

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

g. Primary Containment Leakase Rate Testinq Proqram A program shall be established t o implement the leakage r a t e t e s t i n g of the containment as re uired by 10 CFR 50.54 ( 0 ) a n d 10 C F R 50, Ap endix J ,

?

Option 6, as modi ied by approved exemptions. This program s a l l be in accordance w i t h the gui del i nes contained in Regul atory Gui de 1.163 R

Performance-Based Containment Leakage Test program, dated September 1995, as modified by the following exception t o N E I 94-01, Rev. 0 ,

Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendi x J  :

a. Section 9.2.3: The f i r s t Type A t e s t performed a f t e r May 15, 1998 shall be performed no l a t e r t h a n May 1 5 , 2013.

The peak calculated containment internal pressure for the design basis loss o f coolant accident, Pa, i s 44.0 psig.

The maximum allowable primary containment leakage r a t e , La, a t Pa, shall be 0.5% of primary containment a i r weight per d a y .

Leakage r a t e acceptance c r i t e r i a are:

a. Primary Containment leakage r a t e acceptance c r i t e r i o n i s l e s s t h a n or equal t o 1.0 La. During the f i r s t u n i t startup following t e s t i n g in accordance with thi s program, the leakage rate acceptance c r i t e r i a are l e s s t h a n or equal t o 0.60 La for the Type B a n d Type C t e s t s a n d l e s s t h a n or equal t o 0.75 La for Type A t e s t s ;

b. Air lock t e s t i n g acceptance c r i t e r i a are:

1) Overall airlock leakage rate i s l e s s t h a n or equal t o 0.05 La when tested a t greater t h a n or equal t o P,.

2) Seal leakage rate i s l e s s t h a n or equal t o 5 scf per hour when the g a p between the door seals i s pressurized t o 10 psig.

The provisions of Specification 4 . 0 . 2 do n o t apply t o the t e s t frequencies specified in the Primary Containment Leakage Rate Testing Program.

The rovisions of Specification 4.0.3 are applicable t o the t e s t s described R

in t e Primary Containment Leakage Rate Testing Program.

h. Technical Specifications ( T S ) Bases Control Prosram This program provides a means for processing changes t o the Bases o f these Technical Specifications.

a. Changes t o the Bases of the TS shall be made under appropriate administrative controls a n d reviews.

b. Licensees may make changes t o Bases without prior NRC approval provided the changes d o n o t require e i t h e r o f the following:

A change i n the TS incorporated in the license; or A change t o the UFSAR or Bases t h a t requires NRC approval pursuant t o 10 CFR 50.59.

c. The Bases Control Program shall contain provisions t o ensure t h a t the Bases are maintained consistent with the UFSAR.

d. Proposed changes t h a t meet the c r i t e r i a of b . above shall be reviewed and approved by the NRC prior t o implementation. Changes t o the Bases implemented without prior NRC approval shall be provided t o the NRC on a frequency consistent with 10 CFR 50.71(e).

LIMERICK - UNIT 1 6- 1 4 ~ Amendment No. M, 4-62,

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) g* Primar,y Containment Leakaqe Rate T e s t i n q Proqram A program s h a l l be e s t a b l i s h e d t o implement t h e l e a k a g e r a t e t e s t i n g o f t h e c o n t a i n m e n t as r e u i r e d b y 10 CFR 50.54 (01 and 10 CFR 50, Ap e n d i x J ,

7 O p t i o n B, as modi i e d b y approved e x e m p t i o n s . T h i s p r o g r a m s a l l be i n accordance w i t h t h e g u i d e l i n e s c o n t a i n e d i n R e g u l a t o r y Guide 1.163 R

Performance-Based Containment Leakage T e s t program, d a t e d September 1995, as m o d i f i e d by t h e f o l l o w i n g e x c e p t i o n t o N E I 9 4 - 0 1 , Rev. 0, I n d u s t r y G u i d e l i n e f o r I m p l e m e n t i n g Performance-Based O p t i o n o f 10 CFR 50, Appendix 3 :

a. S e c t i o n 9.2.3: The f i r s t Type A t e s t p e r f o r m e d a f t e r May 21, 1999 s h a l l be p e r f o r m e d no l a t e r t h a n May 21, 2014.

1 The peak c a l c u l a t e d containment i n t e r n a l p r e s s u r e f o r t h e d e s i g n b a s i s loss o f c o o l a n t a c c i d e n t , Pa, i s 44.0 p s i g .

The maximum a l l o w a b l e p r i m a r y c o n t a i n m e n t l e a k a g e r a t e , La, a t Pa, s h a l l be 0.5% o f p r i m a r y c o n t a i n m e n t a i r w e i g h t p e r day.

Leakage r a t e acceptance c r i t e r i a a r e :

a. Primary Containment l e a k a g e r a t e acceptance c r i t e r i o n i s l e s s t h a n o r e q u a l t o 1.0 La. D u r i n g t h e f i r s t u n i t s t a r t u p f o l l o w i n g t e s t i n g i n accordance w i t h t h i s program, t h e 1 eakage r a t e acceptance c r i t e r i a a r e l e s s t h a n o r equal t o 0.60 La f o r t h e Type 6 and Type C t e s t s and l e s s t h a n o r equal t o 0.75 La f o r Type A t e s t s ;

b. A i r 1 ock t e s t i n g acceptance c r i t e r i a a r e :

1) O v e r a l l a i r l o c k l e a k a g e r a t e i s l e s s t h a n o r e q u a l t o 0.05 La when t e s t e d a t g r e a t e r t h a n o r e q u a l t o Pa.

2 ) Seal l e a k a g e r a t e i s l e s s t h a n o r equal t o 5 s c f p e r h o u r when t h e gap between t h e d o o r s e a l s i s p r e s s u r i z e d t o 10 p s i g .

The p r o v i s i o n s o f S p e c i f i c a t i o n 4.0.2 do n o t a p p l y t o t h e t e s t f r e q u e n c i e s s p e c i f i e d i n t h e Primary Containment Leakage Rate T e s t i n g Program.

The r o v i s i o n s o f S p e c i f i c a t i o n 4.0.3 a r e a p p l i c a b l e t o t h e t e s t s d e s c r i b e d R

i n t e Primary Containment Leakage Rate T e s t i n g Program.

h. T e c h n i c a l S p e c i f i c a t i o n s (TS) Bases C o n t r o l Proqram T h i s p r o g r a m p r o v i d e s a means f o r p r o c e s s i n g changes t o t h e Bases o f t h e s e Techni c a l Speci f ic a t i o n s .

a. Changes t o t h e Bases o f t h e TS s h a l l be made under a p p r o p r i a t e admi n i s t r a t i ve c o n t r o l s and r e v i e w s .

b. Licensees may make changes t o Bases w i t h o u t r i o r NRC approval p r o v i d e d Y

t h e changes do n o t r e q u i r e e i t h e r o f t h e f o l owing:

A change i n t h e TS i n c o r p o r a t e d i n t h e l i c e n s e ; o r A change t o t h e UFSAR o r Bases t h a t r e q u i r e s NRC a p p r o v a l p u r s u a n t t o 10 CFR 50.59.

C. The Bases C o n t r o l Program s h a l l c o n t a i n p r o v i s i o n s t o e n s u r e t h a t t h e Bases a r e m a i n t a i n e d c o n s i s t e n t w i t h t h e UFSAR.

d. Proposed changes t h a t meet t h e c r i t e r i a o f b . above s h a l l be r e v i e w e d and approved by t h e NRC p r i o r t o i m p l e m e n t a t i o n . Changes t o t h e Bases implemented w i t h o u t p r i o r NRC a p p r o v a l s h a l l be p r o v i d e d t o t h e NRC on a f r e q u e n c y c o n s i s t e n t w i t h 10 CFR 5 0 . 7 U e ) .

LIMERICK - UNIT 2 6-14~ Amendment No. %A, 424,

ATTACHMENT 4 Risk Assessment for LGS, Units 1 and 2 to Support ILRT (Type A) Interval Extension Request

I RM DOCUMENTATIONNO.

_ _ _ _ ~

LG-2006-LAR-0I REV: 0 PAGENO. 1 I STATION: LIMERICK UNIT(S) AFFECTED: 1 and 2 I

I TITLE: Risk Assessment for Limerick Unit Iand Unit 2 To Support I L R l (Type A) Interval Extension Request I

SUMMARY

(Include UREs incorporated): The purpose of this analysis is to provide an assessment of the risk associated with implementing a one-time extension of the Limerick Unit 1 and Limerick Unit 2 containment Type A integrated leak rate test (ILRT) interval from 10 years to 15 years.

Internal RM Documentation Electronic Calculation Data Fifes:

(Program Name, Version, File Name extension/size/date/hour/min)

Prepared by: Donald E. Vanover I 2 / IZ//.?/dG Print Sign Date I Reviewed by: Robert J. WolfQanqj Method of Review:

Print

[Xj Detailed [ f Alternate This RM documentation supersedes: NIA in its entirety.

Approved by: Greqow A. Krueqer Print Date/

II External RM Documentation Reviewed by: / /

Pnnt Sign Date Approved by: NIA -I I Print Sign Date 1 Do any ASSUMPTIONS /ENGlNEER/NGJUDGEMEN73 require later v e ~ c a f f o n ? [ ] Yes [ x ] NO

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval TABLE OF CONTENTS Section Paae 1.0 PURPOSE OF ANALYSIS ....................................................................................... 3 1.1 Purpose............................................................................................................ 3 1.2 Background...................................................................................................... 3 1.3 Criteria.............................................................................................................. 5 2.0 METHODOLOGY ..................................................................................................... 7 3.0 GROUND RULES................................................................................................... 13 4.0 INPUTS................................................................................................................... 15 4.1 General Resources Available ........................................................................ 15 4.2 Plant-Specific Inputs ..................................................................................... 23 4.3 Impact of Extension on Detection of Component Failures That Lead to Leakage (Small and Large) ........................................................................... 32 4.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage ......................................................................................................... 34 5.0 RESULTS ............................................................................................................... 40 5.1 Step I- Quantify the Base-Line Risk in Terms of Frequency per Reactor Year ............................................................................................................... 42 SCENARIO TYPE..................................................................................................... 43 5.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose) per Reactor Year .................................................................................................. 48 5.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval From IO-to-15 Years ......................................................................................................... 51 5.4 Step 4 - Determine the Change in Risk in Terms of Large Early Release Frequency...................................................................................................... 55 5.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability ...................................................................................................... 55 5.6 Summary of Results ...................................................................................... 56 6.0 SENSITIVITIES ...................................................................................................... 58 6.1 Sensitivity to Corrosion Impact Assumptions ................................................ 58 6.2 EPRl Expert Elicitation Sensitivity ................................................................. 60

7.0 CONCLUSION

S ..................................................................................................... 63

8.0 REFERENCES

....................................................................................................... 65 2 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval 1.O PURPOSE OF ANALYSIS 1.I Purpose The purpose of this analysis is to provide an assessment of the risk associated with implementing a one-time extension of the Limerick Generating Station Units 1 and 2 containment Type A integrated leak rate test (ILRT) interval from ten years to fifteen years.

The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages. The risk assessment follows the guidelines from NEI 94-01 [I], the methodology used in EPRl TR-I04285 [2], the NEI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals [3, 211, the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a request for a plants licensing basis as outlined in Regulatory Guide (RG) 1.174 [4], and the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval

[19]. The format of this document is consistent with the intent of the Risk Impact Assessment Template for evaluating extended integrated leak rate testing intervals provided in the December 2005 EPRl final report [22].

1.2 Background Revisions to IOCFRSO, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing requirements from three-in-ten years to at least once per ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage was less than normal containment leakage of 1.OL, (allowable leakage).

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and was established in 1995 during development of the performance-based 3 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRTInterval Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493 [5],

Performance-Based Containment Leak Test Program, September 1995, provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertook a similar study. The results of that study are documented in Electric Power Research Institute (EPRI) Research Project Report TR-104285.

The NRC report, Performance Based Leak Test Program, NUREG-1493 [5], analyzed the effects of containment leakage on the health and safety of the public and the benefits realized from the containment leak rate testing. In that analysis, it was determined for a comparable BWR plant, that increasing the containment leak rate from the nominal 0.5 percent per day to 5 percent per day leads to a barely perceptible increase in total population exposure, and increasing the leak rate to 50 percent per day increases the total population exposure by less than 1 percent. Consequently, extending the ILRT interval should not lead to any substantial increase in risk. The current analysis is being performed to confirm these conclusions based on Limerick specific models and available data.

Earlier ILRT frequency extension submittals have used the EPRl TR-104285 methodology to perform the risk assessment. In November and December 2001, NEI issued enhanced guidance (hereafter referred to as the NEI Interim Guidance) that builds on the TR-104285 methodology and intended to provide for more consistent submittals [3,21]. The NEI Interim Guidance was developed for NEI by EPRl using personnel who also developed the TR-104285 methodology. This ILRT interval extension risk assessment for Limerick employs the NEI Interim Guidance methodology.

It should be noted that, in addition to ILRT tests, containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel

~

4 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Code (ASME Code),Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants.

Furthermore, NRC regulations 10 CFR 50m55a(b)(2)(ix)(E),require licensees to conduct visual inspections of the accessible areas of the interior of the containment 3 times every 10 years. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency. Type C tests are also not affected by the Type A test frequency change.

1.3 Criteria The acceptance guidelines in RG 1.I74 are used to assess the acceptability of this one-time extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in core damage frequency (CDF) less than per reactor year and increases in large early release frequency (LERF) less than per reactor year.

Since the Type A test does not impact CDF for Limerick, the relevant criterion is the change in LERF. RG 1.174 also defines small changes in LERF as below lom6 per reactor year provided that the total from all contributors (including external events) can be reasonably shown to be less than 10-5 per reactor year. RG 1.I 74 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the conditional containment failure probability (CCFP) that helps to ensure that the defense-in-depth philosophy is maintained is also calculated.

In addition, the total annual risk (person remlyr population dose) is examined to demonstrate the relative change in this parameter based on the precedent set by previous 5 PO4670060049-2706

Risk Impact Assessment sf Extending Limerick Units 1 and 2 ILRT Interval submittals for ILRT extensions [6, 20, 231. (No criteria have been established for this parameter change.)

6 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval 2.0 METHODOLOGY A simplified bounding analysis approach consistent with the EPRI approach is used for evaluating the change in risk associated with increasing the test interval to fifteen years

[22]. The approach is consistent with that presented in NEI Interim Guidance [3, 211, EPRl TR-I04285 [2], NUREG-1493 [5] and the Calvert Cliffs liner corrosion analysis [19]. The analysis uses results from a Level 2 analysis of core damage scenarios from the current Limerick PRA model and subsequent containment response resulting in various fission product release categories (including no or negligible release). The Limerick Unit 1 model is used in this analysis, but since there is no significant difference between the two Unit models, this risk assessment is applicable to Limerick Units 1 and 2. Confidence in the validity of this assumption can be obtained by examining the comparison of the Unit 1 and Unit 2 Level 1 model results shown in Table 2-1 below, and the Level 2 model results shown in Table 2-2 below.

The six general steps of this assessment are as follows:

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

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

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

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

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

6. Evaluate the sensitivity of the results to assumptions in the liner corrosion analysis and to the fractional contribution of increased large isolation failures (due to liner breach) to LERF.

7 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRTInterval This approach is based on the information and approaches contained in the previously mentioned studies. Furthermore, Consistent with the other industry containment leak risk assessments, the Limerick assessment uses population dose as one of the risk measures. The other risk measures used in the Limerick assessment are LERF and the conditional containment failure probability (CCFP) to demonstrate that the acceptance guidelines from RG I. 174 are met.

This evaluation for Limerick uses ground rules and methods to calculate changes in risk metrics that are similar to those used in the EPRI approach.

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8 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Table 2-1

SUMMARY

OF THE LIMERICK CORE DAMAGE FREQUENCY BY ACCIDENT SEQUENCE SUBCLASS Accident Class 1 LGSU~~~I Freauencv LGS Unit 2 Frequency Designator Subclass Definition (per year) (per year)

Class' I Accident sequences involving loss of inventory makeup in which the reactor pressure remains high.

1 1.19E-06 1.19E-06 B Accident sequences involving a loss of offsite power and loss of coolant 1.17E-06 1.17E-06 inventory makeup.

1 l c Accident sequences involving a loss of coolant inventory induced by an ATWS sequence with containment intact.

2.56E-08 2.56E-08 D Accident sequences involving a loss of coolant inventory makeup in which reactor pressure has been successfully reduced to 200 psi.

1 9.68E-08 9.68E-08 1

E Accident sequences involving loss of inventory makeup in which the reactor 3.39E-09 3.39E-09 pressure remains high and DC power is unavailable.

Class II A Accident sequences involving a loss of containment heat removal with the RPV initially intact; core damage; core damage induced post containment 1.02E-07 1.01E-07 failure.

F Class IIA and IIL except that the vent operates as designed; loss of makeup occurs at some time following vent initiation. Suppressionpool saturated but 6.78%-07 6.78E-07 intact.

L Accident sequences involving a loss of containment heat removalwith the RPV breached but no initial core damage; core damage induced post containment failure. (Note that this is a new category for the 2004 update, and is grouped 3.61E-08 3.61 E-08 with Class IIA for transfer to the Level 2 model for consistency with previous treatment in the Level 2 model evaluation.)

9 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval Table 2-1

SUMMARY

OF THE LIMERICK CORE DAMAGE FREQUENCY BY ACCIDENT SEQUENCE SUBCLASS Accident LGS Unit 1 LGS Unit 2 Class Frequency Frequency Designator Subclass Definition (per year) (per year)

Class I l l A Accident sequences leading to core damage conditions initiated by vessel (LOCA) rupture where the containment integrity is not breached in the initial time phase 9.50E-09 9.50E-09 of the accident.

6 Accident sequences initiated or resulting in small or medium LOCAs for which 3.08E-08 3.08E-08 the reactor cannot be depressurized prior to core damage occurring.

C Accident sequences initiated or resulting in medium or large LOCAs for which 4.55E-08 4.55E-08 the reactor is a low pressure and no effective injection is available.

D Accident sequences which are initiated by a LOCA or RPV failure and for which the vapor suppression system is inadequate, challenging the 1.99E-08 1.99E-08 containment integrity with subsequent failure of makeup systems.

Class IV A Accident sequences involving failure of adequate shutdown reactivity with the 2.22E-07 2.22E-07 (ATWS) RPV initially intact; core damage induced post containment failure.

L Accident sequences involving failure of adequate shutdown reactivity with the RPV initially breached; core damage induced post containment failure. (Note that this is a new category for the 2004 update, and is grouped with Class IIA for 4.29E-08 4.29E-08 transfer to the Level 2 model for consistency with previous treatment in the Level 2 model evaluation.)

Class V --- Unisolated LOCA outside containment. 3.24E-08 3.24E-08 10 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Table 2-2

SUMMARY

OF THE LIMERICK LEVEL 2 MODEL RESULTS BY RELEASE CATEGORY LGS Unit 1 Frequency (per LGS Unit 2 Frequency (per Release Category()

Year) Year)

M/I 1.39E-06 1.39E-06 L/L 2.28E-07 2.28E-07 M/L 1.98E-07 1.98E-07 LLE 1.46E-07 1.46E-07 L/I 1.09E-07 1.09E-07 I I I

MIE 9.27E-08 9.27E-08 LLL 8.92E-08 8.92E-08 H/E 6.80E-08 6.80E-08 L/E 3.09E-08 3.09E-08 H/L 1-18E-08 1.18E-08 H/I 8.53E-09 8.53E-09 LLI I.42E-09 1.43E-09 Totaf2): 2.37E-06 2.37E-06 (I) Refer to Table 2-3 for the release category classification scheme used in the Limerick Level 2 analysis.

(2) The difference between this value and the total CDF value of 3.70E-06 is assigned to the Containment Intact (OK) category.

11 PO4670060049-2706

Risk ImDact Assessment o f Extendina Limerick Units I and 2 ILRT Interval Table 2-3 RELEASE SEVERITY AND TIMING CLASSIFICATION SCHEME (SEVERITY, TIMING)

Classification Category Cs Iodide YOin Classification Category Release Time of Release()

1 Greater than I 0 Late (L) I Greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

/I Moderate (M) 1 to10 Intermediate (I) 1 6 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.1 to 1 Early (E) Less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 1 Low-low (LL) I Less than 0.1 Relative to (I) the declaration of a General Emergency.

~ ~~~~~

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Risk Impact Assessment o f Extendina Limerick Units 1 and 2 ILRTIntewal 3.0 GROUND RULES The following ground rules are used in the analysis:

The Limerick Level I and Level 2 internal events PRA models provide representative results.

It is appropriate to use the Limerick internal events PRA model as a gauge to effectively describe the risk change attributable to the ILRT extension. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose) will not substantially differ if fire and seismic events were to be included in the calculations.

Dose results for the containment failures modeled in the PRA can be characterized by information provided in NUREG/CR-4551 [9]. They are estimated by scaling the NUREG/CR-4551 results by population differences for Limerick compared to the NUREG/CR-4551 reference plant.

Accident classes describing radionuclide release end states are defined consistent with EPRl methodology [2] and are summarized in Section 4.2.

The representative containment leakage for Class 1 sequences is IL,. Class 3 accounts for increased leakage due to Type A inspection failures.

The representative containment leakage for Class 3a sequences is IOL,, based on the previously approved methodology petformed for Indian Point Unit 3 [6, 71.

The representative containment leakage for Class 3b sequences is 35La, based on the previously approved methodology [6, 71.

The Class 3b can be very conservatively categorized as LERF based on the previously approved methodology [6, 71. The Class 3b category increase is used as a surrogate for LERF in this application even though the releases associated with a 35La release would not necessarily be consistent with a Large release for Limerick.

The impact on population doses from containment bypass scenarios is not altered by the proposed ILRT extension, but is accounted for in the EPRl methodology as 13 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRTInterval a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes on the conclusions from this analysis will result from this separate categorization.

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

The use of estimated 2010 population data is adequate for this analysis. Scaling the year 2010 population data to the date of the next ILRT test if extended beyond the current due date would not significantly impact the quantitative results, nor would it change the conclusions.

An evaluation of the risk impact of the ILRT on shutdown risk is addressed using the generic results from EPRl TR-105189 [8].

14 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRTInterval 4.0 INPUTS This section summarizes the general resources available as input (Section 4.1) and the plant specific resources required (Section 4.2).

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

1. NUREG/CR-3539 [lo]

2. NUREG/CR-4220 [I I]

3. NUREG-I273 [I21

4. NUREG/CR-4330 [I31

5. EPRl TR-105189 [8]

6. NUREG-I493 [5]

7. EPRl TR-104285 [2]

8. NUREG-I 150 [I41 and NUREG/CR-4551 [9]

9. NEI Interim Guidance [3, 211

10. Calvert Cliffs liner corrosion analysis [ I 91

11. EPRI I009325 [22]

The first study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to NUREGlCR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different 15 PO4670060049-2706

Risk Imivact Assessment o f Extendina Limerick Units I and 2 ILRT Interval containment leakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRCs cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh study is an EPRl study of the impact of extending ILRT and LLRT test intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the bases for the consequence analysis of the ILRT interval extension for Limerick. The ninth study includes the NEI recommended methodology for evaluating the risk associated with obtaining a one-time extension of the ILRT interval. The tenth study addresses the impact of age-related degradation of the containment liners on ILRT evaluations. Finally, the last study complements the previous EPRl report [2],

integrates the NEI interim guidance, and provides the results of an expert elicitation process to determine the relationship between pre-existing containment leakage probability and magnitude.

NUREGER-3539 [lo]

Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leak rates on public risk in NUREGICR-3539. This study uses information from WASH-I 400 [ I 51 as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

NUREGKR-4220 [I 11 NUREWCR-4220 assessed the large containment leak probability to be in the range of 1E-3 to 1E-2, with 5E-3 identified as the point estimate based on 4 events in 740 reactor years and conservatively assuming a one-year duration for each event. It should be noted that all of the 4 identified large leakage events were PWR events, and the assumption of a one-year duration is not applicable to an inerted containment such as Limerick.

NUREGER-4220 identifies inerted BWRs as having significantly improved potential for leakage detection because of the requirement to remain inerted during power operation.

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Risk ImDact Assessment o f Extendim Limerick Units 1 and 2 ILRT Interval This calculation presented in NUREG/CR-4220 is called an upper bound estimate for BWRs (presumably meaning inerted BWR containment designs).

NUREG-1273 [I21 A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREGKR-4220 database. This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected. In addition, this study noted that local leak rate tests can detect essentially all potential degradations of the containment isolation system.

NUREG/CR-4330 [I31 NUREG/CR-4330 is a study that examined the risk impacts associated with increasing the allowable containment leakage rates. The details of this report have no direct impact on the modeling approach of the ILRT test interval extension, as NUREWCR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals. However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies:

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

EPRl TR-105189 [81 The EPRl study TR-I05189 is useful to the ILRT test interval extension risk assessment because this EPRl study provides insight regarding the impact of containment testing on shutdown risk. This study performed a quantitative evaluation (using the EPRl ORAM software) for two reference plants (a 6WR-4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk.

The result of the study concluded that a small but measurable safety benefit is realized from extending the test intervals. For the BWR, the benefit from extending the ILRT 17 PO4670060049-2706

Risk ImDact Assessment of Extendina Limerick Units 1 and 2 ILRT Interval frequency from 3 per 10 years to 1 per 10 years was calculated to be a reduction of approximately 1E-7lyr in the shutdown core damage frequency. This risk reduction is due to the following issues:

Reduced opportunity for draindown events Reduced time spent in configurations with impaired mitigating systems The study identified 7 shutdown incidents (out of 463 reviewed) that were caused by ILRT or LLRT activities. Two of the 7 incidents were RCS draindown events caused by ILRT/LLRT activities, and the other 5 were events involving loss of RHR and/or SDC due to lLRT/LLRT activities. This information was used in the EPRl study to estimate the safety benefit from reductions in testing frequencies. This represents a valuable insight into the improvement in the safety due to extending the ILRT test interval.

NUREG-1493 151 NUREG-I493 is the NRCs cost-benefit analysis for proposed alternatives to reduce containment leakage testing intervals and/or relax allowable leakage rates. The NRC conclusions are consistent with other similar containment leakage risk studies:

Reduction in ILRT frequency from 3 per I 0 years to 1 per 20 years results in an imperceptible increase in risk.

0 Increasing containment leak rates several orders of magnitude over the design basis would minimally impact (0.2 - 1.0%) population risk.

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

EPRl TR-104285 121 Extending the risk assessment impact beyond shutdown (the earlier EPRl TR-105189 study), the EPRl TR-104285 study is a quantitative evaluation of the impact of extending Integrated Leak Rate Test (ILRT) and (Local Leak Rate Test) LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with 18 PO4670060049-2706

Risk Imlvact Assessment o f Extendina Limerick Units 1 and 2 ILRT Interval NUREG-1I 5 0 Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-I493 in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

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

1 Containment intact and isolated a

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

3. Type A (ILRT) related containment isolation failures

4. Type B (LLRT) related containment isolation failures

5. Type C (LLRT) related containment isolation failures

6. Other penetration related containment isolation failures

7. Containment failure due to core damage accident phenomena

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

These study results show that the proposed CLRT [containment leak rate tests] frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms. For example, for the PWR analyzed, the change is about 0.02 person-rem per year...

Release Cateqory Definitions Table 4.1-1 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI/NEI methodology [2]. These containment failure classifications are used in this analysis to determine the risk impact of extending the Containment Type A test interval as described in Section 5 of this report.

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Risk Impact Assessment of Extendinn Limerick Units I and 2 ILRT Interval Table 4.1-1 EPRVNEI CONTAINMENT FAILURE CLASSIFICATIONS [2]

Class Description 1 Containment remains intact including accident sequences that do not lead to containment failure in the long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant 2 Containment isolation failures (as reported in the IPEs) include those accidents in which there is a failure to isolate the containment.

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

4 Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 3 isolation failures, but is applicable to sequences involving Type B tests and their potential failures. These are the Type B-tested components that have isolated but exhibit excessive leakage.

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

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

7 Accidents involving containment failure induced by severe accident phenomena.

Changes in Appendix J testing requirements do not impact these accidents.

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

NUREG-1150 [I41 and NUREG/CR 4551 [91 NUREG-1150 and the technical basis, NUREG/CR-4551, provide an ex-plant consequence analysis for a spectrum of accidents including a severe accident with the containment remaining intact (i.e., Technical Specification leakage). This ex-plant consequence analysis is calculated for the 50-mile radial area surrounding Peach Bottom. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the Limerick 20 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent Limerick if the population is scaled to represent Limerick. (The meteorology and site differences other than population are assumed not to play a significant role in this evaluation.)

NEI Interim Guidance 13, 211 NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions of Containment Integrated Leakage Rate Test Surveillance Intervals [3] has been developed to provide utilities with revised guidance regarding licensing submittals.

Additional information from NEI on the Interim Guidance was supplied in Reference PI1

• A nine step process is defined which includes changes in the following areas of the previous EPRI guidance:

Impact of extending surveillance intervals on dose Method used to calculate the frequencies of leakages detectable only by ILRTs 0 Provisions for using NUREG-I 150 dose calculations to support the population dose determination.

The guidance provided in this document builds on the EPRI risk impact assessment methodology [2] and the NRC performance-based containment leakage test program

[5], and considers approaches utilized in various submittals, including Indian Point 3

[6,7] (and associated NRC SER) and Crystal River [20].

The approach included in this guidance document is used in the Limerick assessment to determine the estimated increase in risk associated with the ILRT extension. This document includes the bases for the values assigned in determining the probability of leakage for the EPRI Class 3a and 3b scenarios in this analysis as described in Section 5.

21 PO4670060049-2706

Risk Imaact Assessment of Extendina Limerick Units 1 and 2 ILRTIntewal Calvert Cliffs Response to Request for Additional Information Concerning the License Amendment for a One-Time Integrated Leakage Rate Test Extension [I 91.

This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting liner corrosion, and the corresponding change in risk. The methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms were factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner.

EPRl Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals 1221 This report presents a risk impact assessment for extending integrated leak rate test (ILRT) surveillance intervals to 15 years and is consistent in nature with the NEI interim guidance. This risk impact assessment complements the previous EPRl report, TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals.

The earlier report considered changes to local leak rate testing intervals as well as changes to ILRT testing intervals. The original risk impact assessment considers the change in risk based on population dose, whereas the revision considers dose as well as large early release frequency (LERF) and conditional containment failure probability (CCFP). This report deals with changes to ILRT testing intervals and is intended to provide bases for supporting changes to industry (NEI) and regulatory (NRC) guidance on ILRT surveillance intervals.

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

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Risk ImDact Assessment of Extending Limerick Units I and 2 ILRT Interval 4.2 Plant-Specific Inputs The Limerick specific information used to perform this ILRT interval extension risk assessment includes the following:

Level 1 Model results [I61 Level 2 Model results [I61 Population within a 50-mile radius [I71 0 ILRT results to demonstrate adequacy of the administrative and hardware issues (I)

Limerick Internal Events Level 1 PRA Model The current Level 1 PRA model is an event tree / linked fault tree model characteristic of the as-built, as-operated plant. The total internal events core damage frequency (CDF) used in this analysis is 3.70E-O6/yr for both Unit 1 and Unit 2 1161.

Limerick Internal Events Level 2 PRA Model The Level 2 Model that is used for Limerick was developed to calculate the LERF contribution as well as the other release categories evaluated in the model. Table 4.2-1 summarizes the pertinent Limerick results in terms of release category. The total Large Early Release Frequency (LERF) which corresponds to the H/E release category in Table 4.2-1 was found to be 6.80E-8/yr. The total release frequency is 2.37E-O6/yr. With a total CDF of 3.70E-O6/yr, this corresponds to an "OK" release limited to normal leakage of 1.33E-6/yr [I 61.

(I) Limerick ST-1-060-490-1/2 identifies that the 95% Upper Confidence Level (UCL) leak rate shall be less than 0.75La (0.375% wtlday). The following acceptable leak rates were identified [18]:

Unit 1 Unit 2

(%wt/day) (%wtlday) 1984 0.1642 1989 0.2310 1987 0.1468 1993 0.1836 1990 0.2870 1999 0.3272 1998 0.3070 Since the two most recent Type A tests at Limerick Unit Iand Unit 2 have been successful, the current Type A test interval requirement is 10 years.

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Risk Imx7act Assessment of Extending Limerick Units I and 2 ILRT Interval Table 4.2-1 Limerick Level 2 PRA Model Release Categories and Frequencies l Category I Frequencylyr H/E - High Early (LERF) I 6.80E-08 I I M/E - Medium Early I 9.27E-08 1I

/I L/E - Low Early I 3.09E-08 I 1I I I

~ ~~~

LL/E - Low Low Early 1.46E-07 I HA - High Intermediate I 8.53E-09 1I l M/I - Medium Intermediate 1 1.39E-06 I 1I L/I - Low Intermediate I 1.09E-07 l 1 LL/I - Low Low Intermediate I 1.42E-09 R I H/L - High Late I 1.18E-08 l I M/L - Medium Late I 1.98E-07 l I L/L - Low Late I 2.28E-07 I LL/L - Low Low Late 8.92E-08 H Total Release Frequency I 2.37E-06 l Core Damage Frequency I 3.70E-06 I Population Dose Calculations The population dose is calculated by using data provided in NUREG/CR-4551 and adjusting the results for Limerick. Each accident sequence was associated with an applicable collapsed Accident Progression Bin (APB) from NUREG/CR-4551. The collapsed APBs are characterized by 5 attributes related to the accident progression.

Unique combinations of the 5 attributes result in a set of 10 bins that are relevant to the analysis. Information from the Limerick PRA Containment Event Trees (CETs) was used to classify each of the Level 2 sequences using these attributes. The definitions of the I 0 collapsed APBs are provided in NUREG/CR-4551 and are reproduced in Table 4.2-2 for 24 PO4670060049-2706

Risk Imlvact Assessment o f Extendina Limerick Units 1 and 2 ILRT Interval references purposes. Table 4.2-3 summarizes the calculated population dose associated with each APB from NUREG/CR-4551.

Table 4.2-2 Collapsed Accident ProGression Bin (APB) Descriptions r91 Collapsed Description APB Number 1 CD, VB, Early CF, WW Failure, RPV Pressure 200 psi at VB Core damage occurs followed by vessel breach. The containment fails early in the wetwell (i.e., either before core damage, during core damage, or at vessel breach) and the RPV pressure is greater than 200 psi at the time of vessel breach (this means Direct Containment Heating (DCH) is possible).

2 CD, VB, Early CF, WW Failure, RPV Pressure < 200 psi at VB Core Damage occurs followed by vessel breach. The containment fails early in the wetwell (i.e., either before core damage, during core damage, or at vessel breach) and the RPV pressure is less than 200 psi at the time of vessel breach (this means DCH is not possible).

3 CD, VB, Early CF, DW Failure, RPV Pressure > 200 psi at VB Core damage occurs followed by vessel breach. The containment fails early in the drywell (Le., either before core damage, during core damage, or at vessel breach) and the RPV pressure is greater than 200 psi at the time of vessel breach (this means DCH is possible).

4 CD, VB, Early CF, DW Failure, RPV Pressure c 200 psi at VB Core Damage occurs followed by vessel breach. The containment fails early in the drywell (i.e., either before core damage, during core damage, or at vessel breach) and the RPV pressure is less than 200 psi at the time of vessel breach (this means DCH is not possible).

5 CD, VB, Late CF, WW Failure, N/A Core Damage occurs followed by vessel breach. The containment fails late in the wetwell (i.e., after vessel breach during Molten Core-Concrete Interaction (MCCI)) and the RPV pressure is not important since, even if DCH occurred, it did not fail containment at the time it occurred.

6 CD, VB, Late CF, DW Failure, N/A Core Damage occurs followed by vessel breach. The containment fails late in the drywell (i.e., after vessel breach during MCCI) and the RPV pressure is not important since, even if DCH occurred, it did not fail containment at the time it occurred.

~

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Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval Table 4.2-2 Collapsed Accident Progression Bin (APB) Descriptions r91 Collapsed Description APB Number 7 CD, VB, No CF, Vent, NIA Core Damage occurs followed by vessel breach. The containment never structurally fails, but is vented sometime during the accident progression. RPV pressure is not important (characteristic 5 is NIA) since, even if it occurred, DCH does not significantly affect the source term as the containment does not fail and the vent limits its effect.

8 CD, VB, No CF, NIA, NIA Core damage occurs followed by vessel breach. The containment never fails structurally (characteristic 4 is N/A) and is not vented. RPV pressure is not important (characteristic 5 is NIA) since, even if it occurred, DCH did not fail containment. Some nominal leakage from the containment exists and is accounted for in the analysis so that while the risk will be small it is not completely negligible.

9 CD, No VB, NIA, NIA, NIA Core damage occurs but is arrested in time to prevent vessel breach. There are no releases associated with vessel breach or MCCI. It must be remembered, however, that the containment can fail due to overpressure or venting even if vessel breach is averted. Thus, the potential exists for some of the in-vessel releases to be released to the environment.

10 No CD, NIA, NIA, NIA, NIA Core damage did not occur. No in-vessel or ex-vessel release occurs. The containment may fail on overpressure or be vented. The RPV may be at high or low pressure depending on the progression characteristics. The risk associated with this bin is negligible.

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Risk ImDact Assessment of Extending Limerick Units I and 2 ILRT Interval Table 4.2-3 Calculation of P6APS Population Dose Risk at 50 Miles Collapsed Fractional APB NUREG/CR-4551 NUREGlCR-4551 NUREG/CR-4551 Bin # Contributions to Population Dose Collapsed Bin Population Dose Risk (MFCR) (I) Risk at 50 miles Frequencies at 50 miles (From a total of (per year) (3) (Person-rem) (4) 7.9 person-rem/yr, mean) (2) 1 0.021 0.1659 9.55E-08 1.74E+06 2 0.0066 0.05214 I 4.77E-08 3 0.556 4.3924 1.48E-06 4 0.226 1.7854 7.94E-07 5 0.0022 0.01738 I.30E-08 6 0.059 0.4661 2.04E-07 7 0.1 18 0.9322 4.77E-07 0.0005 0.00395 7.99E-07 0.01 0.079 3.86E-07 0 0 4.34E-08 (I)

Mean Fractional Contribution to Risk from Table 5.2-3 of NUREG/CR-4551 The total population dose risk at 50 miles from internal events in person-rem is provided in Table 5.1-1 of NUREG/CR-4551. The contribution for a given APB is the product of the total PDR5O and the fractional APB contribution.

(3)

NUREGKR-4551 provides the conditional probabilities of the collapsed APBs in Figure 2.5-6.

These conditional probabilities are multiplied by the total internal CDF to calculate the collapsed APB frequency.

(4)

Obtained from dividing the population dose risk shown in the third column of this table by the collapsed bin frequency shown in the fourth column of this table.

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Risk Impact Assessment of Extendinn Limerick Units 1 and 2 ILRT Interval Population Estimate Methodolow The person-rem results in Table 4.2-3 can be used as an approximation of the dose for Limerick if it is corrected for the population surrounding Limerick. The total population within a 50-mile radius of Limerick is projected to be 8.05E+06 by the year 2010 [ I 71. This value is slightly less than the projected value of 8.53E+06 from the Limerick UFSAR [24]

since it factors in more recent actual census data from 1990 and 2000 for the projected growth estimates compared to the earlier population data utilized in the UFSAR. The use of the 2010 estimate is judged to be sufficient to perform this assessment. Scaling the year 2010 population data to the date of the next ILRT test if extended beyond the current due date would not significantly impact the quantitative results, nor would it change the conclusions.

This population value is compared to the population value that is provided in NUREGICR-4551 in order to get a Population Dose Factor that can be applied to the APBs to get dose estimates for Limerick.

Total Limerick P ~ p ~ l a t i ~ n =s 8.05E+06 o~il~~

Peach Bottom Population from NUREG/CR-4551= 3.02E+06 Population Dose Factor = 8.05E+06 / 3.02E+06 = 2.67 The difference in the doses at 50 miles is assumed to be in direct proportion to the difference in the population within 50 miles of each site. This does not take into account differences in meteorology data, detailed environmental factors or detailed differences in containment designs, but does provide a first-order approximation for Limerick of the population doses associated with each of the release categories from NUREG/CR-4551.

This is considered adequate since the conclusions from this analysis will not be substantially affected by the actual dose values that are used.

Table 4.2-4 shows the results of applying the population dose factor to the NUREG/CR-4551 population dose results at 50 miles to obtain the adjusted population dose at 50 miles for Limerick.

28 PO4670060049-2706

Risk ImDact Assessment OfExtendinPLimerick Units I and 2 ILRT Interval Table 4.2-4 Calculation of Limerick Population Dose Risk at 50 Miles 1

Accident NUREGICR-4551 Bin Multiplier Limerick Adjusted Progression Population Dose used to obtain Population Dose at Bin # at 50 miles 1 Limerick 50 miles (Person-rem1 Powlation Dose (Person-rem1 1 I 1.74E+O6 2.67 I 4.65E+06 1I 2 1 1.09E+06 2.67 I 2.91E+06 I 3 I 2.97E+06 2.67 7.93E+06 4 I 2.25E+06 2.67 6.01E+06 5 I.34E+06 2.67 3.58Et.06 6 2.28E+06 2.67 6.09E+06 7 I 1.95E+06 2.67 5.21E+06 8 I 4.94E+03 2.67 I.32E+04 9 2.05E+05 2.67 5.47E+05 10 0 2.67 O.OOE+OO Application of Limerick PRA Model Results to NUREWCR-4551 Level 3 Output A major factor related to the use of NUREG/CR-4551 in this evaluation is that the results of the Limerick PRA Level 2 model are not defined in the same terms as reported in NUREWCR-4551. In order to use the Level 3 model presented in that document, it was necessary to apply the Limerick PRA Level 2 model results into a format which allowed for the scaling of the Level 3 results based on current Level 2 output. Finally, as mentioned above, the Level 3 results were modified to reflect the difference in the site demographics that exist between the two sites. This subsection provides a description of the process used to apply the Limerick PRA Level 2 model results into a form that can be used to generate Level 3 results using the NUREG/CR-4551 documentation.

The basic process that was pursued to obtain Level 3 results based on the Limerick Level 2 model and NUREG/CR-4551 was to define a useful relationship between the Level 2 and Level 3 results. Consequently, each non-zero sequence of the Limerick Level 2 model was reviewed and assigned into one of the collapsed Accident Progression Bins (APBs) from NUREG/CR-4551. The Level 2 model contains a

~

29 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval significantly larger amount of information about the accident sequences than what is used in the collapsed APBs in NUREG/CR-4551 and this assignment process required simplification of accident progression information and assumptions related to categorizations of certain items. The relevant assumptions used for these assignments are shown in Table 4.2-5. Other nodes are included in the Limerick Level 2 model, but these were not utilized (or did not contribute) to the APE3 assignment performed here for the ILRT assessment.

Table 4.2-5 Limerick Level 2 Model Nodal Assumptions for Application to the NUREG/CR-4551 Accident Progression Bins Limerick Assumption Containment Event Tree Node IS - Containment If the containment is not isolated, it is assumed that it will be open for the Isolation equivalent of an un-scrubbed release as soon as the vessel is breached. No depressurization is asked prior to this node; it is assumed that RPV pressure is >= 200 psi for these sequences. This is APB #3.

OP - Operator It is assumed that success on this branch results in RPV pressure below 200 Depressurizes the RPV psi that is then used to distinguish between APB #I versus APB #2, or APB

1. 3 versus APB #4.

RX - Core Melt A success on this branch signifies that there is no vessel breach. The Arrested in Vessel sequences following this path are typically grouped in APB #9. However, this assignment is overridden if the containment still fails due to subsequent CZ or HR-MU failures. In these cases, CZ failures are assigned to APB #3 or APB

1. 4 depending on the status of OP, and APB #6 is assigned for HR-MU failures with CV also failed, but to APB #7 is assigned for HR-MU failures with CV success.

CZ - Containment Failure of containment is assumed to result in an un-scrubbed release and is Intact Early grouped in APB #3 or APB #4 depending on RPV pressure.

FC and FD - If containment flooding is initiated and successfully completed without other Sontainment Flooding containment failures, this is assigned to APB #7 based on the interpretation nitiated and Completed that the successful completion of flooding requires venting.

HR - Containment Failure of this branch is assumed to result in a Late DW failure and APB #6 is

-teat Removal generally assigned for these sequences. The exception is the case where Maintained RX is successful as noted above.

CV - Containment Success of these nodes is used to indicate assignment to APB #7 for venting denting as long as the suppression pool is not bypassed in the SP node, and other containment failure nodes are not failed. This assignment applies to sequences with RX failures.

~~

30 PO4670060049-2706

Risk Imlvact Assessment o f Extendina Limerick Units 1 and 2 ILRT Interval Table 4.2-5 Limerick Level 2 Model Nodal Assumptions for Application to the NUREG/CR-4551 Accident Progression Bins Limerick Assumption Containment Event Tree Node

~ _ _

suppression pool bypass node is considered in the CETs to deirmine Not Bypassed whether the vent volume passes through the suppression pool or not. This node is used to distinguish between a W or DW failure as described in the other node assumption descriptions above.

MU - Inventory Success of this node in combination with success of RX is used to assign Makeup Available APB #9 unless otherwise overridden as described in the RX node discussion above.

DI - Drywell Intact / These nodes were utilized to distinguish between early Drywell Failures (APB WW - Wetwell Airspace #3, APB #4) and Wetwell Failures (APB #I, APB #2). All Late failures were Breach assumed to be drywell failures (APB #6), and therefore no sequences were assigned to APB #5.

RB - Release Mitigated The RB node, release mitigated in reactor building, is not credited as a in Reactor Building scrubbing mechanism in this analysis. The only scrubbing accounted for in the collapsed bins is distinguished by indicating a WW release (with the success of the SP node).

31 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRilntewal 4.3 Impact of Extension on Detection of Component Failures That Lead to Leakage (Small and Large)

The ILRT can detect a number of component failures such as liner breach, failure of certain bellows arrangements and failure of some sealing surfaces, which can lead to leakage. The proposed ILRT test interval extension may influence the conditional probability of detecting these types of failures. To ensure that this effect is properly accounted for, the EPRl Class 3 accident class as defined in Table 4.1-1 is divided into two sub-classes representing small and large leakage failures. These subclasses are defined as Class 3a and Class 3b, respectively.

The probability of the EPRl Class 3a failures may be determined, consistent with the NEI Guidance [3], as the mean failure estimated from the available data (i.e., 5 small failures in 182 tests leads to a 5/182=0.027 mean value). For Class 3b, using the original NEI Guidance [3], a non-informative prior distribution would be assumed for no large failures in 182 tests (i.e., 0.54 182+1) = 0.0027).

In a follow-on letter [21] to their ILRT guidance document [3], NEI issued additional information concerning the potential that the calculated delta LERF values for several plants may fall above the very small change guidelines of the NRC regulatory guide 1.I 74. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for delta LERF. NEI describes ways to demonstrate that, using plant-specific calculations, the delta LERF is smaller than that calculated by the simplified method.

The supplemental information states:

The methodology employed for determining LERF (Class 3b frequency) involves conservatively multiplying the CDF by the failure probability for this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, some plant-specific accident classes leading to core damage are likely to include individual sequences that either may already (independently) cause a LERF or could never cause a LERF, and are thus not associated with a postulated large Type A containment leakage path (LERF). These contributors can be removed from Class 3b 32 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval in the evaluation of LERF by multiplying the Class 3b probability by only that portion of CDF that may be impacted by type A leakage.

The application of this additional guidance to the analysis for Limerick, as detailed in Section 5, means that the Class 2 and Class 8 sequences are subtracted from the CDF that is applied to Class 3b. To be consistent, the same change is made to the Class 3a CDF, even though these events are not considered LERF. Class 2 and Class 8 events refer to sequences with either large pre-existing containment isolation failures or containment bypass events. These sequences are already considered to contribute to LERF in the Limerick Level 2 PRA analysis.

Consistent with the NEI Guidance [3], the change in the leak detection probability can be estimated by comparing the average time that a leak could exist without detection.

For example, the average time that a leak could go undetected with a three-year test interval is 1.5 years (3 yr / 2), and the average time that a leak could exist without detection for a ten-year interval is 5 years (10 yr / 2). This change would lead to a non-detection probability that is a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRT testing, given a 10-year vs. a 3-yr interval.

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

It should be noted that using the methodology discussed above is very conservative compared to previous submittals (e.g., the IP3 request for a one-time ILRT extension that was approved by the NRC [6]) because it does not factor in the possibility that the failures could be detected by other tests (e.g., the Type B local leak rate tests that will still occur.) Eliminating this possibility conservatively over-estimates the factor increases attributable to the ILRT extension.

33 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval 4.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage An estimate of the likelihood and risk implications of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is evaluated using the methodology from the Calvert Cliffs liner corrosion analysis [I 91.

The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner. The Limerick primary containment is a pressure-suppression BWR/Mark II containment type that also includes a steel-lined reinforced concrete structure.

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

Differences between the containment basemat and the containment walls The historical steel liner flaw likelihood due to concealed corrosion The impact of aging 0 The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw 34 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Assumptions Consistent with the Calvert analysis, a half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures. (See Table 4.4-1, Step 1.)

The two corrosion events used to estimate the liner flaw probability in the Calvert Cliffs analysis are assumed to be applicable to the Limerick containment analysis. These events, one at North Anna Unit 2 and one at Brunswick Unit 2, were initiated from the non-visible (backside) portion of the containment Iiner.

For consistency with the Calvert Cliffs analysis, the estimated historical flaw probability is limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data were not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date (and have been performed since the time frame of the Calvert Cliffs analysis), and there is no evidence that additional corrosion issues were identified. (See Table 4.4-1, Step I . )

Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages. (See Table 4.4-1, Steps 2 and 3.) Sensitivity studies are included that address doubling this rate every ten years and every two years.

0 In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere given that a liner flaw exists was estimated as 1.I % for the containment walls and dome region and 0.1 1% (10% less) for the basemat. These values were determined from an assessment of the containment fragility curve versus the ILRT test pressure. For Limerick the containment failure probabilities are conservatively assumed to be 10% for the drywell and wetwell outer walls, and since the basemat for the Limerick Mark II containment is in the suppression pool, it is judged that failure of this area would not lead to LERF. In any event, a 1% probability is conservatively assigned. Sensitivity studies are included that increase and decrease the probabilities by an order of magnitude. (See Table 4.4-1, Step 4.)

Consistent with the Calvert analysis, a 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used. To date, all liner corrosion events have been detected through visual inspection. (See Table 4.4-1, Step 5.) Sensitivity studies are included that evaluate total detection failure likelihood of 5% and 15%, respectively.

35 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval 0 Consistent with the Calvert analysis, all non-detectable containment failures are assumed to result in early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

Table 4.4-1 Steel Liner Corrosion Base Case Step Description Containment Wall Containment Basemat 1 Historical Steel Liner Flaw Events: 2 Events: 0 Likelihood (assume half a failure)

Failure Data: Containment location 2/(70

• 5.5) = 5.2E-3 0.5/(70

• 5.5) = 1.3E-3 specific (consistent with Calvert Cliffs analysis).

2 Age Adjusted Steel Liner Flaw Failure Failure Likelihood Year Rate During 15-year interval, assume 2.1E-3 1 5.OE-4 failure rate doubles every five years avg 5-10 1.3E-3 (14.9% increase per year). The 1.4E-2 15 3.5E-3 average for gfhto 10' year is set to the historical failure rate (consistent 15 year average = 15 year average =

with Calvert Cliffs analysis). 6.27E-3 1S7E-3 3 Flaw Likelihoodat 3,10, and 15 0.71% (1 to 3 years) 0.18% (1 to 3 years) years 4.06% (1 to I 0 years) 1.02% (I to I 0 years)

Uses age adjusted liner flaw 9.40% (1 to 15 years) 2.35% (1 to 15 years) likelihood (Step 2), assuming failure (Note that the Calvert Cliffs (Note that the Calvert Cliffs rate doubles every five years analysis presents the delta analysis presents the delta (consistent with Calvert Cliffs between 3 and 15 years of between 3 and 15 years of analysis - See Table 6 of Reference 8.7% to utilize in the 2.2% to utilize in the 1191). estimation of the delta- estimation of the delta-LERF LERF value. For this value. For this analysis the analysis the values are values are calculated based calculated based on the 3, on the 3, 10, and 15 year 10, and 15 year intervals.) intervals.)

4 Likelihood of Breach in 10% 1%

Containment Given Steel Liner Flaw The failure probability of the containment is assumed to be 10%

(compared to 1.1% in the Calvert Cliffs analysis). The basemat failure probability is assumed to be a factor of ten less, 1%, (compared to 0.1 1%

in the Calvert Cliffs analysis).

36 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRTlntewal Table 4.4-1 Steel Liner Corrosion Base Case Step I Description Containment Wall Containment Basemat 5 Visual Inspection Detection 10% 100%

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

Utilize assumptions consistent with flaws plus 5% likelihood Calvert Cliffs analysis. that the flaw is not visible (not through-cylinder but could be detected by ILRT).

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

I I Likelihoodof Non-Detected 0.0071% (at 3 years) 0.0018% (at 3 years)

Containment Leakage 0.71Yo

• 10%

• 10% 0.18%

• 1%* 100%

(Steps 3

• 4* 5) 0.0406% (at 10 years) 0.0102% (at 10 years) 4.06%

• 10%

• 10% 1.02%

• 1O h

• looo/o 0.0940% (at 15 years) 0.0235% (at 15 years) 9.40%

• 10%

• 10% 2.35%

• 1%

• 100%

The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the containment walls and the containment basemat:

At 3 years: 0.0071% + 0.0018% = 0.0089%

At I 0 years: 0.0406% + 0.0102% = 0.0508%

At 15 years: 0.0940% + 0.0235% = 0.1 175%

Limerick Past ILRT Results The surveillance frequency for Type A testing in NEI 94-01 under option B criteria is at least once per ten years based on an acceptable performance history (i.e., two consecutive periodic Type A tests at least 24 months apart where the calculated performance leakage rate was less than 1.0 La) and consideration of the performance factors in NEI 94-01, Section 11.3.

37 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Based on completion of two successful ILRTs at each of the Liemrick units, the current ILRT interval is once per ten years. The next Type A test for Limerick Unit Iis currently due to be completed by 3/2008, and by 3/2009 for Unit 2 [18].

Note that the probability of a pre-existing leakage due to extending the ILRT interval is based on the industry wide historical results as discussed in the NEI Guidance document, and the only portion of Limerick specific information utilized is the fact that the current ILRT interval is once per ten years.

NEI Interim Guidance This analysis uses the approach outlined in the NEI Interim Guidance. [3, 211 The nine steps of the methodology are:

1. Quantify the baseline (nominal three year ILRT interval) frequency per reactor year for the EPRl accident categories of interest. Note that EPRI categories 4, 5, and 6 are not affected by changes in ILRT test frequency.

2. Determine the containment leakage rates for EPRl categories 1 and 3 where category 3 is subdivided into categories 3a and 3b for small and large isolation failures, respectively.

3. Develop the baseline population dose (person-rem) for the applicable EPRl categories.

4. Determine the population dose rate (person-rem/year) by multiplying the dose calculated in Step (3) by the associated frequency calculated in Step (1).

5. Determine the change in probability of leakage detectable only by ILRT, and associated frequency for the new surveillance intervals of interest. Note that with increases in the ILRT surveillance interval, the size of the postulated leak path and the associated leakage rate are assumed not to change, however the probability of leakage detectable only by ILRT does increase.

38 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval

6. Determine the population dose rate for the new surveillance intervals of interest.

7. Evaluate the risk impact (in terms of population dose rate and percentile change in population dose rate) for the interval extension cases.

8. Evaluate the risk impact in terms of LERF.

9. Evaluate the change in conditional containment failure probability.

The first seven steps of the methodology calculate the change in dose. The change in dose is the principal basis upon which the Type A ILRT interval extension was previously granted and is a reasonable basis for evaluating additional extensions. The eighth step in the interim methodology calculates the change in LERF and compares it to the guidelines in Regulatory Guide 1.I 74. Because there is no change in CDF for Limerick, the change in LERF forms the quantitative basis for a risk informed decision per current NRC practice, namely Regulatory Guide 1.174. The ninth and final step of the interim methodology calculates the change in containment failure probability, referred to as the conditional containment failure probability, CCFP. The NRC has previously accepted similar calculations [7] for the change in CCFP as the basis for showing that the proposed change is consistent with the defense in depth philosophy. As such this last step suffices as the remaining basis for a risk informed decision per Regulatory Guide 1.I 74.

This group consists of all core damage accident sequences in which the containment is failed due to a pre-existing small leak in the containment structure that would be identifiable only from an ILRT (and thus affected by ILRT testing frequency).

39 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval 5.0 RESULTS The application of the approach based on NEI Interim Guidance [3, 211, EPRI-TR-104285

[2] and previous risk assessment submittals on this subject [6, 7, 20, 231 have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report. Table 5.0-1 lists these accident classes.

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

Core damage sequences in which the containment remains intact initially and in the long term (EPRI TR-104285 Class 1 sequences).

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

Core damage sequences in which containment integrity is impaired due to containment isolation failures of pathways left opened following a plant post-maintenance test. (For example, a valve failing to close following a valve stroke test.

(EPRI TR-104285 Class 6 sequences). Consistent with the NEI Guidance, this class is not specifically examined since it will not significantly influence the results of this analysis.

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

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

40 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRTInterval Table 5.0-1 ACCIDENT CLASSES Accident Classes (Containment Release Type) Description I No Containment Failure 2 Large Isolation Failures (Failure to Close)

I 3a I Small Isolation Failures (liner breach)

It 3b I Large Isolation Failures (liner breach) 1I 4 I Small Isolation Failures (Failure to seal -Type B)

I 5 I Small Isolation Failures (Failure to seal-Type C) 1I 6 I Other Isolation Failures (e.g., dependent failures)

I 7 I Failures Induced by Phenomena (Early and Late)

I 8 I Bypass (SGTR and Interfacing System LOCA)

I CDF 1 All CET End states (including very low and no release)

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

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

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

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

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

Step 5 - Determine the impact on the Conditional Containment Failure ProbabiI ity (CCFP) 41 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval 5.1 Step 1 - Quantify the Base-Line Risk in Terms of Frequency per Reactor Year This step involves the review of the Limerick containment event trees (CETs) and Level 2 accident sequence frequency results. The CETs characterize the response of the containment to important severe accident sequences. As described in Section 4.2, the Limerick CETs were examined and each endstate was applied to one of the Accident Progression Bins as defined in NUREGICR-4551. The correlation between the NUREGICR-4551 Accident Progression Bins to the EPRl containment release categories is shown in Table 5.1-1. This application combined with the Limerick dose (person-rem) results determined from Table 4.2-4 forms the basis for estimating the population dose for Limerick.

As previously described, the extension of the Type A interval does not influence those accident progressions that involve large containment isolation failures, Type B or Type C testing, or containment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks is included in the model. (These events are represented by the Class 3 sequences in EPRl TR-104285). 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 5.0-1 were developed for Limerick based on the assumptions shown in Table 4.2-5, determining the frequencies for Classes 3a and 3b, and then determining the remaining frequency for Class 1.

Furthermore, adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion of the steel liner per the methodology described in Section 4.4. The eight containment release class frequencies were developed consistent with the definitions in Table 5.0-1 as described following Table 5.1-1.

42 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Table 5.1-1 Containment Release Type Assignment from the NUREG/CR-4551 Consequence Model l

1 EPRl TR-104285 Containment Release NUREGICR-4551 0 TYPE Dose (Person-Rem)

Accident Limerick Dose 1.32E+04(I)

(VB, No CF, No Vent)

I0 O.OOE+OO (No core damage) 7.93E+06 3 7.93 E+06 (VB, Early DW, Hi Press) 3.97E+6 (*) 1 4.65 E+06 (VB, Early WW, Hi Press)

I I

2 2.91E+06 (VB, Early WW, Lo Press) 5 3.58 E+06 (VB, Late WW) 6 6.09E+06 (VB, Late DW)

~,

5.21E+06 (NoVB, No CF, No Vent) 6.01E+6 4 (VB, Early DW, Lo Press)

II 6.01E+06 (I No specific Release Bin for this category exists in NUREG/CR-4551. For simplicity, all sequences assigned to APB #3 is used in this analysis to represent EPRl Class 2 and all sequences assigned to APB

1. 1. 4are assigned to EPRl Class 8. This will not impact the calculated change for the proposed ILRT extension.

(*) Given that multiple NUREG/CR-4551 discrete scenarios apply to the broader EPRl type, the EPRI type dose is based on a weighted average (weights based on Limerick PRA scenario frequencies) of the applicable NUREG/CR-4551 APB doses.

43 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Class 1 Sequences. This group consists of all core damage accident progression bins for which the containment remains intact (modeled as Technical Specification Leakage).

The frequency per year for these sequences is 1.22E-O6/yr and is determined by subtracting all containment failure end states including the EPRVNEI Class 3a and 3b frequency calculated below, from the total CDF. For this analysis, the associated maximum containment leakage for this group is ILa, consistent with an intact containment evaluation.

Class 2 Sequences. This group consists of all core damage accident progression bins for which a failure to isolate the containment occurs. For simplicity, the frequency is obtained from all sequences that were assigned to APB #3 for Limerick which is 2.18E-O8/yr.

Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (e.g., containment liner) exists.

The containment leakage for these sequences can be either small (2La to 35L,) or large

(>35La).

The respective frequencies per year are determined as follows:

PROBclass-3a = probability of small pre-existing containment liner leakage

= 0.027 [see Section 4.31 PROB~lass-3b= probability of large pre-existing containment liner leakage

= 0.0027 [see Section 4.31 As described in section 4.3, additional consideration is made to not apply these failure probabilities on those cases that are already LERF scenarios (i.e., the Class 2 and Class 8 contributions).

Class-3a = 0.027 * (CDF-Class 2-Class 8)

= 0.027 * (3.70E 2.18E 4.65E-08) = 9.80E-O8/yr Class-3b = 0.0027 * (CDF-Class 2-Class 8)

= 0.0027 * (3.70E 2.18E 4.65E-08) = 9.80E-O9/yr For this analysis, the associated containment leakage for Class 3a is IOL, and for Class 3b is 35La. These assignments are consistent with the NEI Interim Guidance.

44 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRTlntewal Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis.

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

Class 6 Sequences. This group is similar to Class 2. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs. These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution.

Consistent with the NEI Interim Guidance, however, this accident class is not explicitly considered since it has a negligible impact on the results.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs. For this analysis, the associated radionuclide releases are based on the application of the Level 2 endstates to the Accident Progression Bins from NUREG/CR-4551 as described in Section 4.2. The Class 7 Sequences are divided into 6 categories which consists of Bins 1, 2, 5, 6, 7, and 9 from NUREG/CR-4551. The failure frequency and population dose for each specific APB is shown below in Table 5.1-2. The total release frequency and total dose are then used to determine a weighted average person-rem for use as the representative EPRl Class 7 dose in the subsequent analysis. Note that the total frequency and dose associated from this EPRl class does not change as part of the ILRT extension request.

45 PO4670060049-2706

Risk ImDact Assessment o f Extending Limerick Units 1 and 2 ILRTlntewal Table 5.1-2 ACCIDENT CLASS 7 FAILURE FREQUENCIES AND POPULATION DOSES

/LIMERICK BASE CASE LEVEL 2 MODEL) 7a (APB #I) 7.40E-09 4.65 E+06 3.44E-02 7b (APB #2) 2.39E-07 2.91E+06 6.97E-0I 7c (APB #5) O.OOE+OO 3.58E+06 O.OOE+OO 7d (APB #6) 7.76E-08 6.09 E+06 4.72E-01 1.47E-06 5.21E+06 7.67E+00 7f (APB #9) 5.09E-07 5.47E+05 2.78E-01 1 Class 7 Total 1 2.31E-06 I 3.97E+06(3) I 9.15E+00 I (I)

Population dose values obtained from Table 4.2-4 based on the Accident Progression Bin.

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

(3)

The weighted average population dose for Class 7 is obtained by dividing the total population dose risk by the total release frequency.

Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass occurs. For simplicity, the frequency is obtained from all sequences that were assigned to APB #4 for Limerick which is 4.65E-OWyr.

Summaw of Accident Class Frequencies In summary, the accident sequence frequencies that can lead to radionuclide release to the public have been derived consistent with the definition of Accident Classes defined in EPRI-TR-104285. Table 5.1-3 summarizes these accident frequencies by Accident Class.

46 PO4670060049-2706

Risk Impact Assessment of Extendina Limerick Units 1 and 2 ILRT Interval Table 5.1-3 RADIONUCLIDE RELEASE FREQUENCIES AS A FUNCTION OF ACCIDENT CLASS (LIMERICK BASE CASE)

Accident Classes (Containment Frequency Release Type) Description (per Rx-yr)

It 1 1 No Containment Failure I 1.22E-06 2(1) Large Isolation Failures (Failure to Close) 2.18E-08 3a Small Isolation Failures (liner breach) 9.8OE-08 3b Large Isolation Failures (liner breach) 9.80E-09 I 4 I Small Isolation Failures (Failure to seal -Type B) 1 N/A 5 Small Isolation Failures (Failure to seal-Type C) N/A Other Isolation Failures (e.g., dependent failures) N/A I 7 1 Failures Induced by Phenomena (Early and Late) 1 2.31E-06 8(l) Bypass (Interfacing System LOCA) 4.65E-08 CDF All CET End states (including very low and no release) 3.70E-06 (I) The EPRl Class 2 and Class 8 scenarios are assumed to be LERF in the ILRT methodology, and the sum of these sequence contributions from the simplified APB assignment of 6.83E-O8/yr agrees quite well with the Limerick detailed Level 2 PRA model reported LERF value of 6.8OE-O8/yr.

47 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILR T Interval 5.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose) per Reactor Year Plant-specific release analyses were performed to estimate the person-rem doses to the population within a 50-mile radius from the plant. The releases are based on information provided by NUREGICR-4551with adjustments made for the site demographic differences compared to the reference plant as described in Section 4.2, and summarized in Table 4.2-

4. The results of applying these releases to the EPRVNEI containment failure classification are as follows:

Class 1 1.32E+04 person-rem (at I.OLa) (I)

Class 2 7.93E+06(2)

Class 3a 1.32E+04 person-rem x 1OLa = 1.32E+05 person-rem (3)

Class 3b 1.32E+04 person-rem x 35La = 4.62E+05 person-rem (3)

Class 4 Not analyzed Class 5 Not analyzed Class 6 Not anaIyzed Class 7 3.97E+06 person-rem (4)

Class 8 6.01E+06 person-rem (5)

The Class 1, containment intact sequences, dose is assigned from the APB #8 (No CF, No Vent) from the NUREGKR-4551 adjusted dose for Limerick as shown in Table 4.2-4.

The Class 2, containment isolation failures, dose is approximated from APB #3 (VB, Early DW, Hi Press) from Table 4.2-4.

The Class 3a and 3b dose are related to the leakage rate as shown. This is consistent with the NEI Interim Guidance.

The Class 7 dose is assigned from the weighted average dose calculated from APBs # I , 2, 5, 6, 7, and 9 from Table 4.2-4 as detailed in Table 5.1-2 above.

Class 8 sequences involve containment bypass failures; as a result, the person-rem dose is not based on normal containment leakage. As an approximation, the releases for this class are assigned from APB #4 (VB, Early DW, Lo Press) from Table 4.2-4.

In summary, the population dose estimates derived for use in the risk evaluation per the EPRl methodology [2] containment failure classifications, and consistent with the NEI guidance [3] are provided in Table 5.2-1.

48 PO4670060049-2706

Risk Impact Assessment ofExtending Limerick Units 1 and 2 ILRT Interval Table 5.2-1 LIMERICK POPULATION DOSE ESTIMATES FOR POPULATION WITHIN 50 MILES Accident Representative Classes Accident Person-Rem Description (Containment Progression (50 miles)

Release Type) Bin (APB)

I I No Containment Failure (1 La)

~~ ~~

1 8 1.32E+04 Large Isolation Failures (Failure 3

to Close) 3a 1OLa 1 Small Isolation Failures (liner breach) 3b 35La 1 Large Isolation Failures (liner breach)

Small Isolation Failures (Failure 4.62E+05 4 NIA NA to seal -Type B)

Small Isolation Failures (Failure 5 N/A NA to seal-Type C)

Other Isolation Failures (e.g.,

6 NIA NA dependent faiIures)

Failures Induced by 7 3.97E+06 Phenomena (Early and Late)

Bypass (SGTR and Interfacing 8 4 6.01E+06 System LOCA)

The above dose estimates, when combined with the frequency results presented in Table 5.1-3, yield the Limerick baseline mean consequence measures for each accident class.

These results are presented in Table 5.2-2.

~ _ _ _ ~

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Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Table 5.2-2 LIMERICK ANNUAL DOSE AS A FUNCTlON OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR lLRT REQUIRED 3/10 YEARS Accident NEI Methodology Plus Change NEI Methodology Classes Person- Corrosion Due to (Containment Description Rem Person- Person- Corrosion Release (50 miles) Frequency Rem/yr Frequency Rem/yr Person-Type) (per Rx-Y~) Remlyr)

(Per Rx-Yr) (50 miles) (50 miles)

~~~~~

1 No Containment Failure (*) 1.32E+04 1.22E-06 1.60E-02 1.22E-06 1.60E-02 -4.25E-06 Large Isolation Failures (Failure to --

7.93E+06 2.18E-08 1.73E-01 2.18E-08 1.73E-01 Close)

Small Isolation Failures (liner --

1.32E+05 9.80E-08 1.29E-02 9.80E-08 1.29E-02 breach)

Large Isolation Failures (liner 4.62E+05 9.80E-09 4.52E-03 1.01E-08 4.67E-03 I.49E-04 breach)

Small Isolation Failures (Failure to NA NIA NIA NIA NIA NIA seal -Type B)

Small Isolation Failures (Failure to NA NIA NIA NIA NIA NIA seal-Type C)

Other Isolation Failures (e.g.,

NIA I 8 --

I CDF I.44E-04

1) Only release Classes 1 and 3b are affected by the corrosion analysis.

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

~~

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Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval The calculated dose for Limerick compares favorably with other locations given the relative population densities surrounding each location:

7 Annual Dose Plant Reference (Person-RemNr)

Indian Point 3 14315 [el Peach Bottom 6.2 ~ 3 1 Crystal River 1.4 [201 1 Limerick 1 9.6 I [Table 52-21 I 5.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval From 10-to-I5 Years The next step is to evaluate the risk impact of extending the test interval from its current ten-year value to fifteen-years. To do this, an evaluation must first be made of the risk associated with the ten-year interval since the base case applies to a 3-year interval (i.e., a simplified representation of a 3-in-I 0 interval).

Risk Impact Due to 10-year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and 3b sequences is impacted. The risk contribution is changed based on the NEI guidance as described in Section 4.3 by a factor of 3.33 compared to the base case values. The results of the calculation for a 10-year interval are presented in Table 5.3-1 for Limerick.

Risk Impact Due to 15-Year Test Interval The risk contribution for a 15-year interval is calculated in a manner similar to the 10-year interval. The difference is in the increase in probability of leakage in Classes 3a and 3b. For 51 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval this case, the value used in the analysis is a factor of 5.0 compared to the 3-year interval value, as described in Section 4.3. The results for this calculation are presented in Table 5.3-2.

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Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval Table 5.3-1 LIMERICK ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR ILRT REQUIRED 1/10 YEARS Accident NEI Methodology CI asses Person-(Containment Description Release Type) Rem Person-(50 miles) Frequency Rem/yr (per Rx-yr) (50 miles)

I

~p 1.32E+04 9.65E-07 Ip1.27E-02 Large Isolation Failures (Failure to Close) 7.93E+06 I 2.18E-08 1 1.73E-01 2.18E-08 1.73E-01 I --

Small Isolation Failures (liner breach) 3.26E-07 4.30E-02 I --

Large Isolation Failures (liner breach) 3.45E-08 I.59E-02 8.50E-04 Small Isolation Failures (Failure to seal -

N/A ~

Type 6)

Small Isolation Failures (Failure to seal-Type C)

N/A N/A I I

N/A I

Other Isolation Failures (e.g., dependent

, failures) N/A I N/A I N/A Failures Induced by Phenomena (Early and Late) 3.97E+06 I I

2.31E-06 1 I

9.15E+00 2.31E-06 1 9.15E+00 I --

6.01E+06 1 4.65E-08 I 2.79E-01 I 3.70E-06 I 9.67

('1 Only reie ise classes 1 and 3b are affected by the corrosion analysis.

(*) Characterized as ILa release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs.

Release classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

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Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Table 5.3-2 LIMERICK ANNUAL DOSE AS A FUNCTION OF ACCIDENT CLASS; CHARACTERISTIC OF CONDITIONS FOR ILRT REQUIRED 1/15 YEARS Accident NEI Methodology Person- Due to Classes Corrosion Description Rem Person-(Containment (50 miles) Rem/yr Release Type)

(per Rx-yr) (50 miles)

I

~ _ _ _ _ _ _ _ _ .

1 No Containment Failure (2) 1.32E+04 7.85E-07 1.03E-02 7.80E-07 1.03E-02 I -5.63E-05 2 I I

Large Isolation Failures (Failure to Close) I I

7.93E+06 2.18E-08 1.73E-01 2.18E-08 I 1.73E-01 I 3a I Small Isolation Failures (liner breach) I 1.32E+05 4.9OOE-07 6.46E-02 3b I Large Isolation Failures (liner breach) I 4.62E+05 4.900E-08 2.26E-02 Small isolation Failures (Failure to seal -

I NA 11 N/A N/A N/A I N/A I N/A I

.~

I 5

I Small Isolation Failures (Failure to seal-Type C) I NA N/A 1 N/A NIA N/A NIA 6 I I

Other Isolation Failures (e.g., dependent failures) I I

NA I

N/A I N/A N/A 1 N/A I N/A Failures Induced by Phenomena (Early 3.97E+06 I

and Late) 8 I Bypass (SGTR and ISLOCA) 6.01 E+06 CDF All CET end states 3.70E-06 9.70 3.70E-06 9.70 1.9 1E-03 (I) Only release classes 1 and 3b are affected by the corrosion analysis.

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

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Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval 5.4 Step 4 - Determine the Change in Risk in Terms of Large Early Release Frequency Regulatory Guide 1,174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 1O?yr and increases in LERF below 10-7/yr,and small changes in LERF as below 10-6/yr.Because the ILRT does not impact CDF, the relevant metric is LERF.

For Limerick, 100% of the frequency of Class 3b sequences can be used as a conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the NEI guidance methodology). Based on the original 3/10 year test interval assessment from Table 5.2-2, the Class 3b frequency is 1.01E-O8/yr. Based on a ten-year test interval from Table 5.3-1, the Class 3b frequency is 3.45E-O8/yr; and, based on a fifteen-year test interval from Table 5.3-2, it is 5.33E-O8/yr. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is 4.32E-O8/yr. Similarly, the increase due to increasing the interval from 10 to 15 years is 1.88E-08/yraAs can be seen, even with the conservatisms included in the evaluation (per the NEI methodology), the estimated change in LERF is below the threshold criteria for a very small change in risk when comparing the 15 year results to the current 10-year requirement or to the original 3-in-10 year requirement.

5.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability Another parameter that the NRC guidance in RG 1.I 74 states can provide input into the decision-making process is the change in the conditional containment failure probability (CCFP). The change in CCFP is indicative of the effect of the ILRT on all radionuclide releases, not just LERF. The CCFP can be calculated from the results of this analysis. One of the difficult aspects of this calculation is providing a definition of the failed containment.

In this assessment, the CCFP is defined such that containment failure includes all 55 PO4670060049-2706

Risk ImDact Assessment o f Extending Limerick Units I and 2 ILRT Interval radionuclide release end states other than the intact state. The conditional part of the definition is conditional given a severe accident (i.e., core damage).

The change in CCFP can be calculated by using the method specified in the NEI Interim Guidance. The NRC has previously accepted similar calculations [7] as the basis for showing that the proposed change is consistent with the defense-in-depth philosophy.

CCFP CCFP CCFP ACCFPg5-3 AccFP1s-10 3 in IOyrs 1 in IOyrs 1 in 15yrs 64.48% 65.14% 65.65% 1.17% 0.51%

CCFP = [ I - (Class 1 frequency + Class 3a frequency) / CDF]

• 100%

The change in CCFP of slightly more than 1% by extending the test interval to 15 years from the original 3-in-10 year requirement is judged to be insignificant.

5.6 Summary of Results The results from this ILRT extension risk assessment for Limerick are summarized in Table 5.6-1.

56 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval Table 5.6-1 Limerick ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (Including Age Adjusted Steel Liner Corrosion Likelihood)

Base Case Extend to Extend to EPRl 1 in 10 Years Per-Rem CDFNr I Per-RemNr I CDFNr I Per-RemNr 1 I 1.32E+04 I 1.22E-06 I 1.60E-02 I 9.63E-07 I 1.27E-02 I 7.80E-07 I 1.03E-02 2 I 7.93E+06 I2.18E-08 I 1.73E-01 I 2.18E-08 I 1.73E-01 I 2.18E-08 I 1.73E-01 3a I 1.32E+05 I9.80E-08 1 1.29E-02 I 3.26E-07 I 4.30E-02 I 4.90E-07 I 6.46E-02 3b I 4.62E+05 I 1.01E-08 I 4.67E-03 I 3.45E-08 I 1.59E-02 I 5.33E-08 I 2.46E-02 7 3.97E+06 2.31 E-06 9.15E+00 2.31E-06 I 9.15E+00 I 2.31E-06 I 9.15E+00 8 6.01 E+06 4.65E-08 2.79E-01 4.65E-08 [ 2.79E-01 I 4.65E-08 I 2.79E-01 Total I I3.70E-06 I 9.63 I 3.70E-06 I 9.67 I 3.70E-06 I 9.70 ILRT Dose Rate from 3a and 3b 1.76E-02 5.90E-02 8.92 E-02 Delta I From 3 yr I --- I 3.80E-02 I 6.59E-02 Total lose Rate From lo yr --- 2.79E-02 3b Frequency (LERF) I 1.01E-08 1 3.45E-08 I 5.33 E-08 Delta From 3 yr --- 2.44E-08 4.31E-08 LERF From 10 yr --- --- 1.88E-08 CCFP % 64.48% 65.14% 65.65%

Delta From 3 yr --- 0.66% 1.17%

CCFP %

I From 10 yr I --- --- 0.51%

57 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval 6.0 SENSITIVITIES 6.1 Sensitivity to Corrosion Impact Assumptions The results in Tables 5.3-1, 5.3-2, and 6.1-1 show that including corrosion effects calculated using the assumptions described in Section 4.4 does not significantly affect the results of the ILRT extension risk assessment.

Sensitivity cases were developed to gain an understanding of the sensitivity of the results to the key parameters in the corrosion risk analysis. The time for the flaw likelihood to double was adjusted from every five years to every two and every ten years. The failure probabilities for the containment wall and basemat were increased and decreased by an order of magnitude. The total detection failure likelihood was adjusted from 10% to 15%

and 5%. The results are presented in Table 6.1-1. In every case the impact from including the corrosion effects is very minimal. Even the upper bound estimates with very conservative assumptions for all of the key parameters yield increases in LERF due to corrosion of only 1.26E-Vyr. The results indicate that even with very conservative assumptions, the conclusions from the base analysis would not change.

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Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval Table 6.1-1 Steel Liner Corrosion Sensitivity Cases Increase in Class 3b Frequency (LERF)

Visual for ILRT Extension Containment Inspection & 3 to 15 years Breach Non-Visua I (per Rx-yr)

Flaws (Step 4 in the corrosion (Step 5 in the Increase Due to analysis) corrosion Total Increase Corrosion analysis)

Base Case Base Case Doubles every (10% Wall, 10% Wall, 4.31E-08 3.94E-09 5 yrs 1.O% Basemat) 100% Basemat Doubles every Base Base 4.82E-08 9.01 E-09 2 yrs Doubles every Base Base 4.25E-08 3.32E-09 10 yrs Base Base 15% Wall 4.47E-08 I 5.52E-09 Base Base 5% Wall 4.16E-08 I 2.37E-09 100% Wall, Base Base 7.86E-08 3.94E-08 10% Basemat 1.0% Wall, Base Base 3.96E-08 3.94E-10 0.1% Basemat P

59 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval 6.2 EPRl Expert Elicitation Sensitivity An expert elicitation was performed to reduce excess conservatisms in the data associated with the probability of undetected leaks within containment [22]. Since the risk impact assessment of the extensions to the ILRT interval is sensitive to both the probability of the leakage as well as the magnitude, it was decided to perform the expert elicitation in a manner to solicit the probability of leakage as a function of leakage magnitude. In addition, the elicitation was performed for a range of failure modes which allowed experts to account for the range of mechanisms of failure, the potential for undiscovered mechanisms, un-inspectable areas of the containment as well as the potential for detection by alternate means. The expert elicitation process has the advantage of considering the available data for small leakage events, which have occurred in the data, and extrapolate those events and probabilities of occurrence to the potential for large magnitude leakage events.

The basic difference in the application of the ILRT interval methodology using the expert elicitation is a change in the probability of pre-existing leakage in the containment. The basic methodology uses the Jefferys non-informative prior and the expert elicitation sensitivity study uses the results of the expert elicitation. In addition, given the relationship between leakage magnitude and probability, larger leakage that is more representative of large early release frequency, can be reflected. For the purposes of this sensitivity, the same leakage magnitudes that are used in the basic methodology (i.e., 10 La for small and 35 La for large) are used here. Table 6.2-1 illustrates the magnitudes and probabilities of a pre-existing leak in containment associated with the Jefferys non-informative prior and the expert elicitation statistical treatments. These values are used in the ILRT interval extension for the base methodology and in this sensitivity case. Details of the expert elicitation process, and the input to expert elicitation as well as the results of the expert elicitation, are available in the various appendices of the EPRl report [22].

60 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Table 6.2-1 EPRl Expert Elicitation Results 35 2.7E-03 9.86E-04 64%

A summary of the results using the expert elicitation values for probability of containment leakage is provided in Table 6.2-2. As mentioned previously, probability values are those associated with the magnitude of the leakage used in the Jefferys non-informative prior evaluation (1OLa for small and 35La for large). The expert elicitation process produces a probability versus leakage magnitude relationship and it is possible to assess higher leakage magnitudes more reflective of large early releases but these evaluations are not performed in this study. Alternative leakage magnitudes could include consideration of 100 to 600 La where leakage begins to approach large early releases.

The net affect is that the reduction in the multipliers shown above has the same impact on the calculated increases in the LERF values. The increase in the overall probability of LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is I.43E-O8/yr. Similarly, the increase due to increasing the interval from 10 to 15 years is 5.98E-O9/yr. As such, if the expert elicitation mean probability of occurrences are used instead of the non-informative prior estimates, the change in LERF for Limerick is even further below the threshold criteria for a very small change in risk when compared to the current l-in-10 or original 3-in-10 year requirement. The results of this sensitivity study are judged to be more indicative of the actual risk associated with the ILRT extension than the results from the assessment as dictated by the NEI methodology values, and yet are still conservative given the assumption that all of the Class 3b contribution is considered to be LERF.

61 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Table 6.2-2 Limerick ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (Based on EPRI Expert Elicitation Leakage Probabilities)

Base Case Extend to Extend to EPRI DOSE 3 in I 0 Years I in I 0 Years 1 in 15 Years Class Per-Rem CDFNr Per-RemNr CDFNr Per-RemNr CDFNr Per-RemNr 1 I 1.32E+04 1 1.31E-06 I 1.72E-02 I 1.26E-06 I 1.67E-02 I 1.24E-06 I 1.63E-02 2 I 7.93E+06 I2.18E-08 I 1.73E-01 I 2.18E-08 I 1.73E-01 I 2.18E-08 I 1.73E-01 3a I 1.32E+05 I 1.41E-08 I 1.86E-03 I 4.69E-08 I 6.19E-03 I 7.04E-08 I 9.29E-03 3b I 4.62E+05 I 3.58E-09 I 1.65E-03 I 1.19E-08 I 5.50E-03 1 1.79E-08 1 8.26E-03 7 1 3.97E+06 I 2.31E-06 1 9.15E+00 I 2.31E-06 I 9.15E+00 I 2.31E-06 I 9.15E+00 8 I 6.01E+06 I4.65E-08 I 2.79E-01 I 4.65E-08 I 2.79E-01 I 4.65E-08 I 2.79E-01 Total I I3.7OE-06 I 9.62 I 3.70E-06 I 9-63 I 3.70E-06 I 9-64 ILRT Dose Rate from 3a and 3b 3.51E-03 I 1.17E-02 1 1.75E-02 Delta From 3 yr --- 7.63 E-03 1.31E-02 Total Dose Rate From 'lo yr --- --- 5.47E-03 3b Frequency (LERF) 1 From i o yr j I I LERF --- 5.98E-09 62 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval

7.0 CONCLUSION

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

Reg. Guide 1.174 [4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Reg. Guide 1.174 defines very small changes in risk as resulting in increases of CDF below 10-6/yrand increases in LERF below 10-7/yr.Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in internal events LERF resulting from a change in the Type A ILRT test interval from three in ten years to one in fifteen years is estimated as 4.31E-81yr using the NEI guidance as written, and at 1.43E-8/yr using the EPRl Expert Elicitation methodology. In either case, the estimated change in LERF is determined to be very small using the acceptance guidelines of Reg. Guide 1. I74.

The change in Type A test frequency to once-per-fifteen-years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.066 person-rem/yr using the NEI guidance, and drops to 0.013 person-rem/yr using the EPRI Expert Elicitation methodology. Therefore, in either case, the risk impact when compared to other severe accident risks is negligible.

The increase in the conditional containment failure frequency from the three in ten year interval to one in fifteen year interval is about 1.2% using the NEI guidance, and drops to about 0.4% using the EPRl Expert Elicitation methodology. Although no official acceptance criteria exist for this risk metric, it is judged to be very small.

Since the increase in LERF falls well below the small change category using the acceptance guidelines of Reg. Guide 1.I 74, a detailed examination of the external events impact is not required, nor would it change the conclusions from this assessment.

63 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units I and 2 ILRT Interval Therefore, increasing the ILRT interval to 15 years is considered to be insignificant since it represents a very small change to the Limerick Generating Station risk profile.

Previous Assessments The NRC in NUREG-1493 [5] has previously concluded that:

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

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

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

64 PO4670060049-2706

Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRT Interval

8.0 REFERENCES

Nuclear Energy Institute, lndustry Guideline for lrnplementing Performance-Based Option of 70 CFR Part 50, Appendix J, NEI 94-01, July I995.

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

Letter from A. Pietrangelo (NEI) to NEI Administrative Points of Contact, Interim Guidance for Performing Risk lmpact Assessments in Support of One-Time Extensions for Containment Integrated Leak Rate Test Surveillance lntetvals, November 13,2001.

U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, July 1998.

Performance-Based Containment Leak-Test Program, NUREG-I493, September 1995.

Letter from R.J. Barrett (Entergy) to US. Nuclear Regulatory Commission, IPN 007, dated January 18,2001.

United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MBOI78), April 17,2001.

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

valuation of Severe Accident Risks: Peach Bottom, Unit 2, Main Report NUREG/CR-4551,SAND86-I309, Volume 4, Revision 1, Part 1, December 1990.

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

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

US. Nuclear Regulatory Commission, Technical Findings and Regulatory Analysis for Generic Safety lssue /I.E.4.3 Containment Integrity Check, NUREG-I273, April 1988.

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Risk Impact Assessment of Extending Limerick Units 1 and 2 ILRTInterval Pacific Northwest Laboratory, Review of Light Water Reactor Regulatory Requirements, NUREG/CR-4330, PNL-5809, Vol. 2, June 1986.

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

U.S. Nuclear Regulatory Commission, Reactor Safety Study, WASH-I 400, October 1975.

Exelon Risk Management Team, Limerick PRA Summary Notebook, 20043 Model, LGS-PRA-013, July 2005.

E-mail from G. Teagarden (ERIN) to D. Vanover (ERIN), Year 2070 Populations for ILRT, August 16,2006.

E-mail from T. Loomis (Exelon) to B. Sloane (ERIN), FW: lnformafion Request in Supporf of Limerick ILRT Extension, December 5,2006.

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.

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

Letter from A. Pietrangelo (NEI) to NEI Administrative Points of Contact, "One-Time Extension of Containment Integrated Leak Rate Test Interval - Additional Information", November 30,2001.

Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI, Palo Alto, CA: 1009325, Revision 1, December 2005.

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, LGS UFSAR Table 2.1-13, "Population Distribution 10-50 Miles (2010)".

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