JAFP-03-0108, Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT) Interval Extension

Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT) Interval Extension

ML032170128
Person / Time
Site:	FitzPatrick
Issue date:	07/28/2003
From:	Ted Sullivan Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-03-0108
Download: ML032170128 (199)

Text

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. Fitzpatrick NPP

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110 Altergy Lycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 T.A. Sullivan July 28, 2003 Site Vice President - JAF JAFP-03-0108 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, DC 20555-0001

SUBJECT:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT) Interval Extension Gentlemen:

In accordance with the provisions of 10 CFR 50.90, Entergy Nuclear Operations, Inc. (ENO) is submitting a request for an amendment to the Technical Specifications (TS) for the James A.

FitzPatrick Nuclear Power Plant (JAFNPP).

The proposed license amendment would revise Technical Specification section 5.5.6 "Primary Containment Leakage Rate Testing Program" to allow a one-time interval extension for the JAF Type A, Integrated Leakage Rate Test (ILRT) of no more than five (5) years. This revision takes a one time exception to the ten (10) year frequency of the performance-based leakage rate testing program for Type A tests as required by NEI 94-01, revision 0, "Industry Guideline For Implementing Performance-Based Option of 10 CFR Part 50, Appendix J", and endorsed by 10 CFR 50, Appendix J, Option B. The one time exception is to the requirement of NEI 94-01 to perform an integrated leak rate test (ILRT) at a frequency of up to ten years, with allowance for a 15 month extension. The exception is to allow ILRT testing within fifteen years from the last ILRT, performed on March 7, 1995. This application represents a cost beneficial licensing change. The integrated leak rate test imposes significant expense on the station while the safety benefit of performing it within 10 years, versus 15 years, is minimal.

This request is made with a risk-informed basis as described in Regulatory Guide 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications." The plant specific risk assessment evaluation performed in support of this request, Engineering Report JAF-RPT-03-00007, "Risk Impact Assessment of Extending Containment Type A Test Interval" is forwarded as Attachment 4. This assessment was performed following the guidelines of NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J", the methodology used in EPRI TR-104285, "Risk Assessment of Revised Containment Leak Rate Testing Intervals," and the guidance provided in NRC Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis". The assessment also followed the guidance and additional information distributed by NEI in November 2001 to their Administrative Points of

%C(

Contact regarding risk assessment evaluation of one-time extensions of containment ILRT intervals and the approach outlined in the Indian Point Unit Three Nuclear Power Plant ILRT extension submittal. Detailed bases and considerations for the risk assessment evaluation are identified in Attachment 4.

The conclusion of the risk assessment is that the one-time increase in the Type A test interval from ten to fifteen years would result in negligible effect on plant risk. In particular, the conclusion of the plant internal events risk associated with extending the Type A ILRT interval from ten to fifteen years is as follows:

1. The increase in risk on the total integrated plant risk as measured by person-rem/year increases for those accident sequences influenced by Type A testing, given the change from a l-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.56%

(0.004 person rem/yr). This value can be considered to be a negligible increase in risk.

2. Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 104/yr and increases in large early release frequency (LERF) below 10- /yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from l-in-I0 years to 1-in-15 years is 1.09 x 10 4/yr. Since Regulatory Guide 1.174 defines very small changes in LERF as below 10i 7/yr, increasing the ILRT interval at FitzPatrick from the currently allowed one-in-ten years to one-in-fifteen years is non-risk significant from a risk perspective.

3. The change in conditional containment failure probability (CCFP) is calculated to demonstrate the impact on 'defense-in-depth'. For the current ten-year ILRT interval, sequences involving no containment failure or small releases contribute 27.4% to the overall plant risk. Alternatively stated, the contribution of sequences involving containment failure for the ten-year interval is 72.6%. These numbers are consistent with those documented in the FitzPatrick IPE. For the proposed fifteen-year interval, the contribution of sequences involving containment failure increased to 73.03%. Therefore, ACCFPI0.15 is found to be 0.43%. This signifies a very small increase and represents a negligible change in the FitzPatrick containment defense-in-depth.

Additional risk considerations (external event hazards, potential containment liner corrosion) were also evaluated. These are summarized in Attachment 4.

Further assurance of containment leak-tight integrity is provided through periodic inservice inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI. More specifically, Subsection IWE provides the rules and requirements for inservice inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations, 10 CFR 50.55a(b)(2)(ix)(E), require licensees to conduct visual inspections of the accessible areas of the containment 3 times every 10 years. These requirements are not affected by 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, as discussed more fully in the accompanying Safety Evaluation (Attachment 2). provides a description of the inservice inspection program for the FitzPatrick containment and a summary of recent inspection results. Inspections conducted to date have not identified any degradation or other condition that would threaten the structural integrity of the containment.

The NRC has approved similar risk-informed submittals relating to a one-time extension of a Type A test interval for a number of plants, including Carolina Power & Light Company's Brunswick Unit 1, Exelon Nuclear's Peach Bottom Unit 3, and Entergy's Indian Point 3 (IP3) nuclear power plant. The IP3 request was submitted on September 6, 2000 (IPN-00-062) and supplemented on January 18, 2001 (IPN-01-007) and on April 2, 2001 (IPN-01-030). The NRC approval was granted on April 17,2001 (TAC No. MB0178).

The signed original of the Application for Amendment to the Operating License is enclosed fbr filing. Attachment 1 contains the proposed new TS pages and Attachment 2 is the Safety Evaluation for the proposed changes. A markup of the affected TS pages is included as . As previously indicated, Attachment 4 provides the supporting risk assessment evaluation, while Attachment 5 provides a synopsis of containment related inservice testing.

There are no TS Bases associated with this request.

ENO requests approval of the proposed license amendment by August 10, 2004 with the amendment being implemented within thirty days following approval. The requested approval date and implementation period will allow sufficient time for effective planning and scheduling of affected activities associated with Refueling Outage 16, scheduled to begin on October 4, 2004.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated New York State official.

These are no new commitments made in this letter. If you should have any questions regarding the submittal, please contact Mr. Andrew Halliday at (315) 349-6055.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on this the ~ay of Jd/y 2003.

Very truly yoursel Site Vice President Attachments: 1. Revised Technical Specification Pages

2. Safety Evaluation

3. Marked Up Technical Specification Pages

4. Plant Specific Risk Assessment Report

5. Containment Inservice Inspection Program Summary cc: Regional Administrator, Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Office of the Resident Inspector U. S. Nuclear Regulatory Commission P. 0. Box 136 Lycoming, NY 13093 Mr. G. Vissing, Project Manager Project Directorate I Division of Licensing Project Management Office or Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop: 8C2 Washington, DC 20555 Mr. Peter R. Smith, Acting President New York State Energy, Research, and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, NY 12203-6399

ATTACHMENT I to JAFP-03-0108 REVISED TECHNICAL SPECIFICATION PAGES Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT)

Interval Extension Entergy Nuclear Operations, Inc.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 Primary Containment Leakage Rate Testing Program This program implements the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50. Appendix J. Option B. as modified by agproved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163. "Performance-Based Containment Leak-Test Program,"

dated September 1995. as modified by the following exceptions:

• NEI 94-01-1995. Section 9.2.3: The first Type A test performed after the March 7. 1995 Type A test shall be performed no later than March 7. 2010.

• Type C testing of valves not isolable from the containment free air space may be accomplished by pressurization in the reverse direction, provided that testing in this manner provides equivalent or more conservative results than testing in the accident direction. If potential atmospheric leakage paths (e.g., valve stem packing) are not subjected to test pressure. the portions of the valve not exposed to test pressure shall be subjected to leakage rate measurement during regularly scheduled Type A testing. A list of these valves. the leakage rate measurement method. and the acceptance criteria, shall be contained in the Program.

a. The peak primary containment internal pressure for the design basis loss of coolant accident. Pal is 45 psig.

b. The maximum allowable primary containment leakage rate, La.

at Pal shall be 1.5% of containment air weight per day.

c. The leakage rate acceptance criteria are:

1. Primary containment leakage rate acceptance criteria is

< 1.0 La. During plant startup following testing in accordance with this program. the leakage rate acceptance criteria are < 0.60 L for the Type B and Type C tests.

and < 0.75 La for the Vype A tests.

2. Air lock testing acceptance criteria are:

(a) Overall air lock leakage rate is < 0.05 La when tested at 2 Pa; and (b) For each door seal, leakage rate is < 120 scfd when tested at 2 Pa.

(continued)

JAFNPP 5.5-5 Amendment

ATTACHMENT 2 to JAFP-03-0108 SAFETY EVALUATION Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT)

Interval Extension Entergy Nuclear Operations, Inc.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 1 of 11 I. DESCRIPTION This application for amendment to the James A. FitzPatrick (JAF) Technical Specifications (TS) proposes to revise Technical Specification 5.5.6 nPrimary Containment Leakage Rate Testing Program". This revision takes a one time exception to the ten (10) year frequency of the performance-based leakage rate testing program for Type A tests as required by NEI 94-01 (Reference 1). The one time exception is to the requirement of NEI 94-01 to perform an integrated leak rate test (ILRT) at a frequency of up to ten years, with allowance for a fifteen month extension. The exception is to allow ILRT testing within fifteen years from the last ILRT, performed on March 7, 1995. This application represents a risk informed, cost beneficial licensing change. The integrated leak rate test imposes significant expense on the station while the safety benefit of performing it within ten years, versus fifteen years, is minimal. The specific change is as follows:

1. TS Section 5.5.6, page 5.5-5 Replace:

"...as modified by the following exception:"

With:

"...as modified by the following exceptions."

Add:

N. NEI 94 1995, Section 9.2.3: The first Type A Test performed after the March 7, 1995 Type A test shall be performed no later than March 7, 2010."

In addition, the existing paragraph beginning with "Type C testing of valves ... Nis bulleted and indented for formatting consistency.

11. PURPOSE OF THE PROPOSED CHANGE The current FitzPatrick ten year Type A test is due on March 7, 2005. This test is currently scheduled to be performed during refuel outage (RO), R016, scheduled for October 2004.

This one time exception will permit deferral of the test beyond RO1 6 to a later outage within the five year extension window. Deferring this test for an additional five (5) years will result in substantial cost savings associated with both direct costs for performing the test and indirect costs associated with critical path outage time. Cost savings have been conservatively estimated for this outage at $660,000, including a minimum reduction of thirty-six hours of critical path outage time.

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 2 of 11 Ill. SAFETY IMPLICATIONS OF THE PROPOSED CHANGE

1. BACKGROUND 10 CFR 50, Appendix J, Option B Requirements:

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 the allowable leakage values specified in the Technical Specifications. The limitation of containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant design basis accident. Appendix J identifies three types of required tests:

Type A tests, intended to measure the primary containment overall integrated leakage rate; Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage-limiting boundaries for primary containment penetrations; and Type C tests, intended to measure containment isolation valve leakage rates. Type B and C tests identify the vast majority of potential containment leakage paths. Type A tests identify overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Type B and C testing.

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 "Performance-Based Requirements." In October 1996, Amendment 234 (Reference 2) was issued to the FitzPatrick Operating License to permit implementation of 10 CFR 50, Appendix J, Option B. Amendment 234 added a Technical Specification section (section 6.20 in custom Technical Specifications, now section 5.5.6 following conversion to standard Technical Specifications) requiring Type A, B and C testing in accordance with Regulatory Guide (RG) 1.163 (Reference 3). Regulatory Guide 1.163 specifies a method acceptable to the NRC for complying with Option B by approving the use of NEI 94-01 and ANSIIANS 56.8 - 1994 (Reference 4), subject to several regulatory positions in the guide. NEI 94-01 specifies an initial Type A test interval of 48 months, but allows an extended interval of ten years, based upon two consecutive successful tests. There is also a provision for extending the test interval an additional fifteen months under certain circumstances.

The adoption of the Option B performance-based containment leakage rate testing program did not alter the basic method by which Appendix J leakage rate testing is performed, but did alter the frequency of measuring primary containment leakage in Type A, B and C tests. Frequency is based upon an evaluation which looks at the *as found" leakage history to determine a frequency for leakage testing which provides assurance that leakage limits will be maintained. The changes to Type A test frequency allowed by Option B do not directly result in an increase in containment leakage, only the interval at which such leakage is measured on an integrated basis. Similarly, the proposed change to the Type A test frequency will not directly result in an increase in containment leakage.

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 3 of 11 The extended frequency interval for testing allowed by NEI 94-01 is based upon a generic evaluation documented in NUREG-1493, uPerformance-Based Containment Leak-Test Program" (Reference 5). NUREG-1493 made the following observations with regard to extending the test frequency:

"Reducing the Type A (ILRT) testing frequency to one per twenty years was found to lead to an imperceptible increase in risk. The estimated increase in risk is small because 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 tests have been only marginally above the existing requirements. Given the insensitivity of risk to containment leakage rate, and the same fraction of leakage detected solely by Type A testing, increasing the interval between ILRT testing had minimal impact on public risk."

• While Type B and C tests identify the vast majority (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.'

Exceptions to the requirements of RG 1.163, are allowed by 10 CFR 50, Appendix J, Option B,Section V.B, "Implementation," which states "The Regulatory Guide or other implementing document used by a licensee, or applicant for an operating license, to develop a performance based leakage-testing program must be included, by general reference, in the plant technical specifications. The submittal for technical specification revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the Commission and endorsed in a regulatory guide." Since exceptions meeting the stated requirements are permitted, Technical Specification amendment applications satisfying these requirements do not require an exemption to Option B.

2. PLANT SPECIFIC INFORMATION 10 CFR 50 Appendix J, Option B Plant Specific Implementation As previously stated, Amendment 234 to the FitzPatrick Operating License permitted implementation of 10 CFR 50, Appendix J, Option B for FitzPatrick. Amendment 234 requires Type A, B and C testing be conducted in accordance with Regulatory Guide (RG) 1.163, which in turn endorses the methodology for complying with Option B identified in NEI 94-01. The surveillance frequency for Type A testing in NEI 94-01 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 1.0L,) and consideration of the performance factors in NEI 94-01, Section 11.3. The two most recent Type A tests at JAF have been successful. The following extract from JAF-RPT-PC-02342, "James A. Fitzpatrick

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 4 of 11 Nuclear Power Plant Primary Containment Leakage Rate Testing Program Plan' (Reference 6), describes the test results:

"5.11 BASELINE CONTAINMENTEVALUATIONS The performance leakage rates are calculated in accordance with NEI 94-01, Section 9.1.1. The performance leakage rate includes the Type A UCL plus the as-left minimum pathway leakage rate for all Type B and C pathways not in service, isolated, or not lined up in their test position. In addition, leakage pathways that were isolated during the performance of the test are included in the test results by adding the as-found minimum pathway leakage rate to the Type A UCL. The performance leakage rate does not include leakage savings (i.e., improvements to Type B and C components made prior to the Type A test).

For the June 1990 Type A test, the Total Time UCL leakaoe rate was 0.27045%

wt./ day. The minimum pathway leakage rate for Type B and C pathways not in service and level corrections was 0.1331 % wtJday. The performance leakage rate was 0.27045 + 0.1331 = 0.4035 % wtlday, which was acceptable. There were no leakage pathways isolated during the performance of the test.

For the March 7. 1995 Periodic Type A test, the total time UCL leakage rate, was 0.0394 % wt.Idav. The minimum pathway leakage rate for Type B and C pathways not in service was 0.023595 % wtlday. Therefore, the performance leakage rate was 0.0394 + 0.023595 = 0.062995 % wtlday."

These results compare with an acceptable design leakage rate for FitzPatrick (1.0 La) of 1.5

% wt./day. Based upon these two consecutive successful tests, the current ILRT interval requirement for JAF is ten years.

The results of the two previous ILRT tests conducted for FitzPatrick (May 1985 and April 1987) were also reviewed. Leakage rates for these tests were 0.281214 % wt.lday and 0.304442 % wt./day respectively.

The results of these four tests conducted over an approximate ten year interval demonstrate consistent low leakage for the FitzPatrick primary containment structure.

Plant Testing and Inspection Programs In addition to periodic Type A testing, various inspections and tests are routinely performed to assure primary containment integrity. These include Type B and C testing performed in accordance with Appendix J, Option B; inspection activities performed as part of the plant Inservice Inspection program; maintenance rule related inspections; and others. The aggregate results of these inspections serve to provide a high degree of assurance of continued primary containment integrity.

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 5 of 11 Type B and Type C Program The FitzPatrick Appendix J, Type B and Type C test programs are described in JAF-RPT-PC-02342, "James A. Fitzpatrick Nuclear Power Plant Primary Containment Leakage Rate Testing Program Plan." Regarding the scope of these programs, the program plan states, in part:

Electrical penetrations, airlocks, hatches, flanges, and valves within the scope of the Appendix J Program Plan and which are not exempt shall be tested in compliance with the requirements of 10 CFR 50 Appendix J Option B and Regulatory Guide 1.163.

The Type B and C test program provides a means to detect or measure leakage across pressure containing or leakage limiting barriers of the primary reactor containment. The resufts of the test program are used to ensure that proper maintenance and repairs are made on the primary reactor containment components over their service life. The Type B and C test program provides a means to protect the health and safety of plant personnel and the public by maintaining the leakage from these components below required levels.

The Type B and C test program consists of local leak rate testing of penetrations which utilize a resilient seal, expansion bellows, double gasketed manways, hatches, and flanges, drywell airlock, and containment isolation valves that serve as a barrier to the release of the post accident primary containment atmosphere.

These components are tested with air or nitrogen at a pressure greater than or equal to 45 psig (Pj), not to exceed 56 psig (this does not account for instrument inaccuracies). Tests performed on-line will assure that full accident differential pressure is applied across the barrier under test, accounting for containment inerting, or system head pressure. The Main Steam Isolation Valves are tested at a greater than or equal to 25 psig per the technical specifications.

As previously noted, Type B and Type C testing evaluate all but a small portion of potential containment leakage pathways. Nothing in this amendment request affects the scope, performance or scheduling of Type B or Type C tests. These programs will continue to provide a high degree of assurance that primary containment integrity is maintained.

Inservice Inspection (ISI) Program Effective September, 1996, the NRC endorsed Subsections IWE and IWL of ASME Section Xi, 1992 Edition including 1992 Addenda. These subsections contain inservice inspection and repair/replacement rules for Class MC and Class CC components. The reactor containment is a free-standing steel containment, to which the requirements of Subsection IWE apply. These requirements are included in the inservice inspection program for FitzPatrick, described in [[::JAF-ISI-0002|JAF-ISI-0002]], 'Third Inservice Inspection Interval Inservice Inspection Programs (Reference 7).

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 6 of 11 Attachment 5 provides a description of ISI program inspection activities for the FitzPatrick containment and a summary of recent inspection results. Basically, these inspections assess the condition of containment structural components and coatings, providing assurance against structural or material degradation. As such, these inspections complement Type A testing and provide a high degree of assurance of continued containment structural integrity.

Additional Tests and Inspections Additional tests and inspections are conducted which assure the continued good material condition of the containment and associated containment integrity. These include Maintenance Rule inspections; periodic surveillance testing conducted, in addition to Type A, B, and C tests, as part of the plant Primary Containment Leakage Rate Testing Program; periodic walkdowns; and post maintenance tests.

Attachment 5 provides further details.

Plant OperationalPerfonnance:

The James A. Fitzpatrick NPP is a power uprated 881 Mwe, General Electric Boiling Water Reactor (BWR4). The reactor is contained in a Chicago Bridge and Iron Works supplied Mark 1, Free Standing Steel Containment Building. The containment consists of two primary interconnected structures: the drywell, housing the reactor and related components, and a toroidal suppression chamber (torus). The drywell, which includes the major primary containment volume, is inerted with nitrogen and maintained at a nominal 1.7 psid positive pressure with respect to the torus. This pressure differential is required by Technical Specifications (LCO 3.2.6.4) and monitored by plant during instrumentation and through periodic surveillance (SR 3.6.2.1). The differential is initially established during drywell inerting by pressurizing the drywell using plant nitrogen.

During plant operation, the combination of a small amount of normal instrument nitrogen leakage within the drywell and leak tightness of the containment structure is such that nitrogen typically does not have to be added to the drywell to maintain the required differential.

Plant Technical Specifications state:

3.6 CONTAINMENT SYSTEMS 3.6.2.4 Drywell-to-Suppression Chamber Differential Pressure LCO 3.6.2.4 The drywell pressure shag be maintained Ž 1.7 psi above the pressure of the suppression chamber.

NOTE Not required to be met for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during Surveillances that cause or require the drywell-to-suppression chamber differential pressure to be outside the limit.

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 7 of 11 APPLICABILITY: MODE 1 during the time period:

a. From 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER is > 15% RTP following startup, to

b. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing THERMAL POWER to < 15% RTP prior to the next scheduled reactor shutdown.

Required action if the differential pressure requirement is not met is to restore the parameter to within limits within eight hours or reduce reactor power to <15% within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. Surveillance testing requirements specify verifying the differential pressure within limit on a 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance frequency.

The major portion of the FitzPatrick containment is thus normally pressurized. Although the pressure is not as significant as that resulting from a Design Basis Accident, the fact that the containment is normally pressurized provides a degree of assurance of containment structural integrity (i.e. no large leak paths in the containment structure).

Significant leakage would be identified using plant instrumentation or through increased nitrogen usage (periodically monitored) needed to maintain the required differential pressure, and would be investigated promptly and addressed within the scope of the plant Corrective Action system. This feature is a complement to periodic visual inspections of the interior and exterior of the containment structure, and serves to provide added assurance of structural integrity for those areas that may be inaccessible for visual examination.

3. PLANT SPECIFIC RISK ASSESSMENT Attachment 4 contains a detailed, plant specific risk assessment performed in support of this amendment request. This assessment evaluates the risk impact of extending the Type A test interval for FitzPatrick from ten to fifteen years. The assessment complements the studies cited in NUREG-1493 that concluded that Type A testing intervals could be extended to as much as twenty years with negligible impact on risk.

The conclusions of the plant specific assessment are that effects on risk from the requested change are negligible or non-risk significant. Methodology and a summary of results are as follows:

Approach and Methodology:

In performing the risk assessment evaluation, the guidelines of NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix JX, the methodology used in EPRI TR-1 04285, ""Risk Assessment of Revised Containment Leak Rate Testing Intervals," and the NRC Regulatory Guide 1.174, VAn Approach for Using Probabilistic Risk Assessment In Risk-informed Decisions On Plant-Specific Changes to the Licensing Basis" were used. The assessment also followed the guidance and additional information distributed by NEI in November 2001 to their Administrative Points of Contact regarding risk assessment evaluation

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 8 of 11 of one-time extensions of containment ILRT intervals and the approach outlined in the Indian Point Unit Three Nuclear Power Plant ILRT extension submittal.

The risk assessment evaluation uses the current FitzPatrick Individual Plant Examination (IPE) internal events model that includes a Level 21 analysis of core damage scenarios and subsequent containment response resulting in various fission product release categories (including no release). The release category end states from the FitzPatrick Level 2 model are also applied to align with those used by the NRC in NUREG/CR-4551 for Peach Bottom Unit 2. This categorization allows the population dose information provided in NUREG/CR-4551 (adjusted by estimated changes in population since the publication of that document) to be used as a consequence model to provide an estimate of the person-rem dose per reactor year associated with various scenarios. The change in plant risk is then evaluated based on the potential change in population dose rate (person-remlyr), change in Large Early Release Frequency (LERF), and the change in conditional containment failure probability (CCFP).

In addition to the internal events risk assessment evaluation, the impact associated with extending the Type A test frequency interval was further examined by considering external event hazard or potential containment liner corrosion. The purpose for these additional evaluations was to assess whether there are any unique insights or important quantitative information associated with the explicit consideration of external event hazard or containment liner corrosion in the risk assessment results. The external event hazards or potential containment liner corrosion evaluation was found not to impact any of the above conclusions.

Summary of Results:

The conclusion of the plant internal events risk associated with extending the Type A ILRT interval from ten to fifteen years is as follows.

1) The increase in risk on the total integrated plant risk as measured by person-rem/year increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.56% (0.004 person rem/yr). This value can be considered to be a negligible increase in risk.

2) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10 /yr and increases in LERF below 10-7/yr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 1-in-10 years to 1-in-15 years is 1.09 x 104 /ry. Since Regulatory Guide 1.174 defines very small changes in LERF as below 107/yr, increasing the ILRT interval at FitzPatrick Level 2 -the evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 9 of 11 from the currently allowed one-in-ten years to one-in-fifteen years is non-risk significant from a risk perspective.

3) The change in conditional containment failure probability (CCFP) is calculated to demonstrate the impact on 'defense-in-depth'. For the current ten-year ILRT interval, sequences involving no containment failure or small releases contribute 27.4% to the overall plant risk. Alternatively stated, the contribution of sequences involving containment failure for the ten-year internal is 72.6%.

These numbers are consistent with those documented in the FitzPatrick IPE. For the proposed fifteen-year interval, the contribution of sequences involving containment failure increased to 73.03%. Therefore, ACCFPO., 5 is found to be 0.43%. This signifies a very small increase and represents a negligible change in the FitzPatrick containment defense-in-depth.

Additional risk considerations (external event hazards, potential containment liner corrosion) were also evaluated, with a similar conclusion that the requested test interval extension poses negligible risk. These evaluations are summarized in Attachment 4.

4. CONCLUSION Previous Type A tests confirm that the JAF reactor containment structure exhibits extremely low leakage and represents minimal risk to increased leakage. The risk is minimized by continued Type B and Type C testing, reinforced by Inservice Inspection (ISI) program and Maintenance Rule inspections, by other periodic walkdowns and inspections, and by operating experience with a containment that normally operates at a positive pressure. These, in aggregate, provide continuing confidence in containment integrity.

This experience is supplemented by studies, including a plant specific risk analysis, that conclude that the risk associated with extending the Type A test interval on a one-time basis as requested is negligibly small.

It is therefore concluded that the cost-beneficial, risk informed change represented by this request is prudent and reasonable, and that the requested change involves no significant hazards as further documented in the following section.

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 10 of 11 IV. EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION In accordance with the requirements of 10 CFR 50.92, the enclosed application is judged to involve no significant hazards based upon the following information:

1. Does the change Involve a sIgnificant Increase In the probability or consequences of an accident previously analyzed?

The change does not involve a significant increase in the probability or consequences of an accident previously analyzed.

The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing. The current test interval of ten years, based on past performance, would be extended on a one time basis to fifteen years from the last Type A test. The proposed extension to Type A testing cannot increase the probability of an accident previously evaluated since the containment Type A testing extension is not a modification and the test extension is not of a type that could lead to equipment failure or accident initiation.

The proposed extension to Type A testing does not involve a significant increase in the consequences of an accident since research documented in NUREG-1493 has found that, generically, very few potential containment leakage paths are not identified by Type B and C tests. The NUREG concluded that reducing the Type A (ILRT) testing frequency to one per twenty years was found to lead to an imperceptible increase in risk.

These generic conclusions were confirmed by a plant specific risk analysis performed using the current FitzPatrick Individual Plant Examination (IPE) internal events model.

Testing and inspection programs in place at FitzPatrick also provide a high degree of assurance that the containment will not degrade in a manner detectable only by Type A testing. The last four Type A tests show leakage to be below acceptance criteria, indicating a very leak tight containment. Type B and C testing required by Technical Specifications will identify any containment opening such as valves that would otherwise be detected by the Type A tests. Inspections, including those required by the ASME code and the maintenance rule are performed in order to identify indications of containment degradation that could affect that leak tightness.

These factors in part and in aggregate show that a Type A test extension of up to five years will not represent a significant increase in the consequences of an accident.

2. Does the change create the possibility of a new or different kind of accident from any accident previously analyzed?

The change does not create the possibility of a new or different kind of accident from any accident previously analyzed. The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing. The current test interval of ten years, based on past performance, would be extended on a one time

Attachment 2 to JAFP-03-00108 SAFETY EVALUATION Page 11 of 11 basis to fifteen years from the last Type A test. The proposed extension to Type A testing cannot create the possibility of a new or different type of accident since there are no physical changes being made to the plant and there are no changes to the operation of the plant that could introduce a new failure mode creating an accident or affecting the mitigation of an accident.

3. Does the change Involve a significant reduction In the margin of safety?

The change does not involve a significant reduction in the margin of safety. The proposed revision to Technical Specifications adds a one time extension to the current interval for Type A testing. The current test interval of ten years, based on past performance, would be extended on a one time basis to fifteen years from the last Type A test. The proposed extension to Type A testing will not significantly reduce the margin of safety. The NUREG 1493 generic study of the effects of extending containment leakage testing found that a 20 year extension in Type A leakage testing resulted in an imperceptible increase in risk to the public. NUREG -1493 found that, generically, the design containment leakage rate contributes about 0.1 percent to the individual risk and that the decrease in Type A testing frequency would have a minimal affect on this risk since 95% of the potential leakage paths are detected by Type C testing. This was further confirmed by a plant specific risk assessment using the current FitzPatrick Individual Plant Examination (IPE) internal events model that concluded the risk associated with this change is negligibly small and/or non-risk significant.

V. REFERENCES

1. NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix JX

2. Amendment No. 234 to Facility Operating License No. DPR-59 (TAC No. M95099)

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

4. ANSI/ANS 56.8 - 1994, "Containment System Leakage Testing Requirements"

5. NUREG-1493, 'Performance-Based Containment Leak-Test Programn

6. JAF-RPT-PC-02342, 'James A. Fitzpatrick Nuclear Power Plant Primary Containment Leakage Rate Testing Program Plan'

7. [[::JAF-ISI-0002|JAF-ISI-0002]], "Third Inservice Inspection Interval Inservice Inspection Program"

ATTACHMENT 3 to JAFP-03-0108 MARKED UP TECHNICAL SPECIFICATION PAGES Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT)

Interval Extension Entergy Nuclear Operations, Inc.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.6 Primary Containment Leakage Rate Testing Program This program implements the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50. Appendix J. Option B. as modified by a proved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163. 'Performance-Based Containment Leak-Test Program."

dated September 1995. as modified by the following e"eeptio_. _

Type C testing of valves not isolable from the containment free excptins:I air space may be accomplished by pressurization In the reverse direction, provided that testing in this manner provides equivalent or more conservative results than testing in the accident direction. If potential atmospheric leakage paths (e.g..

valve stem packing) are not subjected to test pressure, the portions of the valve not exposed to test pressure shall be subjected to leakage rate measurement during regularly scheduled Type A testing. A list of these valves. the leakage rate measurement method, and the acceptance criteria, shall be contained in the Program.

a. The peak primary containment internal pressure for the design basis loss of coolant accident. Pa. is 45 psig.

b. The maximum allowable primary containment leakage rate. L.,

at Pa, shall be 1.5% of containment air weight per day.

c. The leakage rate acceptance criteria are:

1. Primary containment leakage rate acceptance criteria is 1.0 L,. During plant startup following testing in cd accordance with this program, the leakage rate acceptance criteria are s 0.60 La for the Type B and Type C tests, and i 0.75 La for the Type A tests.

2. Air lock testing acceptance criteria are:

(a) Overall air lock leakage rate is i 0.05 L,when tested at > P.; and (b) For each door seal, leakage rate is s 120 scfd when tested at 2 Pa.

(continued)

JAFNPP 5.5-5 Amendment e-&

Insert A NEI 94-01-1995. Section 9.2.3: The first Type A test performed after the March 7. 1995 Type A test shall be performed no later than March 7.

2010.

ATTACHMENT 4 to JAFP-03-0108 PLANT SPECIFIC RISK ASSESSMENT REPORT Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT)

Interval Extension Entergy Nuclear Operations, Inc.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

Engineering Report No. JAF-RPT 03 00007 Rev. 0 Page 1 Of_ 80 MEnteW ENTERGY NUCLEAR NORTHEAST Engineering Report Cover Sheet Engineering Report

Title:

Risk Impact Assessment of Extending Containment Type A Test Interval Engineering Report Type:

New 0 Revision El Cancelled El Superceded E Applicable Site(s)

IPi Cl IP2 El IP3 0 JAF [9 El E]

PNPS VY Quality-Related: 0 Yes El No Prepared by: John Favara / Kou-John Hong Date: 6-10-03 Responsible Engineer Verified/

Reviewed by: John Bretti Date: 6-10-03 Design Verifier/Reviewer Approved by: Clem Yeh Date: 6-11-03 Supervisor I Multiple Site Review Site Design Verifier/Reviewer I Supervisor Date a .4 .4 I .4 t I .4 4.

I 4 4.

1 .4 4

RECORD OF REVISIONS Engineering Report No: JAF-RPT-03-00007 Page 2of 80 Revision No. Description of Change Reason For Change 0 Original report NA

TABLE OF CONTENTS Description Page Executive Summary 7 Nomenclature 15 Definitions 17

1. Introduction 19 1.1 Purpose 19 1.2 Background 19

2. Evaluation 22 2.1 Method of Analysis 22 2.2 Assumptions 24 2.3 Data and Design Criteria 25 2.4 Internal Events Impact 29 2.4.1 Quantify Baseline Accident Classes Frequencies (Step 1) 29 2.4.2 Containment Leakage Rates (Step 2) 32 2.4.3 Baseline Population Dose Estimate (Step 3) 32 2.4.4 Baseline Population Dose Rate Estimate (Step 4) 40 2.4.5 Change in Probability of Detectable Leakage (Step 5) 41 2.4.6 Population Dose Rate for New ILRT Interval (Step 6) 44 2.4.7 Change in Population Dose Rate Due to New ILRT Interval (Step 7) 46 2.4.8 Change in LERF Due to New ILRT Interval (Step 8) 47 2.4.9 Impact on Conditional Containment Failure Probability (Step 9) 48 2.6 External Events Impact 50 2.6 Containment Liner Corrosion Risk Impact 51

3. Summary of Results 65 3.1 Internal Events Impact 65 3.2 External Events Impact 66 3.3 Containment Liner Corrosion Risk Impact 66

4. Conclusions 72 4.1 Internal Events Impact 72 4.2 External Events Impact 72 4.3 Containment Liner Corrosion Risk Impact 73 5 References 78 Appendix A External Event Assessment During an Extension of the ILRT Interval Appendix B Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval Attachment A FitzPatrick IPE Level 2 Model Results to Support One-Time Exemption

TABLE OF CONTENTS (continued)

Attachment B FitzPatrick Risk Impact of Containment Liner cOrrosion During an Extension of the ILRT Interval Results LIST OF FIGURES Page Figure 2-1 Acceptance Guidelines for Large Early Release Frequency 53

LIST OF TABLES Pam Table ES-I Internal Events Quantitative Results as a Function of ILRT Interval 11 Table ES-2 Internal and External Events Quantitative Results as a Function of ILRT Interval 12 Table ES-3 Liner Corrosion Impact Quantitative Results as a Function of ILRT Interval 13 Table 2-1 Internal Core Damage Frequency Contributions by Plant Damage States 54 Table 2-2 Summary of FitzPatrick IPE LEVEL 2 Release Categories 55 Table 2-3 Summary of FitzPatrick IPE LEVEL 2 Containment Failures 55 Table 2-4 Collapsed Accident Progression Bins (APB) Descriptions 56 Table 2-5 Summary of FitzPatrick Baseline Release Frequencies -Given EPRI TR-104285 Accident Class 58 Table 2-6 Peach Bottom Unit 2 NUREG/CR-4551 Collapsed Accident Progression Bin 50-Mile Population Dose 59 Table 2-7 FitzPatrick Population Doses for the Each APB and Associated Adjustment Factors 59 Table 2-8 FitzPatrick Collapsed Accident Progression Bins Frequencies 60 Table 2-9 FitzPatrick Level 2 Endstates Correlation with the Peach Bottom Unit 2 Collapsed Accident Progression Bins From NUEG/CR-4551 62 Table 2-1 0 FitzPatrick Population Dose Estimates as a Function of EPRI Accident Class Within 50-Mile Radius 62 Table 2-11 FitzPatrick Dose Rates Estimates as a Function of EPRI Accident Class For Population within 50-Miles (Base Line 3 per 10 year ILRT) 63 Table 2-12 EPRI Accident Class Frequency as a Function of ILRT Interval 63 Table 2-13 Baseline Dose Rate Estimates By EPRI Accident Class for Population Within 50Mile 64 Table 3-1 Summary of Risk Impact on Extending Type A ILRT Test Frequency

- Internal Events 68 Table 3-2 Summary of Risk Impact on Extending Type A ILRT Test Frequency

- External Events 69

LIST OF TABLES Page Table 3-3 Summary of Risk Impact on Extending Type A ILRT Test Frequency

- Impact of Containment Steel Liner Corrosion on FitzPatrick ILRT Intervals 70 Table 3-4 Containment Steel Liner Corrosion Sensitivity Cases 71 Table 4-1 Quantitative Results as a Function of ILRT Interval - Internal Events 74 Table 4-2 Quantitative Results as a Function of ILRT Interval - Internal and External Events 75 Table 4-3 Quantitative Results as a Function of ILRT Interval - Liner Corrosion Impact 76

EXECUTIVE

SUMMARY

Revisions to 10CFR 50, Appendix J allow individual plants to extend Type A surveillance testing requirements from three-in-ten years to at least once per ten years. The revised Type A test 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 or 1.0La. James A. FitzPatrick (FitzPatrick) selected the revised requirements as its testing program.

FitzPatrick's current ten-year Type A test frequency is due to be performed during the upcoming refueling outage 16 (RF16), scheduled for October 4, 2004. Prior to the performance of that test, however, FitzPatrick is seeking an extension of the test interval to fifteen years. A substantial cost savings will be realized and unnecessary personnel radiation exposure will be avoided by deferring the Type A test for an additional five years. Cost savings have been estimated for this outage at approximately $660,000.00, which includes labor, equipment and critical path outage time needed to perform the test. In addition, this initiative directly supports site goals related to capacity factor and World Association of Nuclear Operators (WANO) performance by shortening planned outage duration for RO-16.

An evaluation was performed to assess the risk impact of extending the current containment Type A Integrated Leak Rate Test (ILRT) interval. In performing the risk assessment evaluation, the guidelines of NEI 94.01, 'Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J", the methodology used in EPRI TR-104285, ""Risk Assessment of Revised Containment Leak Rate Testing Intervals," and the NRC Regulatory Guide 1.174, mAn Approach for Using Probabilistic Risk Assessment In Risk-informed Decisions On Plant-Specific Changes to the Licensing Basis" were used. The assessment also followed the guidance and additional information distributed by NEI in November 2001 to their Administrative Points of Contact regarding risk assessment evaluation of one-time extensions of containment ILRT intervals and the approach outlined in the Indian Point Unit Three Nuclear Power Plant ILRT extension submittal.

The risk assessment evaluation uses the current FitzPatrick Individual Plant Examination (IPE) internal events model that includes a Level 21 analysis of core damage scenarios and subsequent containment response resulting in various fission product release categories (including no release). The release category end states from the FitzPatrick Level 2 model are also applied to align with those used by the NRC in NUREG/CR-4551 for Peach Bottom Unit 2. This categorization allows the population dose information provided in NUREG/CR-4551 (adjusted by estimated changes in population since the publication of that document) to be used as a consequence model to provide an estimate of the person-rem dose per reactor year associated with various scenarios. The change in plant risk is then evaluated based on the potential change in population dose rate (person-rem/ry), change in Large Early Release Frequency (LERF), and the change in conditional containment failure probability (CCFP).

The risk assessment evaluation examined FitzPatrick's IPE plant specific accident sequences in which the containment integrity remains intact or the containment is impaired. Specifically, the following were considered:

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

1 Level 2 - the evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment

• Core damage sequences in which containment integrity is impaired due to a pre-existing isolation failure of plant components associated with Type A integrated leak rate testing. For example, containment liner breach. (EPRI Class 3 sequences).

• Core damage sequences in which containment integrity is impaired due to pre-existing 'failure-to-seal' failure of plant components associated with either a Type B or Type C local leak rate testing (EPRI Classes 4 and 5 sequences).

• Core damage sequences involving containment isolation failures due to failures-to-close of large containment isolation valves initiated by support system failures, or random or common cause valve failures (EPRI Class 2 sequences) and containment isolation failures of pathways left 'opened' following a plant post-maintenance test, or valve failing to close following a valve stroke test (EPRI Class 6 sequences).

• Core damage sequences involving containment failure induced by severe accident phenomena (EPRI Class 7 sequences) or containment bypassed (EPRI Class 8 sequences).

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

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

2) Determine the containment leakage rates for applicable cases, 3a and 3b.

3) Develop the baseline population dose (person-rem) for the applicable EPRI classes.

4) Determine the population dose rate; also know as population dose risk (person-rem/Ry) 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 (Classes 3a and 3b).

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) Eveluate the change in conditional containment failure probability.

The risk assessment evaluation of the one time ILRT extension is characterized by the following risk metrics: (as used in previously approved ILRT test interval extensions:

• The potential change in population dose rate (person-rem/ry)

• The change in Large Early Release Frequency (LERF)

• The change in conditional containment failure probability (CCFP).

The impact of these risk metrics associated with extending the Type A ILRT interval, are presented in Table ES-1.

The conclusion of the plant internal events risk associated with extending the Type A ILRT interval from ten to fifteen years is as follows.

1) The increase in risk on the total integrated plant risk as measured by person-remlry increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.56% (0.004 person-remlry). This value can be considered to be a negligible increase in risk.

2) Regulatory Guide 1.174 provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10 /yr and increases in LERF below 10- 7 Iyr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 1-in-10 years to 1-in-15 years is 1.09 x 10elry. Since Regulatory Guide 1.174 defines very small changes in LERF as below 10-7/yr, increasing the ILRT interval at FitzPatrick from the currently allowed one-in-ten years to one-in-fifteen years is non-risk significant from a risk perspective.

3) The change in conditional containment failure probability (CCFP) is calculated to demonstrate the impact on 'defense-in-depth'. For the current ten-year ILRT interval, sequences involving no containment failure or small releases contribute 27.4% to the overall plant risk. Alternatively stated, the contribution of sequences involving containment failure for the ten-year interval is 72.6%. These numbers are consisted with those documented in the FitzPatrick IPE. For the proposed fifteen-year interval, the contribution of sequences involving containment failure increased to 73.03%. Therefore, ACCFPO. 15 is found to be 0.43%. This signifies a very small increase and represents a negligible change in the FitzPatrick containment defense-in-depth.

In addition to the internal events risk assessment evaluation, the impact associated with extending the Type A test frequency interval is further examined by considering external event hazard or potential containment liner corrosion. The purpose for these additional evaluations is to assess whether there are any unique insights or important quantitative information associated with the explicit consideration of external event hazard or containment liner corrosion in the risk assessment results.

The external event hazards or potential containment liner corrosion evaluation was found not to impact any of the above conclusions. The results from these cases are presented in Tables ES-2 and ES-3 respectively and summarized below.

Considerations of the combined internal events and external event hazards assessment during an extension of the ILRT Interval yielded the following conclusions:

1) Based on conservative methodologies in estimating the combined core damage frequency for internal events, seismic events, and fires events, the increase in LERF from extending the FitzPatrick ILRT frequency from 1-in-10 years to 1-in-15 years is 1.03 x 107 /ry. This value is slightly above the le/yr criterion of Region 1II, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [6]. Therefore, increasing the ILRT interval at FitzPatrick from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.

2) The combined internal and external events increase in risk on the total integrated plant risk as measured by person-rem/ry increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.16% (0.038 person-rem/ry). This value can be considered to be a negligible increase in risk.

3) The change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-15 years is 0.20%. A change in ACCFP of less than 1% is insignilicant from a risk perspective.

4) Other salient results are summarized in Table ES-2. The key results to this risk assessment are those for the 10-year interval (current FitzPatrick ILRT interval) and the 15-year interval (proposed change).

Recently, the NRC issued a series of Requests for Additional Information (RAls) in response to the one-time relief requests for the ILRT surveillance interval submitted by various licensees. The RAls requested a risk analysis on the potential increase in risk due to drywell/torus liner leakage, caused by age-related degradation mechanisms.

The risk analysis utilizes the referenced Calvert Cliffs Nuclear Power Plant assessment [24] to estimate the risk impact from containment liner corrosion during an extension of the ILRT interval. Consistent with the Calvert Cliffs analysis, the following issues were addressed:

• Differences between the containment basemat and the drywell and torus liner

• The historical drywell/torus steel shell flaw likelihood due to concealed corrosion

• The impact of aging

• The corrosion leakage dependency on containment pressure

• The likelihood that visual inspections will be effective at detecting a flaw Considerations of risk impact of containment liner corrosion during an extension of the ILRT Interval yielded the following conclusions:

1) The impact of including age-adjusted corrosion effects in the ILRT assessment has minimal impact on plant risk and is therefore acceptable.

2) The change in LERF, taking into consideration the likelihood of a containment liner flaw due to age-adjusted corrosion is non-risk significant from a risk perspective. Specifically, extending the interval to 15 years from the current 10 years requirement is estimated to be about 1.22 x 10WIry. This is below the Regulatory Guide 1.174 [6] acceptance criteria threshold of 107 /yr.

3) The age-adjusted corrosion impact in dose increase is estimated to be 4.1 x 103 person-rem/ry or 0.57% from the baseline ILRT 10 years interval.

4) The age-adjusted corrosion impact on the conditional containment failure probability increase is estimated to be 0.6%.

5) A series of parametric sensitivity studies regarding potential age related corrosion effects on the containment steel liner also demonstrated minimal impact on plant risk.

6) Other salient results are summarized in Table ES-3.

DrrLIk REPORT NO. JAF-RPT-03-00007 l Revision 0 lPage I 11 I Of I 80 I a

Table ES-1 Internal Events Quantitative Results as a Function of ILRT Interval Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1-per-15 year ILRT)

Dose Population Dose Population Dose (Person-Rem Accident Rate (Person- Accident Rate (Person-EPRI Within 50 Frequency Rem / Ry Within Frequency Rem / Ry Within Class Category Description miles)° (per ry) 50 miles) (per ry) 50 miles)

I No Containment Failure° 2.91 x 10 3

4.49 x 10'7 1.31 x 1043 3.29 x 10-7 9.58 x 104 2 Containment Isolation System Failure 6.79 x 105 2.44 x 10-9 1.66 x 104 2.44 x 10-9 1.66 x 10-3 3a Small Pre-Existing Failures (2l (3) 2.91 x 104 2.20 x 10-7 6.40 x 10' 3.29 x 10'7 9.60 x 10-3 3b Large Pre-Existing Failures (2), (3) 1.02 x I05 2.20 x 104 2.24 x 104 3.29 x 104 3.36 x 104 4 Type B Failures (LLRT) N/A 0.00 0.00 0.00 0.00 5 Type C Failures (LLRT) N/A 0.00 0.00 0.00 0.00 6 Other Containment Isolation System Failure N/A 0.00 0.00 0.00 0.00 7a Containment Failure Due to Severe Accident (a)(4) 6.11 x 10 1.74 x 10-'7 1.06 x 101' 1.74 x 10-7 1.06 x 1 '

7b Containment Failure Due to Severe Accident (b)(4) 7 3.96 x 105 7.37 x 10 2.91 x I0O' 7.37 x 10 ' 2.91 x 10.1 7c Containment Failure Due to Severe Accident (c) (4) 5 4 5.21 x 10 5.82 x 1i0 3.04 x 10-2 5.82 x 10.8 3.04 x 10-2 7d Containment Failure Due to Severe Accident (d) (4) 5 3.06 x 10 4.83 x 10-7 1.48 x 10-1 4.83 x 10-7 1.48 x 10-8 Containment Bypass Accidents 4.46 x 10'5 2.97 x 10-7 1.32 x 10. 2.97 x 107 1.32 x 10-'

TOTALS: 2.44 x 10 4

_ 0.720 2.44 x o0 4 0.724 Increase In Dose Rate 0.56%

Increase in LERF 1.09x 10.

Increase in CCFP (%) 0.43%

Table ES-2 Internal and External Events Quantitative Results as a Function of ILRT Interval Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1-per-1 5 year ILRT)

Dose Population Dose Population Dose (Person-Rem Accident Rate (Person- Accident Rate (Person-EPRI Within 50 Frequency Rem / Ry Within Frequency Rem / Ry Within Class Category Description miles)" (per ry) 50 miles) (per ry) 50 miles) 1 No Containment Failure 2) 2.91 x 103 2.69 x I 04 7.85 x I 0-3 1.56 x 104 4.55 x 10'3 2 Containment Isolation System Failure 6.79 x 105 5.19 x 10-8 3.53 x 10-2 5.19 x 10-8 3.53 x 10-2 3a Small Pre-Existing Failures 4

( (3) 2.91 x 10 2.06 x 106 6.00 X 10-2 3.09 x 104 9.00 X 10-2 3b Large Pre-Existing Failures (2) (3) 1.02 x 105 2.06 x 10-' 2.10 x 10-2 3.09 x 10-7 3.15 x 10.2 4 Type B Failures (LLRT) N/A 0.00 0.00 0.00 0.00 5 Type C Failures (LLRT) N/A 0.00 0.00 0.00 0.00 6 Other Containment Isolation System Failure N/A 0.00 0.00 0.00 0.00 7a Containment Failure Due to Severe Accident (a)(4) 6.11 x 105 2.45 x 105 1.50 x 101 2.45 x 10-5 1.50 x 101 7b Containment Failure Due to Severe Accident (b)(4) 3.96 x le0 1.57 x 1095 6.20 1.57 x 10-5 6.20 7c Containment Failure Due to Severe Accident (c)(41 5.21 x 1.24x 10- 646x 101 1.24 x 104 6.46 x 10.1 7d Containm ent Failure Due to Severe Accident (d)(4) 3.06 x 105 1.82 x .0 4 5.58 x 10 - 1.82 x 104 5.58 x 10-8 Containment Bypass Accidents 4.46 x 105 3.65 x 10o- 1.63 3.65 x 104 1.63 TOTALS: 5.19 x10 5 24.13 5.19 x 10 5 24.17 Increase In Dose Rate _ _ _ _ 0.16%

Increase In LERF 1.03x 7 Increase in CCFP (%) 0.198%

Table ES-3 Liner Corrosion Impact Quantitative Results as a Function of ILRT Interval Quantitative Resuits as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1-per-15 year ILRT)

Dose Population Dose Population Dose (Person-Rem Accident Rate (Person- Accident Rate (Person-EPRI Within 50 Frequency Rem / Ry Within Frequency Rem / Ry Within Class Category Description miles)( 1 ) (per ry) 50 miles) (per ry) 50 miles)

I No Containment Failure ) 2.91 x 10 3

4.49 x 10'7 1.31 x 10-3 3.26 x 10q7 9.52 x 10'4 2 Containment Isolation System Failure 6.79 x 105 9 3 2.44 x 10- 1.66 x 10- 2.44 x 10'9 1.66 x 10'3 3a Small Pre-Existing Failures (21(3) 2.91 x 104 2.20 x 10-7 6.40 x 10'3 3.29 x 10-7 9.60 x 104 3b Large Pre-Existing Failures (2) (3) 1.02 x 105 2.29 x 10o 2.34 x 10-3 3.51 x 10-8 3.58 x 10-3 4 Type B Failures (LLRT) N/A 0.00 0.00 0.00 0.00 5 Type C Failures (LLRT) N/A 0.00 0.00 0.00 0.00 6 Other Containment Isolation System Failure N/A 0.00 0.00 0.00 0.00 7a Containment Failure Due to Severe Accident (a)(4) 6.11 x 105 1.74 x 107 1.06 x 10.1 1.74 x 10c7 1.06 x 10.1 7b Containment Failure Due to Severe Accident (b)(4) 3.96 x 105 7.37 x 10U7 2.91 x 10-' 7.37 x 10 7 2.91 x 10-'

7c Containment Failure Due to Severe Accident (c) (4) 5.21 x 105 5.82 x 104o 3.04 x 10-2 5.82 x 10.8 3.04 x 10-2 7d Containment Failure Due to Severe Accident (d)(4) 3.06 XI105 4.83 x 10-7 1 1.48 x 10" 4.83 x 0-7 1.48 x 10.1 8 Containment Bypass Accidents 4.46 x 105 2.97 x l0e 1.32 x 101' 2.97 x 10 7 1.32 x 10-.

TOTALS: 2.44 x le" 0.720 2.44 x 106 0.724 Increase in Dose Rate __ ____^______-0.57%

Increase In LERF 122 x 104-Increase in CCFP(%) 0.60% _______r______

Notes to Tables ES-1. Es-2. and ES-3:

1) The population dose associated with the Technical Specification Leakage is based on scaling the population data, the power level, and allowable Technical Specification leakage compared to the Peach Bottom Unit NUREG/CR-4551 reference plant.

2) Only EPRI classes 1, 3a, and 3b are affected by ILRT (Type A) interval changes.

3) Dose estimates for EPRI Class 3a and 3b, per the NEI Interim Guidance, are calculated as 10 times EPRI Class 1 dose and 35 times EPRI Class I dose, respectively.

4) EPRI Class 7, containment failure due to severe accident, was subdivided into four subgroups based on FitzPatrick Level 2 containment failure modes for dose allocation purposes. Note that this EPRI class is not affected by ILRT interval changes.

Nomenclature APB Accident Progression Bin ATWS Anticipated Transient Without Scram CAPB Collapsed Accident Progression Bin CCls Core-Concrete Interactions CCFP Conditional Containment Failure Probability CD Core Damage CDF Core Damage Frequency CDFM Conservative Deterministic Failure Margin CET Containment Event Tree CF Containment Failure DCH Direct Containment Heating DW Drywell EPRI Electrical Power Research Institute EVNTRE Event Progress Analysis Code HCLPF High Confidence Low Probability of Failure ILRT Integrated Leak Rate Testing IPE Individual Plant Examination IPEEE Individual Plant Examination for External Events ISLOCA Interface System Loss of Coolant Accident IP3 Indian Point Unit Three Nuclear Power Plant JAF James A. FitzPatrick Nuclear Power Plant LERF Large Early Release Frequency LLRT Local Leak Rate Testing LOCA Loss of Coolant Accident NEI Nuclear Energy Institute

Nomenclature (continued)

NRC United States Nuclear Regulatory Commission MFCR Mean Factional Contribution to Risk PDS Plant Damage State PRA Probabilistic Risk Analysis PSA Probabilistic Safety Assessment RAI Request for Additional Information RCS Reactor Coolant System RPV Reactor Pressure Vessel RF Refueling Outage SCDF Seismic Core Damage Frequency SMA Seismic Margin Assessment TS Technical Specifications WANO World Association of Nuclear Operations vWv Wetwell

Definitions Accident sequence - a representation in terms of an initiating event followed by a combination of system, function and operator failures or successes, of an accident that can lead to undesired consequences, with a specified end state (e.g., core damage or large early release). An accident sequence may contain many unique variations of events (minimal cut sets) that are similar.

Core damage - uncovery and heatup of the reactor core to the point at which prolonged oxidation and severe fuel damage is anticipated and involving enough of the core to cause a significant release.

Core damage frequency - expected number of core damage events per unit of time.

Cutsets - Accident sequence failure combinations.

End State - is the set of conditions at the end of an event sequence that characterizes the impact of the sequence on the plant or the environment. End states typically include: success states, core damage sequences, plant damage states for Level 1 sequences, and release categories for Level 2 sequences.

Event tree - a quantifiable, logical network that begins with an initiating event or condition and progresses through a series of branches that represent expected system or operator performance that either succeeds or fails and arrives at either a successful or failed end state.

Initiating Event - An initiating event is any event that perturbs the steady state operation of the plant, if operating, or the steady state operation of the decay heat removal systems during shutdown operations such that a transient is initiated in the plant. Initiating events trigger sequences of events that challenge the plant control and safety systems.

ISLOCA - a LOCA when a breach occurs in a system that interfaces with the RCS, where isolation between the breached system and the RCS fails. An ISLOCA is usually characterized by the over-pressurization of a low-pressure system when subjected to RCS pressure and can result in containment bypass Large early release - the rapid, unmitigated release of airborne fission products from the containment to the environment occurring before the effective implementation of off-site emergency response and protective actions.

Large early release frequency - expected number of large early releases per unit of time.

Level 1 - identification and quantification of the sequences of events leading to the onset of core damage.

Level 2 - evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment.

Plantdamage state - Plant damage states are collections of accident sequence end states according to plant conditions at the onset of severe core damage. The plant conditions considered are those that determine the capability of the containment to cope with a severe core damage accident. The plant damage states represent the interface between the Level 1 and Level 2 analyses.

Probability- is a numerical measure of a state of knowledge, a degree of belief, or a state of confidence about the outcome of an event.

Definitions (continued)

Probabilisticrisk assessment- a qualitative and quantitative assessment of the risk associated with plant operation and maintenance that is measured in terms of frequency of occurrence of risk metrics, such as core damage or a radioactive material release and its effects on the health of the public (also referred to as a probabilistic safety assessment, PSA).

Release category- radiological source term for a given accident sequence that consists of the release fractions for various radionuclide groups (presented as fractions of initial core inventory), and the timing, elevation, and energy of release. The factors addressed in the definition of the release categories include the response of the containment structure, timing, and mode of containment failure; timing, magnitude, and mix of any releases of radioactive material; thermal energy of release; and key factors affecting deposition and filtration of radionuclides. Release categories can be considered the end states of the Level 2 portion of a PSA.

Risk - encompasses what can happen (scenario), its likelihood (probability), and its level of damage (consequences).

Risk metrics- the quantitative value, obtained from a PRA analysis, used to evaluate the results of an application (e.g., CDF or LERF).

Severe accident - an accident that involves extensive core damage and fission product release into the reactor vessel and containment, with potential release to the environment.

Split Fraction- a unitless parameter (i.e., probability) used in quantifying an event tree. It represents the fraction of the time that each possible outcome, or branch, of a particular top event may be expected to occur. Split fractions are, in general, conditional on precursor events. At any branch point, the sum of all the split fractions representing possible outcomes should be unity. (Popular usage equates "split fraction" with the failure probability at any branch [a node] in the event tree.)

Vessel Breach - a failure of the reactor vessel occurring during core melt (e.g., at a penetration or due to thermal attack of the vessel bottom head or wall by molten core debris).

SECTION 1 INTRODUCTION 1.1 Purpose The purpose of this report is to provide supplemental information to support the proposed James A.

FitzPatrick (FitzPatrick) Technical Specification change of implementing a one-time extension of the containment Type A integrated leak rate test (ILRT) interval from ten years to fifteen years.

The risk assessment follows the guidelines from NEI 94-01 "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J" [1], the methodology used in EPRI TR-104285 "Risk Assessment of Revised Containment Leak Rate Testing Intervals' [2], NEI's "Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals' 13], NEI 's 'Additional Information for ILRT Extensions" [4], and the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) findings and risk insights in support of a request for a change in a plant's licensing basis as outlined in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-nformed Decisions On Plant-Specific Changes to the Licensing Basis" [5].

In addition, the results and findings from the FitzPatrick Individual Plant Examination (IPE) update [6] are used for this risk assessment report.

1.2 Background In October 26, 1995, the Nuclear Regulatory Commission (NRC) revised 10 CFR 50, Appendix J. The revision to Appendix J provided a performance based option, Option B "Performance-Based Requirements", for leakage-rate testing of light-water-cooled containments.

Under Option B, the Integrated Leak Rate Testing (ILRT) Type A surveillance testing requirements was extended from three-in-ten years to at least once per ten years. The revised Type A test 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 is less than the maximum allowable containment leakage limit of 1.0L8.

In accordance with the revised containment leakage-rate testing for Appendix J, FitzPatrick selected the requirements under Option B as its testing program. FitzPatrick's current ten-year Type A test is due to be performed during refueling outage sixteen (RF16, scheduled for October 4, 2004). However, FitzPatrick seeks a one-time exemption based on the substantial cost savings of $660,000.00 from extending the test from the RF16 schedule to RF18. In addition, this initiative directly supports site goals related to capacity factor and World Association of Nuclear Operators (WANO) performance by shortening planned outage duration for RO-16.

The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, which was established in 1995 during development of the performance-based Option B to Appendix J [1].

Section 11.0 of NEI 94-01 states that NUREG-1493, "Performance-Based Containment Leak-Test Program," [7] dated September 1995, provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J.

The NUREG-1493 [] report examined the impact of containment leakage on public health and safety associated with a range of extended leakage rate test intervals. The NUREG analyzed both Boiling Water Reactors (Peach Bottom and Grand Gulf) and Pressurized Water Reactors (Surry, Sequoyah, and Zion). For Peach Bottom, (a comparable Boiling Water Reactor plant to FitzPatrick's), it was found that increasing the containment leak rates several orders of magnitude over the design basis (0.5 percent per day to 50 percent per day), results in a negligible increase in total population exposure. Therefore, extending the ILRT interval does not result in any significant increase in risk.

To supplement the NRC's rulemaking basis, NEI undertook another similar study. The results of that study are documented in EPRI research project report TR-104285 [2]. The EPRI Methodology [2] used a simplified risk model-PRA containment event trees (CETs). These CETs provide a risk framework for evaluating the effect of containment isolation failures affected by leakage testing requirements. The complexity of the CET models however is not necessary to evaluate the impact of containment isolation system failures. Therefore, a simplified risk model was developed to distinguish between those accident sequences that are affected by the status of the containment isolation system versus those that are a direct function of severe accident phenomena. The simplified risk model allowed for a smaller number of CET scenarios to be evaluated to determine the baseline risk as well as subsequent analysis to quantify risk effects of extending test intervals. The methodology regrouped core damage accident sequences reported in PRAs reviewed in the study into eight classifications to permit the appropriate delineation among containment isolation failure and containment failure due severe accident phenomena. The eight EPRI accident classes in the simplified model are:

1) Containment remains intact initially and in the long term. The release of fission products (and accident consequences) is determined by the maximum allowable containment leakage.

2) Core damage accident sequences in which containment integrity is impaired due independent (or random) containment isolation failures that include those accident s sequences in which the containment isolation system function fails during the accident progression (i.e., failures-to-dose of large containment isolation valves initiated by support system failures, or random or common cause valve failures).

3) Core damage sequences in which containment integrity is impaired due to a pre-existing isolation failure of plant components associated with Type A integrated leak rate testing. For example, containment liner breach.

4) Core damage sequences in which containment integrity is impaired due to an independent (or random) pre-existing isolation failure-to-seal of plant components associated with Type B integrated leak rate testing. These are the Type B-tested components that have isolated but exhibit excessive leakage.

5) Core damage sequences in which containment integrity is impaired due to an independent (or random) pre-existing isolation failure-to-seal of plant components associated with Type C integrated leak rate testing.

6) Core damage sequences in which containment integrity is impaired due to containment isolation failures that include those leak paths not identified by containment leak rate tests. The type of failures considered under this Class includes those valves left open or valves that did not properly seal following test or maintenance activities.

7) Core damage sequences involving containment failure induced by severe accident phenomena.

Changes in ILRTs or LLRTs requirements do not impact these accidents.

8) Core damage sequences in which the containment is bypassed (either as an initial condition or induced by accident phenomena). Changes in ILRTs or LLRTs requirements do not impact these accidents.

Building upon the methodology of the EPRI TR-104285 [2] study, the Indian Point Unit Three (IP3)

Methodology [8], quantified leakage from accident sequences in endstate 3 (reclassified as 3a and 3b).

Accident sequence endstates 3a and 3b have the potential to result in a change in risk associated with changes in ILRT intervals since a pre-existing leak is assumed to be present for these endstates. By manipulating the probability of a pre-existing leak of sufficient leak size, an evaluation of the change in large early release frequency (LERF) can be performed. The NRC [9] considered this an improvement on the EPRI study. Similar information is contained in the Crystal River Nuclear Power Plant submittal [10].

Based on the improved methodology, NEI issued in November 2001 enhanced guidance "Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals' 13], and Additional Information for ILRT Extensions," [4] that builds on the EPRI TR-104285 [2], IP3 [8] and Crystal River submittal [10]

methodology and is intended to provide for more consistent submittals to the NRC.

The FitzPatrick evaluation assesses the change in the predicted population dose rate associated with the interval extension. The assessment also evaluated the risk increase resulting from extending the ILRT interval in terms of Large Early Release Frequency (LERF), and the impact on Conditional Containment Failure Probability (CCFP). Regulatory Guide 1.174 [5] provides guidance for using PRA in risk-informed decisions for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 [5] defines very small changes in the risk acceptance guidelines as increases in Core Damage Frequency (CDF) of less than 104 per reactor year and increases in LERF of less than 10-7 per reactor year. Since the Type A test does not impact CDF, the only relevant criterion is the change in LERF. Regulatory Guide 1.174 [5] also encourages the use of risk analysis techniques to help ensure and demonstrate that key risk metrics such as defense-in-depth philosophy, are satisfied. Based on that, the increase in the CCFP, which helps to ensure that the defense-in-depth philosophy is maintained, was evaluated.

SECTION 2 EVALUATION 2.1 Method of Analysis The FitzPatrick risk assessment analysis uses the approach outlined in the Indian Point Unit Three Nuclear Power Plant (1P3) methodology [8], EPRI's TR-104285 [2], NEI's Interim Guidance (3], NEI 's uAdditional Information for ILRT Extensions' 14], and the NRC regulatory guidance on the use of PRA findings and risk insights in support of a request for a change in a plant's licensing basis as outlined in Regulatory Guide 1.174 [5].

The EPRI TR-104285 methodology [2] involves a quantitative evaluation on the change in public risk of the affect of extending the ILRT and Local Leak Rate Test (LLRT) intervals. The EPRI TR-104285 study combined IPE Level 22 models with NUREG-1 150 "Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plantse [11] Level 33 population dose models to perform the analysis. This study also used the approach of NUREG-1493 [7] in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-1 04285 used a simplified containment event tree (CET) to provide a risk-based framework for evaluating the effects of containment isolation failures impacted by Appendix J penetrations testing requirements. The CET regrouped core damage accident sequences into eight accident classes of containment response. These eight accident classes are:

1) Containment intact and isolated

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

3) Type A (ILRT) related containment isolation failures

4) Type B (LLRT) related containment isolation failures

5) Type C (LLRT) related containment isolation failures

6) Other penetration related containment isolation failures

7) Containment failure due to core damage accident phenomena

8) Containment bypass These eight accident classes allow the isolation failures modes and type of penetration analyzed to be correlated directly with Types A, B, and C test relaxation benefits. Each of the eight classes was categorized according to certain release characterization to determine the baseline incremental risk.

The IP3 methodology [8] modified the EPRI TR-104285 [2] accident class 3 to include the probability of a containment breach (due to excessive leakage) at the time of core damage. Class 3 was redefined as accident sequence endstates 3a (small containment breach) and 3b (large containment breach). This reclassification resulted in the change in risk associated with changes in ILRT intervals (since a pre-existing leak is assumed to be present for these endstates). Furthermore, by changing the probability of a pre-existing leak, an evaluation of the change in large early release frequency (LERF) can be performed. The NRC, 'Indian Point Nuclear Generating Station Unit No. 3 - Issuance of Amendment Re:

Frequency of Performance-Based Leakage Rate Testing, April 17, 2001" [9] considered this an 2 Level 2 - the evaluation of containment response to severe accident challenges and quantification of the mechanisms, amounts, and probabilities of subsequent radioactive material releases from the containment.

3 Level 3 -A measure of containment failure sequences leading to public health effects and their frequencies.

improvement on the EPRI study. Similar information is contained in the Crystal River submittal,

'Supplemental Risk Informed Information in Support of License Amendment Request No. 267" [10].

NEI's Interim Guidance documents [3 and 4] improve on the above methods. Therefore, the FitzPatrick risk assessment analysis uses the approach outlined in the NEI's Interim Guidance [3] and NEI 's

'Additional Information for ILRT Extensions' [4]. The nine steps of the methodology are:

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

2) Determine the containment leakage rates for applicable cases, 3a and 3b.

3) Develop the baseline population dose (person-rem) for the applicable EPRI classes.

4) Determine the population dose rate; also know as population dose risk (person-rem/ry) 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 (Classes 3a and 3b). 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.

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 latest FitzPatrick IPE Level 14 and Level 2 [6] were used to evaluate the change in population dose rate (person-rem/ry), change in Large Early Release Frequency (LERF), and the change in conditional containment failure probability. In order to assess the impact on offsite dose, Peach Bottom information, "Evaluation of Severe Accident Risks: Peach Bottom, Unit 2" [12] was used to estimate the FitzPatrick offsite dose.

The first seven steps of the methodology calculate the change in dose. The change in dose is the primary basis upon which the Type A ILRT interval extension was previously granted for IP3 [8, 9] and other subsequent extensions [10].

The eighth step in the interim methodology calculates the change in LERF and compares it to the guidelines in Regulatory Guide 1.174 [5]. Because the change in ILRT test interval does not impact the CDF, the relevant criterion is LERF. The final step of NEI's interim methodology calculates the change in containment failure probability given the change of ILRT test interval from once-per-10 years to once-per-15 years.

4 Level I - identification and quantification of the sequences of events leading to the onset of core damage.

2.2 Assumptions~

1) The surveillance frequency for Type A testing in NEI 94-01 [1] is at least once per ten years based on an acceptable performance history. Based on the consecutive successful ILRTs performed in the early 1990's, the current ILRT interval for FitzPatrick is once per ten years [13].

2) The FitzPatrick (Revision 1) Level I and Level 2 internal events IPE models provide representative results for the analysis [6].

3) Radionuclide release categories defined in this report are consistent with the EPRI TR-1 04285 methodology. 12]

4) The EPRI methodology concluded that Severe Accident Phenomena and Bypass Classes accident sequences (e.g., drywell liner melt-through, ATWS or Interface system LOCA, ISLOCA) contribution to population dose is unchanged by the proposed ILRT extension. These Classes are included for comparison purposes. As such, no changes in this analysis will alter this conclusion.

5) The reliability of containment isolation valves to close in response to a containment isolation signal is not impact by the change in ILRT frequency.

6) The maximum containment leakage for Class 1 sequences is ILa (21. (La is the Technical Specification maximum allowable containment leakage rate).

7) The maximum containment leakage for Class 3a sequences per the NEI Interim Guidance 13]

and previously approved methodology (8, 9] is 10La.

8) The maximum containment leakage for Class 3b sequences per the NEI Interim Guidance (3]

and previously approved methodology [8, 91 is 35La.

9) Class 3b release is categorized as LERF, based on the previously approved IP3 ILRT extension

[8, 9] and NEI's interim methodology [3].

10) Containment leak rates greater than 2La but less than 35La indicate an impaired containment.

The leak rate is considered 'small' per the NEI Interim Guidance [31 and previously approved methodology (2, 8, and 91. Furthermore, these releases have a break opening of greater than 0.5-inch but less than 2-inch diameter [8, 9].

11) Containment leak rates greater than 35La indicates a containment breach. This leak rate is considered 'large' per the NEI Interim Guidance [3] and previously approved methodology [8, 9].

12) Containment leak rates less than 2La indicates an intact containment This leak rate is considered as 'negligible' per the NEI Interim Guidance [3] and previously approved methodology [8, 9].

13) EPRI accident Class 2 (Large Containment Isolation Failures) potential releases can be consider similar to a release associated with early drywell failure at high reactor pressure vessel (RPV) pressure.

14) Because EPRI Class 8 sequences are containment bypass sequences, potential releases are directly to the environment. Therefore, the containment structure will not impact the release magnitude.

15) An evaluation of the risk impact of the ILRT on shutdown risk is addressed using the generic results from EPRI TR-104285 [2] as augmented by NEI Interim Guidance [3, 41.

16) Although the meteorology data could play a role in the early health effects calculations, the meteorology and site topography for Peach Bottom Unit 2 and FitzPatrick are assumed to be sufficiently similar that any differences are assumed not to play a significant role in this evaluation of total population dose.

2.3 Data and Deslan Criteria

1) The FitzPatrick Level 1 and 2 IPE update is used as input to this analysis reflects the as built, as-operated plant. [61

2) The CDF value, as reported in the Fitzpatrick IPE, Revision 1 is 2.44 x 10ery. [6]

3) The FitzPatrick Level 2 IPE is used to calculate the release frequencies for the accidents evaluated in this assessments. Table 2-1 summarizes the FitzPatrick Level 1 IPE internal mean frequency results by core damage plant damage states (PDS). (See Attachment A)

4) Table 2-2 summarizes the pertinent FitzPatrick Level 2 IPE results in terms of release modes.

The total release frequency is 1.75 x 104 /ry; with a total CDF of 2.44 x I eiry. Table 2-3 summarizes the correlation of the FitzPatrick Level 2 IPE results for containment failure accident progression bins in terms of release magnitudes. (See Attachment A)

5) The random large containment isolation failure probability, from the FitzPatrick IPE, Revision 1, Section 4.5 [61 is = 10-3. This value reflects the mean probability of a solenoid valve failing to close on demand. This value is bounding, since other pathways involve redundant closed pathways with failure probabilities less than 10 6.

6) The conditional failure probability of having a small pre-existing containment leak is 0. 027. This value is based on work performed in the IP3 ILRT submittal [8] and the NEI Interim Guidance (3].

From the IP3 submittal, the probability that a liner leak will be small made use of the data presented in NUREG-1493 [7]. The data reported in NUREG-1493 found that 23 of 144 tests had allowable leak rates in excess of 1.0La. However, of these 23 'failures' only 4 were found by an Type A ILRT, the others were found by Type B and C testing or errors in test alignments.

Therefore, the number of failures considered for 'small releases' are 4-of-144. Recent data collected by NEI and documented in the NEI Interim Guidance [31 found that an additional 38 ILRT have been performed since 1/11/95, with only one failure occurring. This indicates a failure probability of 5/182 (0.027) for a type A ILRT.

7) The conditional failure probability of having a large pre-existing containment leak is 0.0027. This value is derived from the NEI Interim Guidance [3]. It's based on the Jeffreys non-informative prior distribution6 for zero failures. The formula is as follows:

5 The Level 2 analysis used a point estimate CDF of 2.2 x 104Iry. However, this analysis uses the mean CDF value incalculating the eight accident dasses' frequencies.

6 Application of the Jeffreys non-informative prior is one of a number of statistical analysis approaches to estimating probabilities when no failures have been experienced. The approach was used in NUREG-1 150 and more recently inNUREGICR-5750.

NUREGICP-5750 is now the preferred source of initiating event data, which also involves rare event approximations. The selected approach is more conservative than many other statistical approaches.

Number of Failures + 1/2 Number of Tests +1 The number of large failures is zero, so the probability is 0.5/183=0.0027.

8) The total population dose risk for Peach Bottom Unit 2 is taken from NUREG/CR-4551, Table 5.1-1 [12]. This value is 7.9 person-rem per reactor year.

9) The mean internal event core damage frequency for Peach Bottom Unit 2 is taken from NUREG/CR-4551, Figure S-3 [12]. This value is 4.34 x 10 per reactor year.

10) The population dose data in NUREG/CR-4551 for Peach Bottom Unit 2 [12] is reported in ten distinct collapsed accident progression bins (APBs). These collapsed APBs are composed essentially of five characteristics: the occurrence of core damage,7 the occurrence of vessel breach, primary system pressure at vessel breach, the location of containment failure, and the timing of containment failure. A description of these characteristics and their associated attributes are presented in Table 2-4.

11) The internal events mean fractional contribution to risk (MFCR) for each collapsed accident progression bin (APB) for Peach Bottom Unit 2 is taken from NUREG/CR-4551, Table 5.2-3 [12].

These are as follows:

Collapsed APB MFCE 1 0.0210 2 0.0066 3 0.5560 4 0.2260 5 0.0022 6 0.0590 7 0.1180 8 0.0005 9 0.0100 10 0

12) The internal events individual conditional probabilities for each collapsed APB for Peach Bottom Unit 2 is taken from NUREG/CR-4551, Figure S-3 [12]. These are as follows:

Collased APB Individual Conditional Probabilities 0.0220 2 0.0110 3 0.3410 4 0.1830 5 0.0030 6 0.0470 7 0.1100 8 0.1840 9 0.0890 10 0.0100 7 Core damage here implies substantial core melt and relocation inside the reactor pressure vessel.

13) The 50-mile radius Peach Bottom Unit 2 population data used to characterize the population dose calculations is 3.02 x 106 [14]. This value is based on the population data presented in the Peach Bottom Unit 2 reactor risk study presented below.

NUREG/CR-4551 Peach Bottom Unit 2 Population Data at Different Radii From the Plant [12]

Distance From Plant (Km) (Miles) Population 1.6 1.0 118 4.8 3.0 1822 16.1 10.0 28,647 48.3 30.0 989,356 160.9 100.0 14,849,112 563.3 350.0 68,008,584 1609.3 1000.0 154,828,144 The analysis utilizes an estimate of the population density within 50 miles for the Peach Bottom Unit 2 plant. (See Appendix C of Reference 14 for more details). The value of 3.02 x 106 is used to relate that 50-mile population dose calculation from Peach Bottom Unit 2 to FitzPatrick; the population information is needed to appropriately scale the calculated dose from Peach Bottom Unit 2 to FitzPatrick.

14) The 50-mile radius FitzPatrick population data is obtained from a study performed for the Nine Mile Point and James A. FitzPatrick Nuclear Stations 'Ingestion Pathwa5 y Population" [15]. The value used to characterize the population dose calculations is 8.98 x 10 [15]. This value is based on the information contain in Table M-2 [15] and reproduced below.

Nine Mile Point and James A. FitzPatrick Nuclear Stations Population Estimate [15]

MLES ______POPULAIO FROL ZONE rFACILTY 2000 2009 2010 1 0-1 8 8 8 2 1-2 342 343 344 3 2-3 926 030 031 4 3-4 1.454 1.46D 1.461 6 4-5 1.422 1.428 1.429 6 5 4.3261 4.343 4,348 7 6-7 0251 9.288 9.297 6 7-8 T 7X5 10.83 10,49

_9 1 i .155 &1n &181 10 0-10 4.53 4.51 4.576 15 10-15 32.843 32080 33.012 20 15-20 28,433 2.46 28574 25 1 20-25 45.046 44.285 44.097 30 1 25-30 1092 107Z19 107X1 35 M 30-35 16.X3 162M5 162.034 40 3540 219X 5 214A82 21331 45 40-45 131.Z2 13051 130X9 50 4 127.023 12Z4 120,188 TOTAL: 49767 i 78823 8U3440 Vff4#1NMILFS:- 4,1 4160 ti73 ITHN 10 MLES 38.232 38381 38.424 TION IlTHN 25 MLESI 144* 142.118 14,107

15) The Peach Bottom Unit 2 reactor power level used in the NUREG/CR-4551 [12] consequence calculations is 3293 MWth.

16) The Peach Bottom Unit 2 maximum allowed containment leakage used in the NWREG/CR-4551 consequence calculations is 0.5 volume%/day [12].

17) The Peach Bottom Unit 2 containment volume used in the NUREGICR-4551 [12] consequence calculations is 307,000 ft3.

18) The FitzPatrick reactor power level used in this report is the power uprated adjusted value of 2536MWth [16].

19) The FitzPatrick technical specification maximum allowed containment leakage is 1.5 volume%/day [17].

20) The Fitzpatrick containment free volume used in this report is 264,000 ft3 (drywell free volume of approximately 150,000 ft3 and torus free volume of approximately 114,000 ft@) (18].

2.4 Internal Events Impact This section provides a step-by-step summary of the NEI guidance [3] as applied to the James A.

FitzPatrick ILRT interval extension risk assessment. Each subsection addresses a step in the NEI guideline [3].

2A.1 Quantify Baseline Accident Classes Frequencies (Step 1)

This step involves the quantification of the baseline frequencies for each of the EPRI TR-104285 accident classes [2].

Frequency of EPRI Class I Sequences. This group consists of all core damage accident progression sequences in which the containment remains isolated and intact (or containment leakage at or below maximum allowable Technical Specification leakage).

Consistent with NEI Interim Guidance [3], the frequency per reactor year for these sequences is calculated by subtracting the frequencies of EPRI Classes 3a and 3b from the sum of all severe accident progression sequence frequencies in which the containment is isolated and intact:

CLASS 1-FREQUENCY = NCF - CLASS 3a FREQUENCY - CLASS-3b-FREQUENCY Where:

CLASS_-_FREQUENCY = frequency of EPRI Class I given a 3-n-10 years ILRT interval NCF = frequency in which containment leakage is at or below maximum allowable Technical Specification leakage

= 6.91 x 10 71ry [Table 2-21 CLASS_3aFREQUENCY = frequency of small pre-existing containment liner leakage

= 6.59 x 10 /ry [See below write-up]

CLASS_3bFREQUENCY = frequency of large pre-existing containment liner leakage

= 6.59 x 10 1ry [See below write-up]

Therefore:

CLASSIFREQUENCY = 6.91 x I1 7 - 6.59 x 10 -6.59 x 109 CLASS_1.JREQUENCY = 6.19 x 10T7/ry Frequency of EPRI Class 2 Sequences. This group consists of all core damage accident progression bins in which the containment isolation system function fails during the accident progression. These sequences are dominated by failure-to-close of large (>2-inch diameter) containment isolation valves [6].

The frequency per reactor year for these sequences is determined as follows:

CLASS_2 FREQUENCY = PROB heCI

• CDF Where:

CLASS_2_FREQUENCY = frequency of EPRI Class 2 given a 3-in-10 years ILRT interval

PROB ,,I ,C = random large containment isolation failure probability (i.e. large valves)

= 10[ Section 2.3, input#5]

CDF = FitzPatrick PE core damage frequency = 2.44 x 104Iry [Section 2.3, input #2]

Therefore:

CLASS-2 FREQUENCY = 10 *2.44 x 104 CLASS_2_FREQUENCY = 2.44 x 109/ry Freauency of EPRI Class 3a Sequences. This group consists of all core damage accident progression bins for which a small pre-existing leakage in the containment structure (i.e. containment liner) exists.

This type of failure is identifiable only from an ILRT and therefore, affected by a change in ILRT testing frequency.

Consistent with NEI Interim Guidance [3], the frequency per reactor year for this category is calculated as:

CLASS_3aFREQUENCY = PROBd,,, 3a

• CDF Where:

CLASS_3a-FREQUENCY = frequency of EPRI Class 3a given a 3-in-10 years ILRT interval PROBdSIS_3a = probability of small pre-existing containment liner leakage

= 0.027 [Section 2.3, input#6]

CDF = FitzPatrick PE core damage frequency = 2.44 x 104 /ry [Section 2.3, input#2]

Therefore, CLASS 3aFREQUENCY = 0.027

• 2.44 x leO CLASS_3aFREQUENCY = 6.59 x 10Try Freguencv of EPRI Class 3b Seguences. This group consists of all core damage accident progression bins for which a large pre-existing leakage in the containment structure (i.e. containment liner) exists.

This type of failure is identifiable only from an ILRT and therefore, affected by a change in ILRT testing frequency.

Consistent with NEI Interim Guidance (3], the frequency per reactor year for this category is calculated as:

CLASS_3b FREQUENCY = PROBdass 3b XCDF Where:

CLASS_3bFREQUENCY = frequency of EPRI Class 3b given a 3-in-10 years ILRT interval PROBdass 3b = probability of large pre-existing containment liner leakage

= 0.0027 [Section 2.3, input #7]

CDF = FitzPatrick PE core damage frequency = 2.44 x 10 6 ry [Section 2.3, input # 2]

i!~~Ente yJ REPORT No. JAF-RPT-03-00007 Revision 0 l Page I 31 IO, I 80 I Therefore, CLASS_3a FREQUENCY = 0.0027 - 2.44 x 10 CLASS_3aFREQUENCY = 6.59 x 10Iry Frequency of EPRI Class 4 Sequences. This group consists of all core damage accident progression sequences in which the containment isolation system function fails due to a pre-existing failure-to-seal of Type B test component(s). Consistent with NEI Interim Guidance [31, because these failures are detected by Type B tests and not by the Type A ILRT, this group is not evaluated further in this analysis.

Frequency of EPRI Class 5 Sequences. This group consists of all core damage accident progression sequences in which the containment isolation system function fails due to a pre-existing failure-to-seal of Type C test component(s). Consistent with NEI Interim Guidance [3], because these failures are detected by Type C tests, this group is not evaluated any further.

Frequency of EPRI Class 6 Sequences. This group consists of all core damage accident sequences in which the containment isolation function is failed due to 'other' pre-existing failure modes (e.g., pathways left open or misalignment of containment isolation vales following a test/maintenance evolution).

Consistent with NEI Interim Guidance [3], because these failures are detected by Type B or C tests, this group is not evaluated any further.

Frequency of EPRI Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (i.e. liner melt-through).

Consistent with NEI Interim Guidance [3], the frequency per reactor year for this class is based on the plant Level 2 PSA results.

Because the Fitzpatrick IPE Level 2 containment failure results are summarized into four different release bins (Table 2-3), EPRI Class 7 is sub-divided in this report to reflect this sub-division of the FitzPatrick Level IPE results. The following sub-classes are defined:

• Class 7a: severe accident induced early drywell failures resulting in early high magnitude releases.

• Class 7b: severe accident induced early torus failures resulting in early medium high or early medium low releases.

• Class 7c: severe accident induced late drywell failures resulting in late high magnitude releases.

• Class 7d: severe accident induced early torus failures resulting in late medium high or late medium low releases.

The frequency of Category 7a is the total frequency of the FitzPatrick Level 2 IPE early drywell failures release bin. Based on the FitzPatrick Level 2 IPE results summarized earlier in Table 2-3, the frequency of Category 7a is 1.74 x 10 7/ry.

The frequency of Category 7b is the total frequency of the FitzPatrick Level 2 IPE early torus failures release bin. Based on the FitzPatrick Level 2 IPE results summarized earlier in Table 2-3, the frequency of Category 7b is 7.37 x 10 7/ry.

The frequency of Category 7c is the total frequency of the FitzPatrick Level 2 IPE late dryvell failures release bin. Based on the FitzPatrick Level 2 IPE results summarized earlier in Table 2-3, the frequency of Category 7c is 5.82 x 1IOry.

The frequency of Category 7d is the total frequency of the FitzPatrick Level 2 IPE late torus failures release bin. Based on the FitzPatrick Level 2 IPE results summarized earlier in Table 2-3, the frequency of Category 7d is 4.83 x 10 7 /ry.

Frequency of EPRI Class 8 Sequences. This group consists of all core damage accident progression bins in which the accident is initiated by a containment bypass scenario (i.e., ATWS with high power oscillations or Interfacing Systems LOCA). Based on the Fitzpatrick Level I IPE results summarized earlier in Table 2-3, the frequency of Class 8 is 2.97 x 1O7/ry.

Note: for this class the maximum release is not based on the maximum allowable containment leakage, because the releases are released directly to the environment. Therefore, the containment structure will not impact the release magnitude.

The EPRI TR -104285 Class frequencies that result in radionuclide releases to the public are derived in accordance with NEI Interim Guidance [3]. The EPRI TR-104285 Class accident sequence frequency results are summarized in Table 2-5.

2.4.2 Containment Leakage Rates (Step 2)

This step defines the containment leakage rates for EPRI accident Classes 3a and 3b. As defined in Step 1, accident Class 3a and 3b are plant accidents with pre-existing containment leakage pathways (designated as 'small' and "large") that are identifiable only when performing a Type A ILRT.

The NEI Interim Guidance [3] recommends containment leakage rates of 1OLa and 35La for accident Classes 3a and 3B, respectively. These values are consistent with previous ILRT frequency extension submittal applications [81. La is the plant Technical Specification maximum allowable containment leak rate; for FitzPatrick La isl.5% of containment air weight per day (per FitzPatrick Technical Specification).

By definition, and per the NEI Interim Guidance [3] and previously approved methodology [8] the containment leakage rate for Class I (i.e., accidents with containment leakage at or below maximum allowable Technical Specification leakage) is 2 La.

2.4.3 Baseline PopulatIon Dose Estimate (Step 3)

This step estimates the baseline population dose (person-rem) for each of the EPRI TR-104285 accident classes 13]. The NEI Interim Guidance 13] recommends two options for calculating population dose for the EPRI accident classes:

• Use of NUREG-1150 dose calculations [11

• Use of plant-specific dose calculations The use of generic dose information for NUREG-1 150 [11] is recommended by NEI to make the ILRT risk assessment methodology more usable for plants that do not have a Level 3 PRA. Because FitzPatrick does not have a Level 3 PRA or associated plant-specific dose calculations for the EPRI's accident

classes, this calculation uses NUREG-1 150 dose results. Specifically, the doses for Peach Bottom Unit 2, as documented in NUREG/CR-4551 112] are used.

The following substeps describes the methodology use for obtaining population dose result estimates for FitzPatrick using information from NUREG/CR-4551 for Peach Bottom Unit 2 [121:

• Calculate Peach Bottom Unit 2 NUREGICR-4551 Study Population Dose

• Calculate FitzPatrick Person-Rem Per Collapsed Accident Progression Bin (APB)

• Calculate FitzPatrick Person-Rem Frequency by Accident Progression Bin (APB)

• Calculate EPRI 's TR -104285 50-Mile Population Dose Calculate Peach Bottom Unit 2 NUREGICR-4551 Study Populatlon Dose.

Given that the NUREG/CR-4551 does not document dose results as a function of collapsed APBs, calculations are performed to obtain these values. The calculation is as follows:

The Peach Bottom Unit 2 individual population dose (person-rem) for each collapsed APB is calculated by dividing the individual mean population dose (person-rem per reactor year) for each collapsed APB by the individual mean frequency contributions for each collapsed APB, or PDCAPB - MPDCAPB / MFCAPB Where:

MPDCAPB = individual mean population dose (person-rem per reactor year) for each collapsed APB

= MFCRCApB

• MPDso Where:

MFCRCAPB -

mean fractional contribution to risk for each collapsed APB [Section 2.3 Input#11]

MPD 50 mean population dose at 50 miles for Peach Bottom Unit 2 [Section 2.3 lnput#8]

Therefore, M PDCAP8 I 0.0210 7.9 0.1659 MPDCAPB2 0.0066 7.9 0.05214 MPDCAPB3 0.5560 7.9 4.3924 MPIDCAPS4 0.2260

• 7.9 1.7854 MPDCAP85 0.0022 7.9 0.01738 MPDCAP86 0.0590 7.9 0.4661 0.1180 7.9 0.9322 MPDCAPBS

• 0.0005 7.9 0.00395 MPDCAPB9 0.0100 7.9 0.079 MPD)CAPBIO 0.000 7.9 0.0 Where:

MFCCAPB = individual mean frequency contributions for each collapsed APB

= IMFCCAPB

• CDFPBU 2 Where:

IMFCCAPB individual mean fractional contributions for each collapsed APB [Section 2.3 lnput#12]

CDFpBU2 = Peach Bottom Unit 2 core damage frequency [Section 2.3 Input#9]

Therefore, MFCCAPB1 = 0.0220

• 4.34 x 10' = 9.55 x 10, MFCCAPB2 = 0.0110

• 4.34 x 104 = 4.77 x 104 MFCCAPB3 = 0.3410

• 4.34 x 10 = 1.48 x 104 MFCCAPB4 = 0.1830

• 4.34 x 104 = 7.94 x 107 MFCCApB5 = 0.0030

• 4.34 x I04 = 1.30 x 10" MFCCAPB6 = 0.0470

• 4.34 x IO 4 = 2.04 x 107 MFCCApB7 = 0.1100

• 4.34 x 10 4 = 4.77 x 10 7 MFCCAPB8 = 0.1840

• 4.34 x IO' = 7.99 x 107 MFCCAPB9 = 0.0890

• 4.34 x 1O = 3.86 x IO' MFCCApBlo = 0.0100

• 4.34 x 104 = 4.34 x 10 4 Therefore, P DCAPB1 = 0.1659 / 9.55x104 = 1.74x106 (person-rem)

PDCAPB2 = 0.05214 / 4.77 x 104 = 1.09 x 10 (person-rem)

PDCAPB3 = 4.3924 / 1.48 x 10 = 2.97 x 106 (person-rem)

PDcAPB4 = 1.7854 / 7.94 x 10 7 = 2.25 x 106 (person-rem)

PDcAPBs = 0.01738 / 1.30x104 = 1.34x106 (person-rem)

PDCAPB6 = 0.4661 I 2.04 x 10-7 = 2.28 x 106 (person-rem)

PDCAPB7 = 0.9322 / 4.77 x10-7 = 1.95 x 106 (person-rem)

PDCAPBS = 0.00395 / 7.99 x 107 = 4.94 x 10 3 (person-rem)

PDCAp19 = 0.079 / 3.86 x107 = 2.05 x 105 (person-rem)

PDCAPB1O = 0.0 4.34x I0 = 0.0 (person-rem)

Table 2-6 shows the 50-mile population dose (person-rem) for each APB considered in the NUREG/CR-4551 Peach Bottom Unit 2 study [121.

Calculate FitzPatrick Person-Rem Per Collapsed Accident Progresslon Bin (APB).

The Peach Bottom Unit [2] NUREG/CR-4551 consequences summarized in Table 2-6 (and calculated above) should be adjusted for use in this analysis to account for differences in the following parameters:

reactor power level, technical specification allowed containment leakage rate, and population.

Reactor Power Level Adjustment The adjustment factor for reactor power level is defined as the ratio of the power level at FitzPatrick to that at Peach Bottom Unit 2. This adjustment factor is calculated as follows:

AFper = PLF / PLP Where:

AFpm~r =th e adjustment factor for reactor power level PLF = the power level at FitzPatrick = 2536 MWth (Section 2.3 Input #18]

PLP = the power level at Peach Bottom Unit 2 = 3293 M'th [Section 2.3 Input #15]

Therefore, AFper = 2536 I 3293 = 0.77

Technical Specification Allowed Containment Leakace Rate Adjustment The adjustment factor for technical specification (TS) allowed containment leakage is defined as the ratio of the containment leakage at FitzPatrick to that at Peach Bottom Unit 2. This adjustment factor is calculated as follows:

AFLwkage = LRF I LRP {equation 1I Where:

AFLeakage = the adjustment factor for TS allowed containment leakage LRF = the TS allowed containment leakage at FitzPatrick LRP = the TS allowed containment leakage at Peach Bottom Unit 2 Because the leakage rates are in terms of the containment volume, the ratio of containment volumes is needed to relate the leakage rates. Therefore, LRF = TSJAF* VOLJAF {equation 2)

LRP = TSPB

• VOLPB {equation 3)

Where:

TSJAF= TS maximum allowed containment leakage is 1.5 volume%/day [Section 2.3 Input #19]

VOLJAF FitzPatrick containment free volume = 264, 000 ft3 [Section 2,3 Input #20]

TSPB TS maximum allowed containment leakage is 0.5 volume%Iday [Section 2.3 Input #16]

VOLPB Peach Bottom Unit containment free volume = 307, 000 ft3 [Section 2,3 Input #17]

Therefore, substituting equation 2 and 3 into I yields, AFiekawge (1.5

• 264000) / (0.5 307000)

AFLwkag 2.58 Population Adjustment The adjustment factor for population is defined as the ratio of the population within 50-mile radius of FitzPatrick to that of Peach Bottom Unit 2. This adjustment factor is calculated as follows:

AFpopulation = POPF / POPP Where:

AF12opulation = the adjustment factor for population POPF = population within 50-mile radius of FitzPatrick = 8.98 x 10 [Section 2.3 Input #14]

POPP = population within 50-mile radius of Peach Bottom Unit 2

= 3.02x106 [Section 2.3 Input#131 Therefore, AFpopuwn = 8.98 x I05 I 3.02 X Id AFpopum. = 0.297

The above adjustment factors that are use in adjusting the population dose (person-rem) of the Peach Bottom Unit 2 for the FitzPatrick site and plant differences are as follows:

• Consequence categories dependent on the wINTACT Tech Spec Leakage (collapsed accident progression bins 8 and 10)

AF8, 10 = AFpower

• AFLeakage

• AFPopulation Where:

adjustment factor for collapsed accident progression bins 8 and 10 AFpower the adjustment factor for reactor power level = 0.77 [from above]

AFLeakage the adjustment factor for TS allowed containment leakage = 2.58 [from above]

AFpoptdabon the adjustment factor for population = 0.297 [from above]

Therefore, AFe,1 o = 0.77

• 2.58

• 0.297 = 0.59

• Consequence categories not dependent on the Technical Specification Leakage (collapsed accident progression bins 1, 2, 3, 4, 5, 6, 7 and 9)

AF 1, 2,3,4.5.6,7,9 - Apower

• Al~population Where:

AFI 2 ,3. 4. 5,6,7,9 = adjustment factor for collapsed accident progression bins 1, 2, 3, 4, 5, 6, 7, and 9 AFpawer = the adjustment factor for reactor power level = 0.77 [from above]

AFPopu[aMon = the adjustment factor for population = 0.297 [from above]

Therefore,

= 0.77

• 0.297 = 0.22869 Based on the above adjustments factors for intact and non-intact accident progressions and the 50-mile population dose (person-rem) for each APB considered in the NUREG/CR-4551 Peach Bottom Unit 2 study, Table 2-6, the FitzPatrick doses (person-rem) are calculated as follows:

JAFMPDCAPBI PDCAP1BI AF1,Z 34 ,5,6,7,9

• JAFMPDCAPB 2 PDCAP1B2 JAFMPDCAPB 3 PDCAPB3 JAFMPDbAPB 4 PDwAB JAFMPDCAPBS AFI. Z3.4,5.e,7, 9

• PDCAP1B5 4

JAFMPDCAPBr, PDCA1BB JAFMPDCAPB 7 AFI, 2.3,4,a, 6,7. 9 PDW13p 7 JAFMPDCAPBS PDwABs JAFMPDwpBg AF8, 10 PDCAP139 JAFMPDCAB10 PDcB10i Therefore, JAFMPDc~pBi 0.22869

• 1.74 x 106 3.98 x 10l JAFMPDcAPB 2 0.22869

• 1.09 x 106 2.49 x 105 JAFMPDcAPB 3 0.22869 2.97 x 106 6.79 x 105 JAFMPDcAPB 4 0.22869 2.25x 106 5.15 x 105 JAFMPDCAPBS 0.22869

• 1.34 x 10' 3.06 x 105 JAFMPDcAPB6 0.22869

• 2.28 x 16 5.21 x 105 JAFMPDcAPB7r 0.22869

• 1.95 x 106 4.46 x 105

JAFMPDcAPB8 = 0.59

• 4.94 x 105 = 2.91 x 103 JAFMPDcApB9 = 0.22869

• 2.05 x 10 = 4.69 x 104 JAFMPDCAPBlo = 0.59 °0.0 0.0 Table 2-7 summarizes the Peach Bottom Unit 2 NUREG/CR-4551 [12] doses after adjustment for changes in population, reactor power level, and containment leakage rate for application to FitzPatrick.

Calculate FitzPatrick Accident Progression Bin (APB) Freguency The FitzPatrick person-rem frequency is calculated in terms of collapsed accident progression bins (Table 2-4). The calculation is performed by running the FitzPatrick Level 2 containment event tree model to match the criteria for the Peach Bottom Unit 2 NUREG/CR-4551 study [12]. The results of this sort is presented in Table 2-8 and detailed in Attachment A.

Calculate EPRI 'l TR-104285 50-Mile Populatlon Dose The FitzPatrick's person-rem results (Table 2-7) are converted to match to the EPRI TR-104285 release classes. The calculation assigns each of the FitzPatrick Level 2 source term category endstates (Tables 2-2 and 2-3) to the equivalent NUREGICR-4551 Peach Bottom Unit 2 [12] collapsed accident progression bin category (Table 2-4).

This is required because the FitzPatrick IPE Level 2 results are not defined in the same terms as reported in NUREG/CR-4551. Therefore, in order to use the Level 3 results presented in NUREG/CR-4551 for Peach Bottom Unit 2 [121, the FitzPatrick IPE Level 2 results needs to be converted into a format that allows the use of the Peach Bottom Unit Level 3 results.

The FitzPatrick IPE provides a grouping of containment failure modes and subsequent release categories resulting from severe accident challenges. Tables 2-2 and 2-3 provide this breakdown for the FitzPatrick IPE. The FitzPatrick release endstates of Tables 2-2 and 2-3 were reviewed and assigned into one of the collapsed accident progression bins from NUREG/CR-4551 [12]. The result of this review is presented in Table 2-9.

The following discussion provides the basis for the assignment of population dose for each EPRI accident class.

The 50-miles population dose for the EPRI accident class 'no containment failure' is based on the FitzPatrick's collapsed accident progression bins 8 and 10 (Table 2-7) as the ones closest to the definition of an intact containment. The population dose is calculated as the weighted mean 50-mile population dose for each collapsed accident progression bin.

CLASS-1-DOSE = JAFWPDCAPSO + JAFWPDCAPBIO Where:

CLASSIDOSE = 50-miles population dose for the EPRI accident class "no containment failure' JAFWPDCAPBa = FitzPatrick 50-miles collapsed APB 8 weighted mean population dose (person-rem)

JAFWPDcApBio = FitzPatrick 50-miles collapsed APB 10 weighted mean population dose (person-rem)

JAFWPDcAPB8 = JAF-FcpB JAF-FCAPB

• JAFMPDcAPB8 L JAF-FCAPBB + JAF-FcApBlo

JAFWPDcAPBIO = JAFJAFWPDCAP~~~~lO= JA'FCAAPPDCPI JAFMPDCAPBIO I JAF-FCAPBS + JAF-FCAP131 I Where:

JAF-FCAPBS = FitzPatrick collapsed APB 8 mean frequency [Table 2-8]

JAF-FCAPBIO = FitzPatrick collapsed APB 10 mean frequency [Table 2-8]

JAFMPDCAPBB = FitzPatrick 50-miles collapsed APB 8 mean population dose (person-rem) [Table 2-7]

JAFMPDCAPB10 = FitzPatrick 50-miles collapsed APB 10 mean population dose (person-rem) [Table 2-7]

Therefore, CLASSIDOSE = 2.91 x 103 person-rem The 50-miles population dose for the EPRI accident Class 2 (Large Containment Isolation Failures, failure-to-close) is based on the FitzPatrick's collapsed accident progression bin 3 (Table 2-7) as the one closest to the definition of large containment isolation failure. This selection is based on assuming that the containment isolation failure of EPRI accident Class 2 occurs concurrent with early drywell failure at high RPV pressure. Collapsed accident progression bin 3 results in the highest dose of all of the FitzPatrick 'containment failure" collapsed accident progression bins (which is indicative of a containment failure with torus pool and drywell bypass).

Therefore, CLASS_2_DOSE = 6.79 x 10 person-rem The 50-miles population dose for the EPRI accident Class 3a (Small Isolation Failures-Liner breach) and accident Class 3b (Large Isolation Failures-Liner breach), per the NEI Interim Guidance [3], are taken as factors of 1OLa and 35La [4, 8], respectively, times the population dose of EPRI accident Class 1.

Therefore, CLASS_3aDOSE = 10* CLASS 1-DOSE CLASS_3b_DOSE = 35

• CLASS_7_DOSE CLASS_3aDOSE = 10 *2.91 x 10 3 CLASS_3bDOSE = 35

• 2.91 x103 CLASS 3a DOSE = 2.91 x 104 person-rem CLASS_3b_DOSE = 1.02 x 105 person-rem Per the NEI Interim Guidance [3], EPRI accident Classes 4 (Small Isolation Failure - failure-to-seal, Type B test), 5 (Small Isolation Failure - failure-to-seal, Type C test), and 6 (Containment Isolation Failures, dependent failures, personnel errors) are not affected by ILRT frequency and are not analyzed as part of this risk assessment. Therefore no selections of population does estimates are made for these accident classes.

The 50-miles population dose for the EPRI accident Class 7a (Severe Accident Phenomena Induced Early Drywell Failures) is based on the FitzPatrick's collapsed accident progression bins 3 and 4 (Table 2-

7) as the ones closest to the definition of early drywell failures. The population dose is calculated as the weighted mean 50-mile population dose for each collapsed accident progression bin.

CLASS_7aDOSE = JAFWPDCAPB 3 + JAFWPDCAPB4

Where:

CLASS_7aDOSE = 50-miles population dose for the EPRI accident class 7a JAFWPDcAPB3 = FitzPatrick 50-miles collapsed APB 3 weighted mean population dose (person-rem)

JAFWPDcAPB 4 = FitzPatrick 50-miles collapsed APB 4 weighted mean population dose (person-rem)

JAFWPDcAPB 3 =-- JAF-FCAPB 3 I

• JAFMPDcApB 3 JAF-FCAPB 3 + JAF-FCAPB 4 JAFWPDCAP1. 4 = [. JAF-FcAps.4

• JAFMPDcAP8p4 JAF-FCAPB3 + JAF-FcApB4 I Where:

JAF-FCAPB 3 = FitzPatrick collapsed APB 3 mean frequency [Table 2-8]

JAF-FCAPB 4 = FitzPatrick collapsed APB 4 mean frequency [Table 2-8]

JAFMPDWpB 3 = FitzPatrick 50-miles collapsed APB 3 mean population dose (person-rem) [Table 2-7]

JAFMPDcAPB4 = FitzPatrick 50-miles collapsed APB 4 mean population dose (person-rem) [Table 2-7]

Therefore, CLASS_7aDOSE = 6.11 x 105 person-rem The 50-miles population dose for the EPRI accident Class 7b (Severe Accident Phenomena Induced Early Torus Failures) is based on the FitzPatrick's collapsed accident progression bins I and 2 (Table 7) as the ones closest to the definition of early torus failures. The population dose is calculated as the weighted mean 50-mile population dose for each collapsed accident progression bin.

CLASS_7bDOSE = JAFWPDcApBI + JAFWPDcApB 2 Where:

CLASS_7bDOSE = 50-miles population dose for the EPRI accident class 7b JAFWPDCAPBI = FitzPatrick 50-miles collapsed APB 1 weighted mean population dose (person-rem)

JAFWPDCAPB 2 = FitzPatrick 50-miles collapsed APB 2 weighted mean population dose (person-rem)

JAFWPDcApsi = JAF-FCAPBI JAFMPDCAPBI I JAF-Fcp~pal + JAF-FCAPB 2 I

• JAFWPDCAP8 2 =- JAF-FCAPB 2 I

• JAFMPEDCAPB 2 JAF-FcApBI + JAF-FCAPS2 Where:

JAF-FCAPBI = FitzPatrick collapsed APB 1 mean frequency [Table 2-8]

JAF-FCAPB 2 = FitzPatrick collapsed APB 2 mean frequency [Table 2-8]

JAFMPDCAPBl = FitzPatrick 50-miles collapsed APB 1 mean population dose (person-rem) [Table 2-7]

JAFMPDCAPB 2 = FitzPatrick 50-miles collapsed APB 2 mean population dose (person-rem) [Table 2-7]

Therefore, CLASS 7b DOSE = 3.96 x 106 person-rem The 50-miles population dose for the EPRI accident Class 7c (Severe Accident Phenomena Induced Late Drywell Failures) is based on the FitzPatrick's collapsed accident progression bin 6 (Table 2-7) as the one closest to the definition of late drywell failures.

Therefore, CLASS_7cDOSE = 5.21 x 105 person-rem The 50-miles population dose for the EPRI accident Class 7d (Severe Accident Phenomena Induced Late Torus Failures) is based on the FitzPatrick's collapsed accident progression bin 5 (Table 2-7) as the one closest to the definition of late torus failures.

Therefore, CLASS_7d_DOSE = 3.06 x 105 person-rem The 50-miles population dose for the EPRI accident Class 8 (Bypass) is based on the FitzPatrick's collapsed accident progression bin 7 (Table 2-7) as the one closest to the definition of bypass failure.

This selection is based the dominance of ATWS induced bypass failure of the torus for this accident class. Collapsed accident progression bin 7 represents the release due to torus venting, therefore, it 's indicative of containment bypass scenarios.

Therefore, CLASS_8_DOSE = 4.46 x 105 person-rem Using the preceding information, the population dose for the 50-mile radius surrounding Fitzpatrick is summarized in Table 2-10. (Note: the use of dose results for the 50-mile radius around the plant as a

'figure of merit' in the risk evaluation is consistent with past ILRT frequency extension submittals, and the NEI Interim Guidance [3]).

2.4A Baseline Population Dose Rate Estimate (Step 4)

This step calculates the baseline does rates for each of the eight EPRI's accident classes. The calculation is performed by multiplying the dose calculated in Step 3 (Table 2-10) by the associated frequency calculated in Step I (Table 2-5). Since the conditional containment pre-existing leakage probabilities for EPRI accident classes' 3a and 3b are based on a 3-per-10 year ILRT frequency, the calculated baseline results reflect a 3-per-10 year ILRT surveillance frequency.

CLASS_ _DOSERATE CLASSI-DOSE CLASS_1_FREQUENCY CLASS_2_DOSERATE CLASS72 DOSE

• CLASS_2_FREQUENCY CLASS_3aDOSERATE CLASS_.3aDOSE CLASS_3a_FREQUENCY CLASS_3b_DOSERATE CLASS_3bDOSE

• CLASS_3b_FREQUENCY CLASS_7aDOSERATE CLASS_7aDOSE

• CLASS 7aFREQUENCY CLASS_7b_DOSEmTE CLASS_7bDOSE

• CLASS_7b_FREQUENCY CLASS_7cDOSERATE CLASS 7cDOSE CLASS_7cFREQUENCY CLASS_7d_DOSERATE CLASS_7dDOSE

• CLASS_7dFREQUENCY CLASS_8_DOSERATE CLASS_8 DOSE

• CLASS_8_FREQUENCY Therefore, CLASS ... DOSERATE 2.91 x 10 3

• 6.19x 107 1.80 x 10-3 (person-rem/ry)

CLASS_2_DOSE.T 6.79 x 105

• 2.44 x 10' 1.66 x 10-3 (person-rem/ry)

CLASS_3a_DOSE 1. 2.91 x 104

• 6.59 x 10' 1.92 x 10- (person-rem/ry)

CLASSC3b1_DOSER.TE 1.02 x 105

• 6.59 x 10i 6.72 x 104 (person-rem/ry)

CLASS_7aDOSERATE 6.11 x 105

• 1.74 x 10i 7 1.06 x 10 (person-remlry)

CLASS 7b DOSERATE 3.96 x l0O' 7.37 x i O' 2.92 x 10 1 (person-rem/ry)

CLASS_7c_DOSERATE 5.21 x10'

• 5.82 x104 3.03 x 10 2 (person-rem/ry)

CLASS_7eDOSERATE 3.06 x1IO

• 4.83 x10'7 1.48 x 10- (person-rem/ry)

CLASS_8_DOSE. 4.46 x10'

• 2.97 x10 1.32 x 101 (person-rem/ry)

Table 2-11 summarizes the resulting baseline population dose rates by EPRI accident class.

2A.5 Change In Probability of Detectable Leakage (Step 6)

This step calculates 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.

According to NUREG-1493 [7] and per the NEI Interim Guidance [3], the calculation of the change in the probability of a pre-existing ILRT-detectable containment leakage is based on the relationship that relaxation of the ILRT interval results in increasing the average time that a pre-existing leak would exist undetected. Specifically, the relaxation of the Type A ILRT interval from 3-in-10 years to 1-in-10 years will increase the average time that a leak detectable only by an ILRT goes undetected from 18 to 60 months8 ,

a factor of 3.333 increase (60/18). Therefore, the change in probability of leakage due to the ILRT interval extension is calculated by applying a multiplier factor determined by the ratio of the average times of undetection for the two ILRT interval cases.

From Section 2.3 'Data and Design Criteria", the calculated pre-existing ILRT detectable leakage probabilities based on 3 in-10 years ILRT frequency is 0.027 for small pre-existing leakage (EPRI accident class 3a) and 0.0027 for large pre-existing leakage (EPRI accident class 3b).

Since October 1996, the FitzPatrick plant has been operating under a 1-in-10 years ILRT testing frequency consistent with the performance-based Option B of 10 CFR Part 50, Appendix J. [17]. As a result, the baseline leakage probabilities, (which are based on a 3-in-10 years ILRT frequency) must be revised to reflect the current 1-in-10 years FitzPatrick ILRT testing frequency. This is performed as follows:

PROBd.,ak_3 , so = PROBcjasas3a I SURTESTO10 PROBgdMs_3bI0 = PROBcjass-3b I SURTESTO10 Where:

PROBdalss3a_10 = probability of small pre-existing containment liner leakage given a 1-in-10 years ILRT frequency.

PROBdass 3b_10 = probability of large pre-existing containment liner leakage given a 1-in-10 years ILRT frequency.

PROBcass 3a = probability of small pre-existing containment liner leakage given a 3in-10 years ILRT I Multiplying the test interval by 1/2and multiplying by 12 to convert from a year to months calculates the average time for undetection.

frequency = 0.027 (Section 2.3, inputt#6 PROBdass 3b = probability of large pre-existing containment liner leakage given a 3-in-10 years ILRT frequency = 0.0027 [Section 23, input #7]

SURTESTio = surveillance interval of interest, months/2 = 10 years*12mnthsA/2 = 60 months year Therefore, PROBdw,_3 1j = 0.027 * [ 60 l = 0.09 18 PROBdgassUb1 = 0.0027* [ 60 l = 0.009 18 Similarly, the pre-existing ILRT detectable leakage probabilities for the 1-in-15 years ILRT frequency being analyzed by FitzPatrick are calculated as follows:

PROBdw~ 3 ,J 5 = PROBdas_3.

• SURTEST15 18 PROBdaa3bl15 = PROBdass 3b * [ SURTESTI5 J L 18 ]

Where:

PROBdass.3a_15 = probability of small pre-existing containment liner leakage given a 1-in-15 years ILRT frequency.

PROBda,3b-15 - probability of large pre-existing containment liner leakage given a 1-in-15 years ILRT frequency.

SURTESTrs = surveillance interval of interest, months/2 = 15 years*12months/2 = 90 months year Therefore, PROBd.~3,15 = 0.027 * [ 90 l = 0.135 L 18 PROBda.s 3b-1 = 0.0027 * [ 90 I = 0.0135 18 Given the above revised leakage probabilities, the frequencies of the EPRI accident classes calculated in Step 1, also needs to be revised to reflect the increase change in leakage probabilities.

As previously stated, Type A tests impact only Class I and Class 3 sequences. Therefore, EPRI accident Class I frequency changes are calculated similar to Step 1, and the rest of EPRI's Classes; 2, 7 and 8 remain the same.

Revised Frequency of EPRI Class 3a Sequences. Consistent with NEI Interim Guidance [3], the frequency per reactor year for this category is calculated as:

CLASS_3a-FREQUENCY 10 = PROB3d., 3a 1 0 'CDF CLASS 3a FREQUENCY1 5 = PROBdass 3a 15 *CDF Where:

CLASS_3aJFREQUENCY 1o = frequency of small pre-existing containment liner leakage given a 1-in-10 years ILRT interval CLASS_3aFREQUENCYIS5 = frequency of small pre-existing containment liner leakage given a 1-in-15 years ILRT interval PROBdass_3 a_10 = probability of small pre-existing containment liner leakage given a 1-in-l0 years ILRT frequency = 0.09 [See above write-up]

PROBdm3a S = probability of small pre-existing containment liner leakage given a 1-in-15 years ILRT frequency = 0.135 [See above write-up]

CDF = FitzPatrick IPE core damage frequency = 2.44 x 104 /ry [Section 2.3, input#2]

Therefore, CLASS_3a_FREQUENCY 1 0 = 0.090

• 2.44 x 10-5 = 2.20 x 10'7/ry CLASS_3aFREQUENCY15 = 0.135

• 2.44 x i0 4 = 3.29 x 107/ry Freouencv of EPRI Class 3b Sequences. Consistent with NEI Interim Guidance [3], the frequency per reactor year for this category is calculated as:

CLASS_3b-FREQUENCY 10 = PROBdaw,_J 10

• CDF CLASS_3bFREQUENCY 1 5 = PROBd. 3b_15

• CDF Where:

CLASS_3bFREQUENCYjO = frequency of small pre-existing containment liner leakage given a 1-in-10 years ILRT interval CLASS_3bFREQUENCY I 5 = frequency of small pre-existing containment liner leakage given a 1-in-15 years ILRT interval Therefore, CLASS_3bFREQUENCYI0 = 0.0090 *2.44 x l0e = 2.20 x 104 /ry CLASS_3b .FREQUENCY, 5 = 0.0135

• 2.44 x 10 = 3.29 x lO4 /ry Freauency of EPRI Class I Sequences. Consistent with NEI Interim Guidance [3], the frequency per reactor year for these sequences is calculated by subtracting the frequencies of EPRI Classes 3a and 3b from the sum of all severe accident progression sequence frequencies in which the containment is isolated and intact:

CLASSIFREQUENCYio = NCF - CLASS_3aFREQUENCY 1 0 - CLASS_3bFREQUENCYIO

CLASS_1_FREQUENCY1s = NCF - CLASS 3a FREQUENCYi 5 - CLASS_3bFREQUENCYi5 Where:

NCF = frequency in which containment leakage is at or below maximum allowable Technical Specification Leakage = 6.91 x l0e /ry [Table 2-2]

CLASSIFREQUENCYio = frequency of no containment failure given a 1-in-10 years ILRT interval CLASS_1_FREQUENCY 15 = frequency of no containment failure given a 1-in-15 years ILRT interval Therefore:

CLASSIFREQUENCYro = 6.91 x 10-7 -2.20 x 107-2.20 x 10, = 4.49 x10Iery CLASS_1_FREQUENCY 15 = 6.91 x 10 - 3.29 x 10 3.29 x 108 = 3.29 x I0-7/ry The impacted frequencies of the EPRI accident classes are summarized in Table 2-12.

2.4.6 Population Dose Rate for New ILRT Interval (Step 6)

This step, per the NEI Interim Guidance [3], calculates the population dose rate for the new surveillance intervals of interest by multiplying the population dose (Table 2-10) by the frequency for each of the eight EPRI's accident classes (Tables 2-5 and 2-12). In addition, sum the accident class dose rates to obtain the total dose rate.

Per the NEI Interim Guidance [3], EPRI accident Classes 4 (Small Isolation Failure - failure-to-seal, Type B test), 5 (Small Isolation Failure - failure-to-seal, Type C test), and 6 (Containment Isolation Failures, dependent failures, personnel errors) are not affected by ILRT frequency and are not analyzed as part of this risk assessment. Therefore no selections of population dose estimates are made for these accident classes.

The calculation for a 1-in-10 years ILRT interval is as follows:

CLASS 1 DOSERATE-so CLASS _1DOSE CLASS 1 FREQUENCYlo CLASS_2_DOSERATE-io CLASS 27DOSE CLASS_2_FREQUENCY 1o CLASS ..3a DOSERATE-10 CLASS73a DOSE

• CLASS_3a_FREQUENCYIo CLASS_3b6DOSERATE-.o CLASS73b DOSE CLASS_3b_FREQUENCY 1 o CLASS_7aDOSERATE-10 CLASS_7aDOSE CLASS77a FREQUENCYi 0 CLASS.7b7DOSERATE-1o CLASS 7b DOSE CLASS_7byFREQUENCYo CLASS_7cDOSERATE-1o CLASS 7c DOSE CLASS_7cFREQUENCYIo CLASS_7d_DOSERATE-10 CLASS_7dDOSE CLASS_7d_FREQUENCY,,

CLASS_8DOSERATE-io CLASS-8 DOSE CLASS_8_FREQUENCY 10 Where:

CLASS I DOSERATE-1 o EPRI accident Class 1 dose rate given a 1-in-10 years ILRT interval CLASS_2_DOSERATE-1 o EPRI accident Class 2 dose rate given a 1-in-10 years ILRT interval CLASS 3a DOSERATE-1o EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval CLASS_3b_DOSERATE-1o EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval CLASS77a.DOSERATE-io EPRI accident Class 7a dose rate given a 14n-10 years ILRT interval CLASS_7b_DOSERATE-.o EPRI accident Class 7b dose rate given a 1-in-10 years ILRT interval CLASS_7cDOSERATE-10 EPRI accident Class 7c dose rate given a 1in-10 years ILRT interval

CLASS_7d_DOSERATE-.o = EPRI accident Class 7d dose rate given a 1-in-10 years ILRT interval CLASS_8_DOSERATE-1O = EPRI accident Class 8 dose rate given a 1-in-10 years ILRT interval Therefore, CLASSIDOSERATEo10 2.91 x 10 3

• 4.49 x le 1.31 x 10-3 (person-rem/ry)

CLASS_2_DOSERATE.1 o 6.79 x 10' 2.44 x 109 1.66 x 10-3 (person-rem/ry)

CLASS_3aDOSERATE-1 0 2.91 x 104

• 2.20 x 1e 6.40 x 10-3 (person-rem/ry)

CLASS_3bDOSEATE-lo 1.02 x 10'

• 2.20 x 10' 2.24 x 10-3 (person-rem/ry)

CLASS_7aDOSERATE-jo 6.11 x 10' 1.74 x 10- 1.06 x 10-1 (person-rem/ry)

CLASS_7b_DOSERATE-.o 3.96 x 105

• 7.37 x 10- 2.92 x 10.1 (person-rem/ry)

CLASS_7cDOSERATE-10 5.21 x 105

• 5.82 x 10' 3.03 x 102 (person-rem/ry)

CLASS_7dDOSERATE..o 3.06 x 10'

• 4.83 x 10-7 1.48 x 101 (person-rem/ry)

CLASS.- 8.-DOSERATE-1O 4.46 x 10'

• 2.97 x 10- 1.32 x 10- (person-rem/ry)

The calculation for a 1-in-15 years ILRT interval is as follows for the:

CLASSIDOSERATE- 15 CLASSIDOSE

• CLASS 1_FREQUENCY 1 s CLASS 2 DOSERATE.15 CLASS72 DOSE

• CLASS_2_FREQUENCY1 5 CLASS_3a_DOSERATE- 1 5 CLASS_3aDOSE

• CLASS_3a_FREQUENCY15 CLASS_3b DOSERATE-15 CLASS 3b DOSE

• CLASS_3b_FREQUENCY1S CLASS_7a_DOSERATE-,1 5 CLASS_7aDOSE

• CLASS_7aFREQUENCY,,

CLASS_7b_DOSERATE-15 CLASS7b_.DOSE

• CLASS_7b_FREQUENCY15 CLASS_7c DOSERATE- 15 CLASS 7c DOSE

• CLASS_7cFREQUENCY 15 CLASS_7d_DOSERATE-15 CLASS_7d_DOSE

• CLASS_7dFREQUENCY15 CLASS_8_DOSERATE- 1 5 CLASS_8_DOSE

• CLASS_8_FREQUENCY 15 Where:

CLASS 1IDOSERATE-15 EPRI accident Class I dose rate given a 1-in-15 years ILRT interval CLASS_2_DOSERATE-,1 5 EPRI accident Class 2 dose rate given a 1-in-15 years ILRT interval CLASS_3a_DOSERATE-s1 EPRI accident Class 3a dose rate given a 1-in-15 years ILRT interval CLASS_3bDOSERATE-1s EPRI accident Class 3b dose rate given a 1-in-15 years ILRT interval CLASS-7aDOSERATE-15 EPRI accident Class 7a dose rate given a 1-in-15 years ILRT interval CLASS b7.DOSERATE-1s EPRI accident Class 7b dose rate given a 1-n-15 years ILRT interval CLASS_7c-DOSERATE- 15 EPRI accident Class 7c dose rate given a -in-15 years ILRT interval CLASS 7dDOSERATE-1s EPRI accident Class 7d dose rate given a 1-in-15 years ILRT interval CLASS_8_DOSERAT-mis EPRI accident Class 8 dose rate given a 1-in-15 years ILRT interval Therefore, CLASSI DOSERATE.15 2.91 x 103

• 3.29 x 107 9.58 x 104 (person-rem/ry)

CLASS_2 DOSERAT,., 5 6.79 x 105

• 2.44 x 10 9 1.66 x 10-3 (person-rem/ry)

CLASS-3a DOSERATE.15 2.91 x le 3.29 x 10' 9.60 x 10-3 (person-rem/ry)

CLASS_3bDOSERATE- 1s 1.02 x 105

• 3.29 x 104 3.36 x 10-3 (person-rem/ry)

CLASS_7a DOSERATE 1s 6.11 x 10'

• 1.74 x 10-7 1.06 x 10I (person-rem/ry)

CLASS77b7DOSERATE.ls 3.96 x 10'

• 7.37 x 10-7 2.92 x 10- (person-rem/ry)

CLASS_.7cDOSERATE- 1 s 5.21 x 105

• 5.82 x 10' 3.03 x 10-2 (person-rem/ry)

CLASSd.7DOSERATE-IS 3.06 x 105

• 4.83 x 10- 1.48 x 101 (person-rem/ry)

CLASS_8_DOSERATE-15 4.46 x 105

• 2.97 x 10-7 1.32 x 101 (person-rem/ry)

The dose rates per EPRI accident class as a function of ILRT interval are summarized in Table 2-13.

2.4.7 Change In Population Dose Rate Due to New ILRT Interval (Step 7)

This step, per the NEI Interim Guidance [3] calculates the percentage of the total dose rate attributable to EPRI accident Classes 3a and 3b (those accident classes affected by change in ILRT surveillance interval) and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

Based on the results summarized in Table 2-13, for the current FitzPatrick 1-inl0 years ILRT interval, the percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b is calculated as follows:

PERCHG10 = CLASS_3aDOSERATE.10 + CLASS-3b-DOSERATE.10 l 100 L TOT- DOSERATE-1o J Where:

PERCHGjo = percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given a 1-in-1O years ILRT interval TOT- DOSERATE-10 = Total dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval

= 0.720 [Table 2-131 Therefore, PERCHG10 = 6.40 x104 + 2.24 x 10 I

• 100 I 0.720 PERCHG10 = 1.2%

The percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b based on the propose 14n-15 years ILRT interval is calculated as follows:

PER-CHG 1 5 = cLASS-39LDOSERATE-15 + CLASS_3b1DOSERATE.1 5 1

• 100 CLASS_3a TOT- DOSERATE. 15 Where:

PERCHGj 5 = percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given a 1-in-15 years ILRT interval TOT- DOSERATE-15 = Total dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval 0.724 (person-rem/ry) [Table 2-13]

Therefore, PER..CHGiS = [. 9.60x104 +

0.724 3.36x104 I

• 100

PER.CHG15 = 1.8%

Based on the above results, the changes from the 1-in-10 years to 1-in-15 years dose rate is as follows:

INCREASE1I- 15 = [TOT- DOSERATE-i 5 - TOT- DOSERATE..10]

• 100 TOT- DOSERATE-10 Where:

INCREASEI 0 . 15 = percent change from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval Therefore, INCREASEs10. 5 = [0.724 - 0.720] 100 = 0.56%

0.720 The above increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.56%. This value can be considered to be a negligible increase in risk.

2.S8 Change In LERF Due to New ILRT Interval (Step 8)

This step, per the NEI Interim Guidance [3] calculates the change in the large early release frequency with extending the ILRT interval from 1-in-10 years to 1-4I5-years.

The risk impact associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from containment could in fact result in a large release due to failure to detect a pre-existing leak during the relaxation period. For this evaluation only accident Class 3 sequences have the potential to result in large releases if a pre-existing leak were present. Class I sequences are not considered as potential large release pathways because for these sequences the containment remains intact. Therefore, the containment leak rate is expected to be small (less than 2La). A larger leak rate would imply an impaired containment, such as classes 2, 3, 6 and 7.

Late releases are excluded regardless of the size of the leak because late releases are, by definition, not a LERF event. At the same time, sequences in the FitzPatrick IPE [6], which result in large releases (e.g.,

large isolation valve failures), are not impacted because a LERF will occur regardless of the presence of a pre-existing leak. Therefore, the frequency of accident Class 3b sequences (Table 2-12) is used as the LERF for FitzPatrick.

The affect on the LERF risk measure due to the proposed ILRT interval extension is calculated as follows:

ALERF10-15 = CLASS_3b_FREQUENC Y15 - CLASS_3b_FREQUENC Y10 Where:

ALERFiois = the change in LERF from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval CLASS_3bFREQUENC Y15 = frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval = 3.29x 108 /ry[Table2-121

CLASS_3bFREQUENC Y10 = frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval = 2.20 x 104 /ry [Table 2-12]

Therefore, ALERFIo.15 = 3.29 x 105 - 2.20 x10~

ALERFio.15 = 1.09 x 104 Iry Regulatory Guide 1.174 [5] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 [5] defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 104 /yr and increases in LERF below I e 7 /yr. Since the ILRT does not impact CDF, the relevant risk metric is LERF.

This ALERF of 1.09 x 108/ry falls into Region IlIl, Very Small Change in Risk (Figure 2.1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [5]. Therefore, because Regulatory Guide 1.174

[5] defines very small changes in LERF as below 107Iyr, increasing the ILRT interval at FitzPatrick from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.

It should be noted that if the risk increase is measured from the original 3-in-10-year interval, the increase in LERF is as follows:

ALERF3 15 = CLASS_3bFREQUENC Y15 - CLASS_3bFREQUENC Y3 Where:

ALERF3 15 = the change in LERF from 3-in-10 years ILRT interval to 1-in-15 years ILRT interval CLASS_3bFREQUENC Y15 = frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval = 3.29 x 10 /ry [Table 2-121 CLASS_3b_FREQUENC Y3 = frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval = 6.59 x 10 9ry [Table 2-12]

Therefore, ALERF3-1S = 3.29 x 10 - 6.59 x 0-9 ALERF3-j1 = 2.63 x I elry Similar to the ALERFIO-15 result, the ALERF3 15 is also non-risk significant from a risk perspective.

2A.9 Impact on Conditional Containment Failure Probability (Step 9)

This step, per the NEI Interim Guidance [3] calculates the change in conditional containment failure probability (CCFP). The CCFP risk metric ensures and shows that the proposed change in ILRT interval is consistent with the defense-in-depth philosophy expose in Regulatory Guide 1.174 [5]9.

9 The defenso-n-depth philosophy is maintained as a reasonable balance among prevention ofcore damage, containment failure and consequence mitigation.

In this calculation, the change in CCFP tracts the impact of the ILRT on both early (LERF) and late radionuclide releases. Based on the NEI Interim Guidance [3], CCFP consists of all those accident sequences resulting in a radionuclide release other that the intact containment state for EPRI accident Class 1, and small failures state for EPRI accident Class 3a. In addition, the CCFP is conditional given a severe core damage accident. The change in CCFP is calculated by the following equation:

CCFP=1-(lntact Containment Frequency/Total CDF)

Or CCFP= {1-([Class I frequency + Class 3a frequencyyCDF))*1 00, %

For the 1-in-10 years ILRT interval:

CCFP fi- (CLASS 1 FREQUENC Y10 + CLASS~.3a _FREQUENCYIo i 100%

ICFP,{ CDF }

Where:

CCFPjo = conditional containment failure probability given 1-in-10 years ILRT interval CDF = FitzPatrick IPE core damage frequency = 2.44 x 10Iry [Section 2.3, input#2]

CLASSIFREQUENC Y1o = frequency of EPRI accident Class I given a 1-in-10 years ILRT Interval = 4.49 x 10&7/ry [Table 2-12]

CLASS.3aFREQUENC Y 10 = frequency of EPRI accident Class 3a given a 1-in-10 years ILRT Interval = 2.20 x 1077/ry [Table 2-12]

Therefore, CCFP10 (i -CU 4.49 x 10 7 + 2.2Ox 10' 7 I) }*00%

2.44x1 CCFP10 = 72.6%

For the 1-in-15 years ILRT interval:

CCFP15 ii - ( ICLASS 1_FREQUENC Y15 + CLASS 3a FFREQUENCY15

([CLSSLREQENCy 15 CDF

)I* 100%

Where:

CCFP 15 = conditional containment failure probability given 1-in-15 years ILRT interval CDF = FitzPatrick IPE core damage frequency = 2.44 x 10 e/ry [Section 2.3, input#2J

CLASS.1IFREQUENC Yj 5 = frequency of EPRI accident Class I given a 1-in-15 years ILRT Interval = 3.29 x 10"1ry (Table 12]

CLASS_3aFREQUENC Y15 = frequency of EPRI accident Class 3a given a 1-in-15 years ILRT Interval = 3.29 x 10 1/ry [Table 12]

Therefore, CCFPI5 =11- [ 3.29 x 10- 4. 3.29 x107 IJ] }* 0%

2.44 x 104 CCFP1s = 73.03%

Therefore, the change in the conditional containment failure probability from 1in-10 years to 1-in-15 years is:

ACCFPIo., = CCFPjs - CCFPIO A&CCFPIo..15 = 73.03% - 72.6%

ACCFPIo.I 5 = 0.43%

This change in CCFP of less than 1% is insignificant from a risk perspective.

2.5 External Events Impact In response to Generic Letter 88-20, Supplement 4 [19], FitzPatrick submitted an Individual Plant Examination of External Events (IPEEE) in June 1996 [20]. The IPEEE was a review of external hazard risk (i.e., seismic, fires, high winds, external flooding, etc) to identify potential plant vulnerabilities and to understand severe accident risks. The results of the FitzPatrick IPEEE are therefore used in this risk assessment to provide a comparison of the effect of external hazards when extending the current 1-in-1 0 years to 1-in-15 years Type A ILRT interval.

The FitzPatrick IPEEE submittal [20] examined a spectrum of external events hazards based on acceptable screening methods (NRC seismic margin [21, 22], EPRI Fire PRA methodology [23], etc.).

These screening methods use varying levels of conservatism; therefore, it is not practical to incorporate realistic quantitative risk assessments of all external event hazards into the ILRT extension assessment at this time. As a result, external events hazards are evaluated as a sensitivity case to demonstrate that the conclusions of the internal events analysis would not be changed if external events hazards were considered.

The impact of external events on this ILRT risk assessment is summarized in this section (refer to Appendix A for further details).

The purpose of the external events evaluation is to determine whether there are any unique insights or important quantitative information that explicitly impact the risk assessment results when considering only internal events.

The quantitative consideration of external hazards is discussed in more detail in Appendix A of this report.

As can be seen from Appendix A, if the external hazard risk results of the FitzPatrick IPEEE are included in this assessment (i.e., in addition to internal events), the change in LERF associated with the increase in ILRT interval from 10 years to 15 years will be 1.03 x 10 7/ry. This delta LERF is just slightly above the Region IlIl boundary for LERF (Figure 2-1) and falls within NRC Regulatory Guide 1.174 [5] Region II

('Small Changes' in risk). As stated above, this can be attributed to the conservative screening nature of the external event methods available for their quantitative assessment at FitzPatrick.

Other salient results from Appendix A, found the increase in risk on the combined internal and external events total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, to be 0.16% or 0.038 person-rem/ry. In addition, the change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-15 years is 0.20%. A change in CCFP of less than 1% is insignificant from a risk perspective.

Therefore, incorporating external event accident sequence results into this analysis does not change the conclusion of internal events only risk assessment (i.e., increasing the FitzPatrick ILRT interval from 10 to 15 years is an acceptable plant change from a risk perspective). This results is expected, because the proposed ILRT interval extension impacts plant risk in a very specific and limited way.

2.6 Containment Liner Corrosion Risk Impact Recently, the NRC issued a series of Requests for Additional Information (RAls) in response to the one-time relief requests for the ILRT surveillance interval submitted by various licensees. One of the RAls related to the risk assessment performed in this report is provided below.

Reguest for Additional Information:

Inspections of reinforced and steel containments at some facilities (e.g., North Anna, Brunswick D.C.

Cook, and Oyster Creek) have indicated degradation from the uninspectable (embedded) side of the steel shell and liner of primary containments. The major uninspectable areas of the Mark I containment are the vertical portion of the dryweHf shell and part of the shell sandwiched between the drywell floor and the basemat. Please discuss what programs are used to monitor their conditions.

Also, address how potential leakage due to age-related degradation from these uninspectable areas are factored into the risk assessment in support of the requested interval extension.

The impact of the risk assessment portion of the above RAls is summarized in this section (refer to Appendix B for further details).

The containment liner corrosion analysis utilizes the referenced Calvert Cliffs Nuclear Power Plant assessment (24] to estimate the likelihood and risk-implication of degradation-induced leakage occurring and going undetected in visual examinations during the extended test interval. It should be noted that the Calvert Cliffs analysis was performed for a concrete cylinder and dome containment with a steel liner whereas FitzPatrick has a free standing steel containment building. Both sites do, however, have a concrete basemat with a steel liner.

Consistent with the Calvert Cliffs analysis, the following issues are addressed:

• Differences between the containment basemat and the drywell and torus liner

• The historical dryweliltorus steel shell flaw likelihood due to concealed corrosion

• The impact of aging

• The corrosion leakage dependency on containment pressure

• The likelihood that visual inspections will be effective at detecting a flaw Consistent with Calvert Cliffs analysis [241, the following six steps are performed:

1) Determine the historical liner flaw likelihood.

2) Determine aged adjusted liner flaw likelihood.

3) Determine the increase in flaw likelihood between 3, 10 and 15 years.

4) Determine the likelihood of containment breach given liner flaw.

5) Determine the visual inspection detection failure.

6) Determine the likelihood of non-detected containment leakage.

In additions to these steps, the following three additional steps are added to evaluate risk-implication of containment liner corrosion:

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

8) Evaluate the risk impact in terms of LERF.

9) Evaluate the change in conditional containment failure probability.

The quantitative consideration of the containment liner corrosion analysis is discussed in more detail in Appendix B of this report. As can be seen from Appendix B, including corrosion effects in the ILRT assessment would not alter the conclusions from the original internal events analysis. That is, the change in LERF from extending the interval to 15 years from the current 10-year requirement is estimated to be 1.22 x 10s/ry. This value is below the NRC Regulatory Guide 1.174 [5) of 10 7/yr. Therefore, because Regulatory Guide 1.174 [6J defines very small changes in LERF as below 10f7 /yr, increasing the ILRT interval at FitzPatrick from the currently allowed 1-in-10 years to 1-in-15 years and taking into consideration the likelihood of a containment liner flaw due to corrosion is non-risk significant from a risk perspective. Additionally, the dose increase is estimated to be 4.1 x 103 person-rem/ry or 0.57%, and the conditional containment failure probability increase is estimated to be 0.6%. Both of these increases are also considered to be small. As a result, the ILRT interval extension is considered to have a minimal impact on plant risk (including age-adjusted corrosion impacts), and is therefore acceptable.

In addition, a series of parametric sensitivity studies (discussed in more detail in Appendix B of this report) regarding the potential age related corrosion effects on the containment steel liner also predict that even with conservative assumptions, the conclusions from the original internal events analysis would not change.

Figure 2-1 Acceptance Guldelines1 0 for Large Early Release Frequency 16[

10.6 10C-7 RegIo'n-!I

.1O- 10'5 LERtF-1

'° The analysis will be subject to increased technical review and management attention as indicated by the darkness of the shading of the figure. Inthe context of the integrated decsionmaking, the boundaries between regions should not be interpreted as being definitive; the numerical values associated with defining the regions inthe figure are to be interpreted as indicative values only.

Table 2-1 Internal Core Damage Frequency Contributions by Plant Damage States PDS l Simplified Description Point Estimate  % Of I PDS FreI Total CDF 1 Short-term large LOCA with loss of core cooling. Core 1.44 x I04 0.59 damage results at low reactor pressure. Late injection and containment heat removal is available.

2 Short-term medium size LOCA with loss of core cooling. 1.90 x 107' 7.78 Core damage results at reactor pressure. Late injection, and containment heat removal are available.

3 Short-term transient with loss of core cooling. Core damage 1.25 x 10 51.10 results at high reactor pressure. Late injection, and containment heat removal are available.

4 Long-term SBO with battery depletion. Core damage 2.13 x 10 8.75 results at high reactor pressure. All accident-mitigating functions are recoverable when ac power is restored.

5 Short-term SBO with no dc power. Core damage results at 4.58 x 104 1.88 high reactor pressure. All accident-mitigating functions are recoverable when ac power is restored.

6 Long-term SBO involving a loss of high-pressure injection 8.66 x 10 3.55 because of one stuck-open safety relief valve. Core damage results at low reactor pressure. All accident-mitigating functions are recoverable when ac power is restored.

7 Short-term SBO involving a loss of high-pressure injection 6.38 x 10- 0.26 because of two stuck-open safety relief valves. Core damage results at low reactor pressure. All accident mitigating functions are recoverable when ac power is restored 8 Short-term transient with loss of core cooling. Core damage 4.00 x 10 1.64 results at low reactor pressure. Late injection, and containment heat removal are available.

9 Short-term ATWS with one stuck-open safety relief valve 1.62 x 10V 0.66 that leads to early core damage at low reactor pressure following loss of reactivity control. Late injection and containment heat removal is available.

10 Short-term ATWS that leads to early core damage at high 2.81 x 10-' 11.51 reactor pressure following loss of reactivity control. Late injection and containment heat removal is available.

11 Transient with a loss of long-term decay heat removal. 3.00 x 10.' 12.29 Core damage results at high reactor pressure. Late in-vessel and ex-vessel injection is available.

Total j 2.44 x 10 4 100

Table 2-2 Summary of FitzPatrick IPE LEVEL 2 Release Categories 161 l Frequency  % Of Release Category Description UJy) Total CDF Early Low NA 0.00 0.00 Early Medium Low Small early torus failures" 5.04 x 10- 20.66 Early Medium High Large early torus failures or early 5.29 x 10-' 21.70 drywell failures with no core-concrete interactions (drywell flooded)

Early High Early drywell failure with no drywell 1.74 x 10' 7.14 sprays Late Low Vessel breach, no containment failure 9.76 x 10-9 0.004 and no core-concrete interactions Late Medium Low Small late torus failures 4.73 x 10-' 19.40 Late Medium High Large late torus failures 2.71 x 1011 0.00001 Late High Late drywell failure with no drywell 5.82 x 10 2.39 I ______________________ sprays No Containment Failure Containment Intact (no release) 6.91 x 10-' 28.34 Total 2.44 x I04 100 Table 2-3 Summary of FitzPatrick IPE LEVEL 2 Containment Failures [6]

Release Category Frequency %Qo Containment End State I I (ry) Total CDF No Containment Failure No Containment Failure 6.91 x IO-' 28.34 Early Drywell Failure Early High 71.74 x 10- 7.14 Early Wetwell Failure Non-ATWS Early Medium Low, Non- 7.37 x 10-'

ATWS Early Medium High 30.19 Late Drywell Failure Late High 5.82 x le 2.39 Late Wetwell Failure Late Low, Late Medium Low, Late 4.83 x 10-Medium High 19.80 Bypass 12 ATWS Early Medium Low, ATWS Early Medium High 2.97 x 10 12.17 Total 2.44 x 104 100 AdIndudes torus venting.

12 Only ATWS bypass sequences constered, ISLOCA sequences are negligible.

Table 2-4 Collapsed Accident Progression Bins (APB) Descriptions Collapsed Description APB Number CAPB-1 [CD, VB, Early CF, WW, RPV pressure >200 psig at VB]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails either before core damage, during core damage or at vessel breach (Early CF).

The containment failure occurs in the torus (WM), above the water level. RPV pressure is greater than 200 psig at the time of vessel breach (this implies that high pressure induced severe accident phenomena IDCH] is possible).

CAPB-2 [CD, VB, Early CF, WV, RPV pressure <200 psig at VB]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails either before core damage, during core damage or at vessel breach (Early CF).

The containment failure occurs in the torus (WV), above the water level. RPV pressure is less than 200 psig at the time of vessel breach; thus, precluding high pressure induced severe accident phenomena. There are no core concrete interactions (No CCI) due to the present of an overlying pool of water.

CAPB-3 [CD, VB, Early CF, DW, RPV pressure >200 psig at VB]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails either before core damage, during core damage or at vessel breach (Early CF).

The containment failure occurs in the drywell or below the torus water line (DW).

RPV pressure is greater than 200 psig at the time of vessel breach (this implies that high pressure induced severe accident phenomena [DCH] is possible).

CAPB-4 [CD, VB, Early CF, DW, RPV pressure <200 psig at VB]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails either before core damage, during core damage or at vessel breach (Early CF).

The containment failure occurs in the drywell or below the torus water line (DW).

RPV pressure is less than 200 psig at the time of vessel breach (this implies that high pressure induced severe accident phenomena is precluded).

CAPB-5 [CD, VB, Late CF, WMd Core damage (CD) occurs followed by vessel breach (VB). The containment fails late due to a loss of containment heat removal (Late CF). The containment failure occurs in the torus (WM), above the water level. RPV pressure is not important because if a high-pressure severe accident phenomena (such as DCH) occurred, it did not fail containment upon its occurrence.

CD = core damage VB = vessel breach CF = containment failure DW = drywell VV = torus RPV = reactor pressure vessel CCI = core-concrete interactions

Table 2-4 Collapsed Accident Progression Bins (APB) Descriptions (continued)

Collapsed Description APB Number CAPB-6 [CD, VB, Late CF, DWJ Core damage (CD) occurs followed by vessel breach (VB). The containment fails late due to a loss of containment heat removal (Late CF). The containment failure occurs in either the drywell or below the torus water level (DW). RPV pressure is not important, because the occurrence of a high-pressure severe accident phenomenon did not fail containment.

CAPB-7 [CD, VB, Vent]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails because of venting during any of the time periods in the accident.

CAPB-8 [CD, VB, No CF, No CCII Core damage occurs (CD) followed by vessel breach (VB). The containment does not fail structurally and is not vented (No CF). Ex-vessel releases are recovered, therefore precluding the occurrence of core-concrete interactions (No CCI). Although the containment does not fail, vessel breach did occur, therefore the potential exists for some in- and ex-vessel releases to the environment due to containment design leakage. RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurred, it did not fail containment.

CAPB-9 [CD, No VB]

Core damage occurs (CD), but the recovery of RPV injection in time prevents vessel beach (No VB). However, the containment can fail even if vessel breach is averted. Therefore, the potential exists for some of the in-vessel releases to be released to the environment.

CAPB-1 0 (No CD]

Core damage does not occur (No CD). Therefore, containment integrity is not challenged (No CF).

CD = core damage VB= vessel breach CF = containment failure DW = drywell WW = torus RPV = reactor pressure vessel CCI = core-concrete interactions

Table 2-6 Summary of FitzPatrick Baseline Release Frequencies - Given EPRI TR-104285 Accident Class EPRI 1 Class Class Frequency Class I Definition Description I_(fry)

I No Containment Accident sequences in which the containment remains Failure intact and is initially isolated. Only affected by ILRT leak testing frequency due to the incorporation of accident 6.19 x 10-'

classes' 3a and 3b.

2 Large Containment Accident sequences in which the containment isolation Isolation Failures system function fails during the accident progression due to (Failure-to-dose) failures-to-close of large containment isolation valves (>2-inch diameter). This accident class is not affected by ILRT 2.44 x 10-9 leak testing frequency.

3a Small Isolation Accident sequences in which the containment is failed due Failures (Liner to a preexisting small leak in the containment structure or breach) liner that would be identifiable only from an ILRT. 6.59 x 1O4 3b Large Isolation Accident sequences in which the containment is failed due Failures (Liner to a pre-existing large leak in the containment structure or Breach) liner that would be identifiable only from an ILRT. 6.59 x 10-9 4 Small isolation Accident sequences in which the containment is failed due Not Analyzed failure - failure-to- to a pre-existing failure-to-seal of Type B components that seal (Type B test) would not be identifiable from a ILRT.

5 Small isolation Accident sequences in which the containment is failed due Not Analyzed failure - failureto- to a pre-existing failure-to-seal of Type C components that seal (Type C test) would not be identifiable from a ILRT.

6 Containment Accident sequences in which the containment isolation Not Analyzed Isolation Failures system function fails due to mother" pre-existing failure (dependent failures, modes not identifiable by leak rate tests (e.g., pathways left personnel errors) open or misalignment of containment isolation vales following a test/maintenance evolution). Not affected by ILRT leak testing frequency.

7a Severe Accident Accident sequences in which vessel breach occurs and the Phenomena drywell fails either before or at the time of vessel breach. 1.74 x 10-'

Induced Early l____ Drywell Failures 7b Severe Accident Accident sequences in which vessel breach occurs and Phenomena torus fails either before or at the time of vessel breach. 7 37 x 10O' Induced Early Torus Because the drywell does not fail, the entire radionuclide Failures release passes through the torus pool.

7c Severe Accident Accident sequences in which vessel breach occurs, Phenomena however, the drywell does not fail until a late time period. 5.82 x 10 Induced Late Drywell Failures 7d Severe Accident Accident sequences in which vessel breach occurs, Phenomena however, the torus does not fail until a late time period. 4.83 x IO-?

Induced Late Torus Because the drywell does not fail, the entire radionuclide I_____ Failures release passes through the torus pool.

8 Containment Accident sequences in which the containment is bypassed Bypassed (ATWS) (i.e., ATWS with high power oscillations or Interfacing 2.97 x 107' Systems LOCA, ISLOCA).

CDF All Level 2 CET 2.44 x 104 Endstates

Table 2-6 Peach Bottom Unit 2 NUREGICR-4551 Collapsed Accident Progression Bin 50Mile Population Dose PBAPS PBAPS Fractional Population PBAPS APB Population APB Dose Risk Frequencies Dose APB # Contributions' (Person-remlry) 2 (1/ry)3 (Person-rem) 4 1 0.021 0.1659 9.55 x IV 1.74 x 10 2 0.0066 0.05214 4.77 x T sx 1.09 lob 3 0.556 4.3924 1.48 x 10 7 2.97 x 10 4 0.226 1.7854 7.94 x 10-' T 2.25 x lOb 5 0.0022 0.01738 1.30 x le 1.34 x 10 6 0.059 0.4661 2.04 x 10-' 2.28 x leb 7 0.118 0.9322 4.77 x 10' 1.95 x 10 8 0.0005 0.00395 7.99 x 10-' 4.94 x 10i 9 0.01 0.079 3.86 x 10-7 2.05 x 105 10 X 0 0 4.34x10-7 0.0 Totals 1 7.9 4.34x 10 1.38x 10'

1. Obtained from table 5.2-3 of NUREG/CR-4551 [121

2. Derived from the fractional APB contributions times total population dose risk, 0.021 x 7.9 = 0.1659

3. Derived from the conditional probabilities of the APBs and the total internal CDF given in NUREG/CR-4551

[121 (See Step 2.3.3)

4. 0.1659/9.55x10-=1.74x 10 Table 2-7 FitzPatrick Population Doses for the Each APB and Associated Adjustment Factors PBAPS Adjustment Factors JAF Population TS Allowed Population PBAPS Dose Reactor Containment Dose APB # (Person-rem) Population Power Level Leakage Rate (Person-rem)

I 1.74 x 106 0.297 0.770 NA 3.98 x 10 2 T 1.09xl b 0.297 0.770 NA 2.49 x 10 3 2.97 x 10 0.297 0.770 NA 6.79 x 10-4 2.25 x 105 0.297 0.770 NA 5.15 x 10 5 1.34 x10b 0.297 0.770 NA 3.06 x 105 6 i2.28 x l0TO 0.297 0.770 NA 5.21 x 10e 7 1.95x 0 0.297 0.770 NA 4.46 x 10 8 4.94 x 1& 0.297 0.770 2.58 2.91 x le 9 2.05x 10? 0.297 0.770 NA 4.69 x l0e 10 0.0 0.297 0.770 2.58 0.0

Table 2-8 FitzPatrick Collapsed Accident Progression Bins Frequencies Collapsed Description Frequency APB MYAv)

CAPB-1 [CD, VB, Early CF, WVW, RPV pressure >200 psig at VB]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails either before core damage, during core damage or 1.43 x lo, at vessel breach (Early CF). The containment failure occurs in the torus (WW), above the water level. RPV pressure is greater than 200 psig at the time of vessel breach (this implies that high pressure induced severe accident phenomena rDCHI is possible).

CAPB-2 [CD, VB, Early CF, WW, RPV pressure <200 psig at VB]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails either before core damage, during core damage or at vessel breach (Early CF). The containment failure occurs in the 2.27 x 10-torus (W4V), above the water level. RPV pressure is less than 200 psig at the time of vessel breach; thus, precluding high pressure induced severe accident phenomena. There are no core concrete interactions due to the present of an overlying pool of water.

CAPB-3 [CD, VB, Early CF, DW, RPV pressure >200 psig at VB]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails either before core damage, during core damage or at vessel breach (Early CF). The containment failure occurs in the 5.41 x 10 drywell or below the torus water line (DW). RPV pressure is greater than 200 psig at the time of vessel breach (this implies that high pressure induced severe accident phenomena [DCHI is possible).

CAPB-4 [CD, VB, Early CF, DW, RPV pressure <200 psig at VB]

Core damage (CD) occurs followed by vessel breach (VB). The containment fails either before core damage, during core damage or l at vessel breach (Early CF). The containment failure occurs in the 3.86 x 107 drywell or below the torus water line (DW). RPV pressure is less than 200 psig at the time of vessel breach (this implies that high pressure induced severe accident phenomena is precluded).

CAPB-5 [CD, VB, Late CF, WWI Core damage (CD) occurs followed by vessel breach (VB). The containment fails late due to a loss of containment heat removal (Late CF). The containment failure occurs in the torus (WW), above 3.09 x 10-10 the water level. RPV pressure is not important because if a high-pressure severe accident phenomena (such as DCH) occurred, it did not fail containment upon its occurrence.

CD = core damage VB = vessel breach CF = containment failure DW = drywell WW = torus RPV = reactor pressure vessel CCI = core-concrete interactions

Table 2-8 FitzPatrick Collapsed Accident Progression Bins Frequencies (continued)

Collapsed Description Frequency APB MY CAPB-6 [CD, VB, Late CF, DWI Core damage (CD) occurs followed by vessel breach (VB). The 7 containment fails late due to a loss of containment heat removal 1.31 x 10-(Late CF). The containment failure occurs in either the drywell or below the torus water level (DW). RPV pressure is not important, because the occurrence of a high-pressure severe accident l_________ phenomenon did not fail containment.

CAPB-7 [CD, VB, Vent]

Core damage (CD) occurs followed by vessel breach (VB). The 5.45 x le containment fails because of venting during any of the time periods in the accident.

CAPB-8 [CD, VB, No CF]

Core damage occurs (CD) followed by vessel breach (VB). The containment does not fail structurally and is not vented (No CF). Ex-vessel releases are recovered, therefore precluding the occurrence 7 of core-concrete interactions (No CCI). Although the containment 2.92 x 10-does not fail, vessel breach did occur, therefore the potential exists for some in- and ex-vessel releases to the environment due to containment design leakage. RPV pressure is not important because, even though high pressure induced severe accident phenomena (such as direct containment heating [DCH]) occurred, it did not fail containment.

CAPB-9 [CD, No VB]

Core damage occurs (CD), but the recovery of RPV injection in time prevents vessel beach (No VB). Therefore, containment integrity is 4.00 x 10-7 not challenged (No CF) and core-concrete interactions are l ~~~precluded.

CAPB-1 0 [No CD]

0.0 Core damage does not occur (No CD). Therefore, containment l_________ integritv is not challenged (No CF).

CD = core damage VB = vessel breach CF = containment failure DW = drywell VW = torus RPV = reactor pressure vessel CCI = core-concrete interactions

Table 2-9 FitzPatrick Level 2 Endstates Correlation with the Peach Bottom Unit 2 Collapsed Accident Progression Bins From NUEGICR-4551 APB l Description l Release Cateory Containment End State I CD, VB, Early CF, WW, RPV Non-ATWS Early Medium Low, Early Wetwell Failure pressure >200 psi at VB Non-ATWS Early Medium High 2 CD, VB, Early CF, WW, RPV Non-ATWS Early Medium Low, Early Wetwell Failure pressure <200 psig at VB Non-ATWS Early Medium High 3 CD, VB, Early CF, DW, RPV Early High Early Drywell Failure pressure >200 psig at VB 4 CD, VB, Early CF, DW, RPV Early High Early Drywell Failure pressure <200 psig at VB 5 CD, VB, Late CF, WW Late Low, Late Medium Low, Late Wetwell Failure Late Medium High 6 CD, VB, Late CF, DW Late High Late Drywell Failure 7 CD, VB, Vent ATWS Early Medium Low, Bypass ATWS Early Medium High 8 CD, VB, No CF No Containment Failure No Containment Failure 9 CD, No VB, No CF No Containment Failure No Containment Failure 10 No CD No Containment Failure No Containment Failure CD = core damage VB = vessel breach CF = containment failure DW = drywell WW = torus RPV = reactor pressure vessel CCI = core-concrete interactions Table 2-10 FitzPatrick Population Dose Estimates As A Function of EPRI Accident Class within 50-Mile Radius EPRI Accident Class Description Person-Rem Class Within 50 miles I No Containment Failure 2.91 x 103 2 Large Containment Isolation Failures (Failure-to-close) 6.79 x 105 3a Small Isolation Failures (Liner breach) 2.91 x 10 4 3b Large Isolation Failures (Liner Breach) 1.02 x 10° 4 mali isolation failure - failure-to-seal (Type B test) N/A 5 mall isolation failure - failure-to-seal (Type C test) N/A 6 ontainment Isolation Failures (dependent failures, personnel errors) N/A 7a evere Accident Phenomena Induced Early Drywell Failures 6.11 x 10" 7b evere Accident Phenomena Induced Early Torus Failures 3.96 x 10I 7c evere Accident Phenomena Induced Late Drywell Failures 5.21 x 10*

7d evere Accident Phenomena Induced Late Torus Failures 3.06 x 105 8 ontainment Bypassed (ATWS) 4.46 x 105

Table 2-11 FitzPatrick Dose Rates Estimates as a Function of EPRI Accident Class For Population within 50-Miles (Base Line 3 per 10 year ILRT)

Person-Rem Baseline Dose Rate EPRI Accident 5Within 0 Frequency (Person-Class Class Description miles jry) Remlry)

I No Containment Failure 2.91 x 104 6.19 x 10-' 1.80 x 10 2 arge Containment Isolation Failures 6.79 x 1 0 5 2.44 x 10-9 1.66 x 10 (Failure-to-close) 3a Small Isolation Failures (Liner breach) 2.91 x IV 1.92 x 10 3b Large Isolation Failures (Liner Breach) 1.02 x 105 6.59 x 10'9 6.72 x1i0o 4 Small isolation failure - failure-to-seal (Type N/A N/A N/A B test) 5 mall isolation failure - failure-to-seal (Type N/A N/A N/A C test) 6 ontainment Isolation Failures (dependent N/A N/A N/A Failures, personnel errors) 7a evere Accident Phenomena Induced Early 6.11 x 105 1T.74-0 x17.07 1 Drywell Failures 7b evere Accident Phenomena Induced Early 3.96 x 105 7.37 x 1 2.92 x 10 0

Torus Failures 7c evere Accident Phenomena Induced Late 5.21 x 105 5.i82 x 10 - i3.03x10 rywell Failures 7d evere Accident Phenomena Induced Late 3.06 x 105 4.83 x IO-' I.48 x I-"

prywell Failures 8 ontainment Bypassed (ATWS) 4.46 x 105 2.97 x 10-7 1.32 x 10-Total 3.10 x 106 0 7.x Table 2-12 EPRI Accident Class Frequency as a Function of ILRT Interval EPRI Baseline Current Proposed Class (3-per-1O year ILRT) (1-In-10 years ILRT) (1-per-15 year ILRT)

Iry Iry Iry 1 6.19 x le 4.49 x 107 [ 3.29 x le 3a 6.59 x 10 4 2.20 x 10-7 3.29 x 10 3b 6.59 x 109 2.20 x104 3.29 x 10V

Table 2-13 Baseline Dose Rate Estimates By EPRI Accident Class for Population Within 60-Mile Dose Rate as a Function of ILRT Interval (Person-Rem/Rx Year)

Baseline Current Proposed EPRI (3-per-40 (1-per-b (1-in-I5 Class Accident Class Description year ILRT) year ILRT) years ILRT)

I No Containment Failure 1.80 x 1e 1.31 x 10 9.58 x 104 2 Large Containment Isolation Failures 1.66 x 1O- 1.66 x 10 1.66 x 10' (Failure-to-close" 3a Small Isolation Failures (Liner breach) 1.92 x 10- 6.40 x 104 9.60 x 10l 3b Large Isolation Failures (Liner 6.72 x 104 2.24 x 10` 3.36 x 10V Breach) 4 Small isolation failure - failure-to-seal N/A N/A N/A (Type B test) 5 Small isolation failure - failure-to-seal N/A N/A NIA (Type C test)"

6 Containment Isolation Failures N/A N/A N/A (dependent failures, personnel errors) l 7a Severe Accident Phenomena Induced 1.06 x 10-1 1.06 x 10-' 1.06 x 10.'

Early Drywell Failures 7b Severe Accident Phenomena Induced 2.91 x 10'1 2.91 x 10-' 2.91 x 10' Early Torus Failures 7c Severe Accident Phenomena Induced 3.03 x 10= 3.03 x 10- 3.03 x 10Z Late Drywell Failures 7d Severe Accident Phenomena Induced 1.48 x 10-' 1.48 x 10V 1.48 x 10'1 Late Drywell Failures 8 Containment Bypassed (ATWS) 1.32 x 10-' 1.32 x 10-' 1.32 x 10-'

Total 1 0.715 1 0.720 1 0.724

SECTION 3

SUMMARY

OF RESULTS 3.1 Internal Events Impact An evaluation was performed to assess the risk impact of extending the current containment Type A Integrated Leak Rate Test (ILRT) interval. In performing the risk assessment evaluation, the guidance and additional information distributed by NEI in November 2001 to their Administrative Points of Contact

[3,4] regarding risk assessment evaluation of one-time extensions of containment ILRT intervals and the approach outlined in the Indian Point Unit Three Nuclear Power Plant ILRT [8, 9] extension submittal were used. The assessment also followed previous work as outline in NEI 94-01 [1], the methodology used in EPRI TR-104285 [2], and the NRC Regulatory Guide 1.174 [5].

These results demonstrate a very small impact on risk associated with the one time extension of the ILRT test interval to 15 years. The following is a brief summary of some of the key aspects of the ILRT test interval extension risk analysis:

1) The baseline (3-in-10 years) risk contribution (person-rem) associated with containment leakage affected by the ILRT and represented by Classes 3a and 3b accident scenarios is 0.36% of the total risk.

2) When the ILRT interval is 1-in-10 years, the risk contribution of leakage (person-rem) represented by Classes 3a and 3b accident scenarios increases to 1.2% of the total risk.

3) When the ILRT interval is 1-in-15 years, the risk contribution of leakage represented by Classes 3a and 3b accident scenarios increases to 1.8% of the total risk.

4) The increase in risk on the total integrated plant risk as measured by person-rem/reactor year increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.56% (0.004 person-rem/ry).

This value can be considered to be a negligible increase in risk.

5) The risk increase in LERF from reducing the ILRT test frequency from the current once-per-I 0 years to once-per-15 years is 1.09 x 104/ry. This is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174.

6) The risk increase in LERF from the original 3-in-10 years test frequency; to once-per-15 years is 2.63 x 104 /ry. This is also found to be "very small" using the acceptance guidelines in Regulatory Guide 1.174.

7) The change in CCFP of 0.43% is deemed to be insignificant and reflects sufficient defense-in-depth.

8) Other salient results are summarized in Table 3-1. The key results to this risk assessment are those for the 10-year interval (current FitzPatrick ILRT interval) and the I5-year interval (proposed change). The 3-in-10 year ILRT is a baseline starting point for this risk assessment given that the pre-existing containment leakage probabilities (estimated based on industry experience - - refer to Section 1.2) are reflective of the 3-per-10 year ILRT testing.

3.2 External Events Impact This analysis provides an evaluation of external events hazards (seismic, fires, high winds, external flooding, etc) impacts within the framework of the ILRT interval extension risk assessment. Similar to the internal events analysis, the combined impact of internal and external events confirms that the impact (due to the proposed ILRT extension) on the external hazard portion of the FitzPatrick plant risk profile is comparable to that shown for internal events. It is deemed that the calculated risk increase for both internal and external hazards would remain "small".

These results demonstrate a small impact on risk associated with the one time extension of the ILRT test interval to 15 years. The following is a brief summary of some of the key aspects of the ILRT test interval extension risk analysis for the combined internal and external events analysis:

1) The baseline (3-in-10 years) risk contribution (person-rem) associated with containment leakage affected by the ILRT and represented by Classes 3a and 3b accident scenarios is 0.10% of the total risk.

2) When the ILRT interval is 1-in-10 years, the risk contribution of leakage (person-rem) represented by Classes 3a and 3b accident scenarios increases to 0.34% of the total risk.

3) When the ILRT interval is 1-in-15 years, the risk contribution of leakage represented by Classes 3a and 3b accident scenarios increases to 0.5% of the total risk.

4) The combined internal and external events increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.16% (0.038 person-rem/ry). This value can be considered to be a negligible increase in risk.

5) The combined internal and external events risk increase in LERF from reducing the ILRT test frequency from the current once-per-10 years to once-per-15 years is 1.03 x 107 /ry. This is determined to be slightly above the 10 /yr criterion of Region IlIl, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines of Regulatory Guide 1.174.

6) The combined internal and external events change in CCFP of 0.20% is deemed to be insignificant and reflects sufficient defense-in-depth.

7) Other salient results are summarized in Table 3-2.

3.3 Containment Liner Corrosion Risk Impact This analysis provides a sensitivity evaluation of considering potential corrosion impacts within the framework of the ILRT interval extension risk assessment. The analysis confirms that the ILRT interval extension has a minimal impact on plant risk. Additionally, a series of parametric sensitivity studies regarding the potential age related corrosion effects on the steel shell also indicate that even with very conservative assumptions, the conclusions from the original analysis would not change. That is, the ILRT interval extension is judged to have a minimal impact on plant risk and is therefore acceptable.

1) The baseline (3-in-10 years) risk contribution (person-rem) associated with containment leakage affected by the ILRT and represented by Classes 3a and 3b accident scenarios is 0.365% of the total risk.

2) When the ILRT interval is 1-in-10 years, the risk contribution of leakage (person-rem) represented by Classes 3a and 3b accident scenarios increases to 1.21% of the total risk.

3) When the ILRT interval is 1-in-15 years, the risk contribution of leakage represented by Classes 3a and 3b accident scenarios increases to 1.82% of the total risk.

4) The age-adjusted corrosion impact on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-l0 years test interval to a 1-in-15 years test interval, is found to be 0.57% (0.0041 person-remlry). This value can be considered to be a negligible increase in risk.

5) The age-adjusted corrosion impact risk increase in LERF from reducing the ILRT test frequency from the current once-per-10 years to once-per-15 years is 1.22 x 10I/ry. This is determined to be below the 1O7/yr criterion of Region Ill, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines of Regulatory Guide 1.174.

6) This age-adjusted corrosion impact change in CCFP of 0.60% is deemed to be insignificant and reflects sufficient defensesin-depth.

7) Other results (taken from Appendix B) of the updated ILRT assessment including the potential impact from non-detected containment leakage scenarios assuming that 100% of the leakages result in EPRI Class 3b are show in Table 3-3.

Additional sensitivity cases were also developed to gain an understanding of the containment liner corrosion sensitivity to various key parameters. The sensitivity cases are as follows:

• Sensitivity Case I - Flaw rate doubles every 2 years

• Sensitivity Case 2 - Flaw rate doubles every 10 years

• Sensitivity Case 3 - 5% Visual inspection failures

• Sensitivity Case 4 - 15% Visual inspection failures

• Sensitivity Case 5 - Containment breach base point 10 times lower

• Sensitivity Case 6 - Containment breach base point 10 times higher

• Sensitivity Case 7 - Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF (Lower bound)

• Sensitivity Case 8 - Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF (upper bound)

The results of the containment liner corrosion sensitivities cases, taken from Appendix B are summarized in Table 3-4.

Table 31 Summary of Risk Impact on Extending Type A ILRT Test Frequency - Internal Events Base Case Extend to Extend to 3 Years 10 Years 15 Years EPRI CDF Per-Rem Per-Rem CDF I Per-Rem Per-Rem CDF Per-Rem Per-Rem Class (Per ry) (Per ry) (Per ry) (Per ry) (Per my) (Per ry) 1 6.19x10-7 2.91 x103 1.80x10-3 4A9x10-7 2.91 x103 1.31 x10-3 3.29x10- 7 2.91 x103 9.58x10 4 2 2.44x IO-9 6.79x 105 1.66x 10-3 2.44x I0-9 6.79x105 1.66x 10 4 2.44x 10- 6.79x I05 1.66x 10-3 3a 6.59 x 10.0 2.91 x 10 4 1.92 x 10-3 2.20 x 10-7 2.91 x I04 6.40 x 104 3.29 x l0e 2.91 x 10 4 9.60 x 10 3 3b 6.59 x 10-9 1.02 x 105 6.72 x 10 4 2.20 x 10 4 1.02 x 105 2.24 x 10 4 3.29 x 104 1.02 x 105 3.36 x 10-3 4 0.00 N/A 0.00 0.00 N/A 0.00 0.00 NWA 0.00 5 0.00 N/A 0.00 0.00 N/A 0.00 0.00 NWA 0.00 6 0.00 N/A 0.00 0.00 N/A 0.00 0.00 NIA 0.00 7a 1.74x10-7 6.11 x 106 1.06x 10- 1.74x10-7 6.11 x 105 1.06x 10- 1.74x 10 7 6.11 x 10 5 1.06X10-1 7b 7.37 x 10-7 3.96 x 105 2.91 x 10-' 7.37 x 10-7 3.96 x 10 5 2.91 x 10-1 7.37 x 10 7 3.96 x 105 2.91 x 10-'

7c 5.82 x 104 5.21 x 105 3.04 x 10-2 5.82 x 104 5.21 x 105 3.04 x 10-2 5.82 x 104 5.21 x105 3.04 x 10-2 7d 4.83 x 10 7 3.06 x 10 5 1.48 x 10' 4.83 x 10-7 3.06 x 10 5 1A8 x 101: 4.83 x 10- 7 3.06 x 105 1.48 x 10' 8 2.97 x 10-7 4.46 x 106 1.32 x 10 2.97 x 10-7 4.46 x 105 1.32 x 10' 2.97 x 10 7 4.46 x 105 1.32 x 10 Total 2.44 x 10 7.1 x 10- 2.44 x 10 4 7.20 x10' 2.44x104 7.24 x10' ILRT Dose Rate 2.59 x 10-3 8.64 x 10-3 1.30 x 10-2 from 3a and 3b

% Of Tota 0.36% 1.2% 1.8%

Delta Dose Rate 3.97 x I0 4 from 3a and 3b (10 to 15 yr)

LERF from 3b 6.59 x 10 9 2.20 x 104 3.29 x 104 Delta LERF 1.09 x I0 4 (10 to 15 yr)

CCFP % 72.0% 72.6% 73.1%

Delta CCFP % 0.43%

(10 to 15 yr)

Table 3-2 Summary of Risk Impact on Extending Type A ILRT Test Frequency - Effect of Internal and External Events Risk on FitzPatrick ILRT Risk Assessment Base Case Extend to Extend to 3 Years 10 Years _ 15 Years EPRI CDF Per-Rem l Per-Rem CDF l Per-Rem I Per-Rem CDF lPerRem l Per-Rem Class (Per _y) (Per Ry) (Per Ry) IPer (Per Ry) I (Per Ry) 1 4.28 x 104 2.91 x 103 1.25 x 10-2 2.69 x 104 2.91 x 103 7.85 x 10-3 1.56 x 104 2.91 x 103 4.55 x 10-3 2 5.19x104 6.79x105 3.53x10,2 5.19x105 6.79x105 3.53x10-2 5.19x10-5 6.79x10 5 3.53x10 2 3a 6.18 x 10-' 2.91 x 10 4 1.80 x 10-2 2.06 x 104 2.91 x 104 6.00 x 10-2 3.09 x 104 2.91 x I04 9.00 x 10-2 3b 6.18x10- 1.02x10 5 6.30x10-3 2.06x10'7 1.02x10 5 2.10x10-2 3.09x10-7 1.02x105 3.15x10-2 4 0.00 N/A 0.00 0.00 N/A 0.00 0.00 N/A 0.00 5 0.00 N/A 0.00 0.00 N/A 0.00 0.00 N/A 0.00 6 0.00 N/A 0.00 0.00 N/A 0.00 0.00 N/A 0.00 7a 2.45 x 10-5 6.11 x 105 1.50 x 10' 2.45 x 10-5 6.11 x 105 1.50 x 10' 2.45 x 10-5 6.11 x 105 1.50 x 10' 7b 1.57 x 10-5 3.96 x 105 6.20 1.57 x 10-5 3.96 x 105 6.20 1.57 x 10-5 3.96 x 105 6.20 7c 1.24 x 104 5.21 x 105 6.46 x 10.1 1.24 x 104 5.21 x 105 6.46 x 10-' 1.24 x 104 5.21 x 105 6.46 x 101' 7d 1.82 x 104 3.06 x 105 5.58 x 107' 1.82 x 104 3.06 x 105 5.58 x 10-' 1.82 x 104 3.06 x 105 5.58 x 10-'

8 3.65 x 104 4.46 x 105 1.63 3.65 x 104 4.46 x 105 1.63 3.65 x 104 4.46 x 105 1.63 Total 5.19 x 105 2.408 x 10' 5.19 x 10-5 2.413 x 10' 5.19 x 10-5 2.417 x 10'

- -~~~~~~~~~~~~~~~~~~~~~1-ILRT Dose Rate 2.43 x 10-2 8.10 x 10.2 1.22 x 10-'

from 3a and 3b

% Of Total 0.1009% 0.3357% 0.5027%

Delta Dose Rate 3.80 x 10-2 from 3a and 3b (10to 15yr)

LERF from 3b 6.18 x 10-8 2.06 x 10-7 3.09 x 10-7 Delta LERF 1.03 x 10-.

(10 to 15 yr)

CCFP *h 90.57% 90.85% 91.05%

Delta CCFP % 0.198%

(10 to 15 vr)

Table 3-3 Summary of Risk Impact on Extending Type A ILRT Test Frequency - Impact of Containment Steel Liner Corrosion on FitzPatrick ILRT Intervals

- Base Case Extend to Extend to 3 Years _ 10 Years 15 Years EPRI CDF Per-Rem Per-Rem CDF Per-Rem Per-Rem CDF Per-Rem Per-Rem Class (Per Ry) (Per r Ry) I____RPe (Per Ry) (Per Ry) 1 6.18 x 10-7 2.91 x 10 3 1.80 x 103 4.49 x IV- 2.91 x 103 1.31 x 10-3 3.26 x 1iO 2.91 x 103 9.52 x 104 2 2.44 x 10-9 6.79 x 105 1.66 x 10-3 2.44 x 10-9 6.79 x 105 1.66 x 10-3 2.44 x 10-9 6.79 x 105 1.66 x 104 3a 6.59 x 104 2.91 x 104 1.92 x 10-3 2.20 x 10-' 2.91 x 104 6.40 x 10-3 3.29 x I O' 2.91 x 104 9.60 x 10-3 3b 6.75 x 10l 1.02 x 105 6.88 x 10 4 2.29 x 104 1.02 x 105 2.34 x 10-3 3.51 x 104 1.02 x 105 3.58 x 10-3 4 0.0 N/A 0.0 0.0 NtA 0.0 0.0 N/A 0.0 5 0.0 N/A 0.0 0.0 N/A 0.0 0.0 N/A 0.0 6 0.0 N/A 0.0 0.0 N/A 0.0 0.0 N/A 0.0 7b 7 1.74x10 6.11 x105 1.06x10-' 1.74x10- 6.11 x105 1.06x10-' 1.74x10-7 6.11 x105 1.06x1O-'

7b 7.37 x IO7 3.96 x 10 5 2.91 x 10.' 7.37 x 10-' 3.96 x 105 2.91 x 101' 7.37 x I0- 3.96 x 10 5 2.91 x 101' 7c 5.82 x 10- 5.21 x 10 5 3.04 x 10.2 5.82 x 10-8 5.21 x 105 3.04 x 10-2 5.82 x I0-8 5.21 x 105 3.04 x 10-2 7d 4.83 x 10-7 3.06 x 105 1.48 x 101 ' 4.83 x 10-7 3.06 x 10 5 1.48 x 10' 4.83 x 10-7 3.06 x 105 1.48 x 101' 8 2.97 x 10- 4.46 x 105 1.32 x 10-' 2.97 x 107' 4.46 x 105 1.32 x 101' 2.97 x 107 4.46 x 10 5 1.32 x 10-1 Total 2.44 xo10 4 7.1 x -' 2.44x 7.20 x 10' 2.44 x 104 = 7.24 x 10-'

ILRT Dose Rate 2.61 x 104 8.74 x 103 1.32 x 10-2 from 3a and 3b (+1.63 x 10- (+9.48 x 1- (+2.21E-04)

% Of Total 0.365% 1.2128% 1.8199%

(+0.0023%) (+0.0130%) (+0.0300%)

Delta Dose Rate 4.09 x 10J from 3a and 3b (+0.0123%)

(10 to 15 yr) __

LERF from 3b 6.75 x 10-9 2.29 x 10U 3.51 x 101

((+1.60x 10Or (+9.29 x 10-1) (+2.17 x 1C)

Delta LERF 1.22 x j04 (10 to 15 yr) (+1.24 x 10-9)

CCFP

• 71.99% 72.65% 73.15%

(+0.0065%) (+0.0381%) (+0.0889%)'

Delta CCFP % 0.60%

(10 to 15 yr (+0.0508%)

Denotes increase from original values presented in Section 2.4, Steps 7,8, and 9 of this report.

Table 3-4 Containment Steel Liner Corrosion Sensitivity Cases Visual LERF LERF LERF Total LERF Drywell Inspection Likelihood Increase Increase Increase Increase Age Torus & Non- Flaw is LERF From From From From ILRT (Step 2) Breach Visual (EPRI Class Corrosion Corrosion Corrosion Extension (Step 4) Flaws 3b) (3-in-10 (1-in-10 (I to 15 (10 to 15 (Step 5) years) years) years) years)

Base Case Base Case Base Case Base Case Base Case Base Case Base Case Base Case Doubles 0.8171%liner 10% 100% 1.60 x 10-10 9.29 x 10-'0 2.17 x 10-9 1.22 x 10 4 every 5 yrs .0817%hfloor Doubles Base Base Base 4.57 x 10-1 7.75 x 10.-' 4.50 x 1i0 1.47 x 10-'

every 2 yrs Doubles Base Base Base 2.37 x I0O'° 3.25 x 10.'0 4.20 x 10-'. 1.11 x 10.8 every IOyrs Base Base 5% Base 1.53x 10-'° 6.20 x 10-" 1.45x 10-'° 1.11 x104 Base Base 15% Base 1.67 x 10-' 1.45 x 10-'° 3.38 x 10-' 1.12 x 104 Base ).166%line,' 3 Base Base 3.25 x 10-" 1.89 x 10-'. 4.42 x 101-' 1.12 x 1048

).017%floor Base 4.07% line r 4 Base Base 7.95 x 1O-'° 4.63 x 10-9 1.08 x 104 1.72 x I10 0.411 %hfloor4 1 III_ I Lower Bound Doubles Base 5% 10% 4.62 x 10-12 2.00 x 10-" 3.66x 101" 1.IOE-08 every IO yrs 5 Upper Bound Doubles I Base I 15% I 100% I 2.38 x 1010 l 4.03 x 10 l 2.34 x 104-T 3.04E-08 every 2 yrs I I I I I I 3

Base point 10 times lower dian base case of0.0001 at 20 psia 4 Base point 10 times higher than base case of 0.01 at 20 psia.

SECTION 4 CONCLUSIONS 4.1 Internal Events Imnact A risk assessment of the impact of changing FitzPatrick Integrated Leak Rate Test (ILRT) interval from the currently approved 1-in-10 year interval to a one-time extension to 1-in-15 years has been performed.

Based on the above results, the following are main conclusions regarding the assessment of the plant risk associated with extending the Type A ILRT test frequency from ten-years to fifteen years:

1. Regulatory Guide 1.174 [6] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 [6] defines very small changes in risk as resulting in increases of CDF below 10 4 /yr and increases in LERF below 1eyr. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 1-in-10 years to 1-in-15 years is 1.09 x 10o/ry. Since Regulatory Guide 1.174 [61 defines very small changes in LERF as below 107 /yr, increasing the ILRT interval at FitzPatrick from the currently allowed one-in4en years to one-in-fifteen years is non-risk significant from a risk perspective.

2. The increase in risk on the total integrated plant risk as measured by person-rem/reactor year increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.56% (0.004 person-rem/ry).

This value can be considered to be a negligible increase in risk.

3. The change in conditional containment failure probability (CCFP) is calculated to demonstrate the impact on 'defense-in-depth'. The ACCFPo. 15 is found to be 0.43%. This signifies a very small increase and represents a negligible change in the FitzPatrick containment defense-in-depth.

Table 4-1 summarizes the above conclusions.

4.2 External Events Impact Based on the results from Appendix A, *External Event Assessment During an Extension of the ILRT Interval, the following are main conclusions regarding the assessment of the plant risk associated with extending the Type A ILRT test frequency from ten-years to fifteen years:

1. Based on conservative methodologies in estimating the core damage frequency for internal events, seismic events, and fires events, the ALERFcoMBINEDIOs5 of 1.03 x 10 /ry from extendin7 g the FitzPatrick ILRT frequency from 1-in-10 years to 1-in-15 years is slightly above the 10. /yr criterion of Region Ill, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [6]. Therefore, increasing the ILRT interval at FitzPatrick from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.

2. The combined internal and external events increase in risk on the total integrated plant risk as measured by person-rem/reactor year increases for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.16% (0.038 person-remlyr). This value can be considered to be a negligible increase in risk.

3. The change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-15 years is 0.20%. A change in ACCFP of less than 1% is insignificant from a risk perspective.

Table 4-2 summarizes the above conclusions.

4.3 Containment Liner Corrosion Risk Impact Based on the results from Appendix B, "Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval,' the following are main conclusions regarding the assessment of the plant risk associated with extending the Type A ILRT test frequency from ten-years to fifteen years:

1. The impact of including age-adjusted corrosion effects in the ILRT assessment has minimal impact on plant risk and is therefore acceptable.

2. The change in LERF, taking into consideration the likelihood of a containment liner flaw due to age-adjusted corrosion is non-risk significant from a risk perspective. Specifically, extending the interval to 15 years from the current 10 years requirement is estimated to be about 1.22 x 10'/ry. This is below the Regulatory Guide 1.174 [6] acceptance criteria threshold of I 7 /yr.

3. The age-adjusted corrosion impact in dose increase is estimated to be 4.1 x 103 person-rem/ry or 0.57% from the baseline ILRT 10 years interval.

4. The age-adjusted corrosion impact on the conditional containment failure probability increase is estimated to be 0.6%.

5. A series of parametric sensitivity studies regarding potential age related corrosion effects on the containment steel liner also demonstrated minimal impact on plant risk.

Table 4-3 summarizes the above conclusions.

Ask I~ REPORT NO. JAF-RPT-03-00007 I Revision 0 1Page I Il Il 0l i

Table 4-1 Quantitative Results as a Function of ILRT Interval - Internal Events Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1-per-15 year ILRT)

Dose Population Dose Population Dose (Person-Rem Accident Rate (Person- Accident Rate (Person-EPRI Within 50 Frequency Rem / Ry Within Frequency Rem / Ry Within Class Category Description miles)"1 ) (perry) 50 miles) (per ry) 50 miles)

I No Containment Failure M 2.91 x lo 4.49 x 10-7 1.31 x 10'3 3.29 x 10'7 9.58 x IO 2 9 Containment Isolation System Failure 6.79 x 105 2.44 x 1ia 1.66 x 10-3 2.44 x 109 1.66 x 10'3 3a Small Pre-Etxsting Failures (2),(3) 2.91 x 104 2.20 x 10- 6.40 x 10-3 3.29 x 10'7 9.60 x 10-3 3b Large Pre-Existing Failures (2).(3) 1.02 x 105 2.20 x 104 2.24 x 10-3 3.29 x 108 3.36 x 10'3 4 Type B Failures (LLRT) N/A 0.00 0.00 0.00 0.00 5 Type C Failures (LLRT) N/A 0.00 0.00 0.00 0.00 6 Other Containment Isolation System Failure N/A 0.00 0.00 0.00 0.00 7a Containment Failure Due to Severe Accident (a)(4) 6.11 x 105 1.74 x 10-7 1.06 x 10.1 1.74 x 10'7 1.06 x 10-1 7b Containm ent Failure Due to Severe Accident (b) (4) 3.96 x 105 7.37 x IO' 2.91 x 10.1 7.37 x 107 2.91 x 10.

7c Containment Failure Due to Severe Accident (c)(4) 5.21 x 105 5.82 x 104 3.04 x 10.2 5.82 x 104 3.04 x 10-2 5

7d Containment Failure Due to Severe Accident (d) (4) 3.06 x 10 4.83 x 10-7 1.48 x 10 1 4.83 x 10'7 1.48 x 10.1 8 Containment Bypass Accidents 4.46 x 105 2.97 x 10e 1.32 x 10-' 2.97 x 10-7 1.32 x 10-.

TOTALS: 2.44 x 104 0.720 2.44 x 10-8 0.724 Increase in Dose Rate 0.56%

Increase in LERF 1.09x 10.8 Increase in CCFP (%) _ 0.43%

t'~E~"Ente rgy REPORT No. JAF-RPT-03-00007 Revision 0 l Page I 75 Io, I 80 I Table 4-2 Quantitative Results as a Function of ILRT Interval - Internal and External Events Quantitative Results as a Function of ILRT Interval Current Proposed I

(1-per-10yearILRT) (1-per.15yearILRT)

Dose Population Dose Population Dose (Person-Rem Accident Rate (Person- Accident Rate (Person-EPRI Within 50 Frequency Rem / Ry Within Frequency Rem / Ry Within Class Category Description miles)°1' (perry) 50 miles) (per ry) 50 miles) 1 No Containment Failure (2 2.91 x 103 2.69 x 104 7.85 x 103 1.56 x 104 4.55 x 10-3 2 Containment Isolation System Failure 6.79 x IO" 5.19 X 104 3.53 x 10.2 5.19 x 10-l 3.53 x 10 2 3a Small Pre-Existing Failures (2)(3) 2.91 x 104 2.06 x 104 6.00 x 102 3.09 x le 9.00 x 10.2 3b Large Pre-Existing Failures (2),(3) 1.02 x 105 2.06 x 10'7 2.10 x 10- 2 3.09 x 10-7 3.15 x 10-2 4 Type B Failures (LLRT) N/A 0.00 0.00 0.00 0.00 5 Type C Failures (LLRT) N/A 0.00 0.00 0.00 0.00 6 Other Containment Isolation System Failure N/A 0.00 0.00 0.00 0.00 7a Containment Failure Due to Severe Accident (a)(4) 6.11 x lo, 2.45 x 10 1.50 x 101 2.45 x 10'5 1.50 x 101 7b Containment Failure Due to Severe Accident (b)(4) 3.96 x 105 1.57 x 10-5 6.20 1.57 x 10 5 6.20 7c Containment Failure Due to Severe Accident (c)(4) 5.21 x IO' 1.24 x 104 6.46 x 101' 1.24 x 104l 6.46 x 10.1 7d Containment Failure Due to Severe Accident (d)(4) 3.06 x 105 1.82 x 104 5.58 x 10-1 1.82 x 1046 5.58 x 10.

8 Containment Bypass Accidents 4.46 x 10 5 3.65 x 104 1.63 3.65 x 104 1.63 TOTALS: 5 5.19x 10- 24.13 5.19x 10' 5 24.17 Increase in Dose Rate _ _ __ ___________ 0.16%

Increase inLERF 1.03 x I 0r Increase in CCFP (%)

Askdf~e~

REPORT NO. JAF-RPT-03-00007 I Revision 0 l Page l 76 l Of I 80 I a

Table 4-3 Quantitative Results as a Function of ILRT Interval - Liner Corrosion Impact Quantitative Results as a Function of ILRT Interval Current Proposed (1-per-10 year ILRT) (1-per-15yearlLRT)

Dose (Person-Rem Accident I Population Dose Rate (Person- Accident Population Dose Rate (Person-EPRI Within 50 Frequency Rem / Ry Within Frequency Rem / Ry Within Class Category Description miles)( 1} (per ry) 50 miles) (per ry) 50 miles)

I No Containment Failure 2.91 x 103 4.49 x l0e 1.31 x 104 3.26 x 10-7 9.52 x 104 2 Containment Isolation System Failure 6.79 x 105 2.44 x 10-9 1.66 x 103 2.44 x 10-9 1.66x 10 3 3a Small Pre-Existing Failures M(3) 2.91 x 104 2.20 x 10-7 6.40 x 104 3.29 x 10-7 9.60 x 10-3 3b Large Pre-Existing Failures (2)(3) 1.02 x 105 2.29 x 104 2.34 x 104 3.51 x io 4 3.58 x 10,3 4 Type B Failures (LLRT) N/A 0.OOE+00 0.00E+00 0.OOE+00 O.OOE+00 5 Type C Failures (LLRT) N/A 0.OOE+00 0.OOE+00 0.OOE+00 O.OOE+00 6 Other Containment Isolation System Failure N/A 0.OOE+00 0.O0E+00 0.OOE+00 0.O0E+00 7a Containment Failure Dueto SevereAccident(a)(4) 6.11 x lo, 1.74x le 1.06x 101' 1.74x 10i7 1.06 x 10 7b Containment Failure Due to Severe Accident (b)(4) 3.96 x 105 7.37 x le 2.91 x 10.1 7.37 x I07 2.91 x l0-7c Containment Failure Due to Severe Accident (c)(4) 5.21 x 105 5.82 x 104 3.04 x 10.2 5.82 x 10. 3.04 x 10.2 5

7d Containment Failure Due to Severe Accident (d)(4) 3.06 x 10 4.83 x 10-7 1.48 x 101- 4.83 x 10-7 1.48 x 10" 8 Containment Bypass Accidents 4.46x 105 2.97 x 10-7 1.32 x 10.1 2.97 x 10 7 1.32 x 10' TOTALS: 2.44 x 10 4 0.720 2.44 x104 0.724 Increase In Dose Rate _=_=_ _0.57%

Increase in LERF _______________ 1.22x 10.8 Increase in CCFP (%) _ 0.60%

Notes to Tables 15. 16. and 17:

1) The population dose associated with the Technical Specification Leakage is based on scaling the population data, the power level, and allowable Technical Specification leakage compared to the Peach Bottom Unit NUREG/CR-4551 [12] reference plant.

2) Only EPRI accident classes 1, 3a, and 3b are affected by ILRT (Type A) interval changes.

3) Dose estimates for EPRI Class 3a and 3b, per the NEI Interim Guidance, are calculated as 10 times EPRI Class 1 dose and 35 times EPRI Class 1 dose, respectively.

4) EPRI Class 7, containment failure due to severe accident, was subdivided into four subgroups based on FitzPatrick Level 2 containment failure modes for dose allocation purposes. Note that this EPRI class is not affected by ILRT interval changes.

SECTION 5 REFERENCES (1) Nuclear Energy Institute, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J", NEI 94-01, July 26, 1995, Revision 0.

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

(3) Letter from A. Petrangelo (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.

(4) Letter from A. Petrangelo (NEI) to NEI Administrative Points of Contact, 0One-Time Extension of Containment Integrated Leak Rate Test Interval - Additional Information", November 30, 2001.

(5) 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, November 2002, Revision 1.

(6) New York Power Authority, 'James A. FitzPatrick Nuclear Power Plant Individual Plant Examination,' (JAF-RPT-MULTI-02107), April 1998, Revision 1.

(7) U.S. Nuclear Regulatory Commission, "Performance-Based Containment Leak-Test Program",

NUREG-1493, September 1995.

(8) Entergy Nuclear Northeast, 'Risk Impact Assessment of Extending Containment Type A Test Interval", Calculation Number, IP3-CALC-VC-03357, January 3, 2001 Revision 0.

(9) U. S. Nuclear Regulatory Commission, Indian Point Nuclear Generating Station Unit No. 3 -

Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing, April 17, 2001.

(10) Florida Power - Progress Energy, Crystal River Nuclear Plant Letter of June 20, 2001,

'Supplemental Risk Informed Information in Support of License Amendment Request No. 267".

(11) U.S. Nuclear Regulatory Commission, *Severe Accident Risks: An Assessment for Five U.S.

Nuclear Power Plants, NUREG-1150', December 1990.

(12) Payne et al., "Evaluation of Severe Accident Risks: Peach Bottom, Unit 2," NUREG/CR-4551, SAND86-1309, Volume 4, Rev 1, Part 1, December 1990.

(13) NRC letter to James A. FitzPatrick Nuclear Power Plant issuing Technical Specification Amendment to 234 to implement the requirements of 10 CFR 50, Appendix J, Option B, dated October 4, 1996.

(14) Entergy Nuclear Northeast, "Risk Impact Assessment of Extending Containment Type A Test Interval", Calculation Number, JAF-CALC-PC-04544, June 2, 2003 Revision 0.

(15) KLD Associates, Inc., 'Nine Mile Point / James A. FitzPatrick Nuclear Facilities Development of Evacuation Time Estimates,* Appendix M 'Ingestion Pathway Population," KLD-TR-370, June 2003, Revision 0.

(16) Entergy Nuclear Northeast "James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis," Chapter One, Revision 4, May 2001.

(17) NRC letter to James A. FitzPatrick Nuclear Power Plant issuing Technical Specification Amendment to 234 to implement the requirements of 10 CFR 50, Appendix J, Option B, dated October 4, 1996.

(18) Entergy Nuclear Northeast *James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis," TABLE 5.2-1, Revision 2, May 2001.

(19) United States Nuclear Regulatory Commission, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10CFR 50.54(f)," Generic Letter 88-20, Supplement 4, June 28, 1991.

(20) New York Power Authority, "James A. FitzPatrick Nuclear Power Plant Individual Plant Examination for External Events," (JAF-RPT-MISC-0221 1), June 1996, Revision 0.

(21) United States Nuclear Regulatory Commission, "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities,"

NUREG-1407, June 1991.

(22) Electric Power Research Institute, "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," EPRI NP-6041 -SL, Revision 1, August 1991.

(23) Parkinson, W. J., "EPRI Fire PRA Implementation Guide", prepared by Science Applications International Corporation for Electric Power Research Institute, EPRI TR-105928, December 1995.

(24) 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, March 27, 2002.

(25) Entergy Nuclear Northeast, "James A. FitzPatrick Nuclear Power Primary Containment Leakage Rate Testing Program Plan," (JAF-RPT-PC-02342), September 2002, Revision 5.

(26) R.P Kennedy, "Overview of Methods for Seismic PRA and Margin Analysis Including Recent Innovations", Proceedings of the OECD-NEA Workshop on Seismic Risk, Tokyo, Japan, August 1999.

(27) Electric Power Research Institute, "Probabilistic Seismic Hazard Evaluation at Nuclear Plant Sites in the Central and Eastern United States: Resolution of the Charleston Issue," EPRI NP-6395-D, April 1989.

(28) Letter from United States Nuclear Regulatory Commission to James Knubel on "James A.

FitzPatrick Nuclear Power Plant - Review of FitzPatrick Individual Plant Examination of External Events (IPEEE) SUBMITTAL (TAC NO. M83622).

(29) Entergy Nuclear Northeast, "First Containment Inservice Inspection Interval Program and Schedule," (JAF-RPT-PC-04088), January 2003, Revision 3.

(30) Michael Griesmeyer and L.N. Smith, "A Reference Manual for the Event Progression Analysis Code (EVNTRE)-Section 8," Science Applications International Corporation, NUREG/CR-5174, SAND88-1607, September 1989.

Appendix A External Event Assessment During an Extension of the ILRT Interval

Table of Contents Page No.

A1.0 Introduction A-2 A2.0 FitzPatrick IPEEE Seismic Analysis A-2 A2.1 Seismic Analysis Methodology Selection A-2 A2.2 Seismic Analysis Assumptions A-2 A2.3 Seismic Analysis Input A-3 A2.4 Seismic Analysis Method of Analysis A-3 A3.0 FitzPatrick IPEEE Fire Analysis A-6 A3.I Fire Analysis Methodology Selection A-6 A3.2 Fire Analysis Method of Analysis A-6 A4.0 Other External Hazards A-7 A5.0 Effect of External Events Hazard Risk on ILRT Risk Assessment A-7 A5. 1 Effect of External Events Hazard Risk on ILRT Risk Assessment Assumptions A-7 A5.2 Effect of External Events Hazard Risk on ILRT Risk Assessment Method of Analysis A-7 A5.3 Effect of External Events Hazard Risk on ILRT Risk Assessment Input A-8 A6. Conclusions A-16 Tables Page No.

Table A JAF EPRI Site Seismic Hazard Curve A-3 Table A Effect of External Events Hazard Risk on FitzPatrick ILRT Risk Assessment A-17 Table A Effect of Internal and Extemal Events Risk on FitzPatrick ILRT Risk Assessment A-18 A-l

A1.0 Introduction This appendix discusses the risk-implication associated with external hazards in support of the FitzPatrick Integrated Leak Rate Testing (ILRT) interval extension risk assessment.

In response to Generic Letter 88-20, Supplement 4 [21], FitzPatrick submitted an Individual Plant Examination of External Events (IPEEE) in June 1996 [20]. The IPEEE was a review of external hazard risk (i.e., seismic, fires, high winds, external flooding, etc) to identify potential plant vulnerabilities and to understand severe accident risks. The results of the FitzPatrick IPEEE are therefore used in this risk assessment to provide a comparison of the effect of external hazards when extending the current 1-in-1 0 years to 1-in-15 years Type A ILRT interval.

A2.0 FitzPatrick IPEEE Seismic Analysis A2.1 Seismic Analysis Methodology Selection The FitzPatrick plant has been designed to accommodate a safe-shutdown earthquake (SSE) with 0.15g-peak ground acceleration. The seismic analysis performed in the IPEEE study is intended to act as a performance check on the design, estimating seismic capacity beyond the SSE.

The seismic analysis methodology implemented for FitzPatrick satisfied the NRC requirements for performing a seismic IPEEE as presented in Generic Letter 88-20, Supplement 4 1211. The methodology comprises an NRC seismic margin assessment (SMA) following the guidance of NUREG-1407 "Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities," [21] and EPRI NP-6041 "A Methodology for Assessment of Nuclear Power Plant Seismic Margin," [22] and a containment performance analysis. A seismic margin can be expressed in terms of the earthquake motion level that compromises plant safety-the seismic margin assessment determines whether there is high confidence that the plant can survive a given earthquake.

No core damage frequency sequences were quantified as part of the IPEEE seismic risk analysis.

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

1. The overall plant HCLPF (High Confidence Low Probability of Failure) capacity at JAF is 0.22g PGA.

2. No unique decay heat removal vulnerabilities to seismic events at full power operation were found.

Because the overall plant HCLPF capacity with respect to decay heat removal is estimated to be 0.30g PGA, it can be concluded that the decay heat removal pathways are seismically robust with a considerable margin above the 0.15g safe shutdown design basis earthquake.

3. Seismic-induced flooding does not pose major risks.

4. Seismic-induced fires do not pose major risks.

5. No unique seismic induced containment failure mechanisms were identified.

A2.2 Seismic Analysis Assumptions

1) The Simplified Hybrid Method as presented in OECD-NEA Workshop on Seismic Risk, 'Overview of Methods for Seismic PRA and Margin Analysis Including Recent Innovations' 126] is used to approximate the FitzPatrick seismic-induced core damage frequency based on the seismic margin analysis results found in the FitzPatrick IPEEE submittal [20].

A-2

A2.3 Seismic Analysis Input

1) The FitzPatrick external events individual HCLPF values for each of the seven seismic accident types are as follows [20]:

Seismic Accident T1De Individual HCLPF Station Blackout 0.22g PGA Structural 0.30g PGA Loss-of-OfFaite Power 0.27g PGA Small LOCA 0.27g PGA Medium LOCA 0.22g PGA Large LOCA 0.31g PGA Loss-of-Containment Heat Removal 0.30g PGA

2) The 10% NEP standard normal variable is -1.282. This value is derived from the OECD-NEA Workshop on Seismic Risk [26].

3) The 1% NEP standardized normal variable is -2.326. This value is derived from the OECD-NEA Workshop on Seismic Risk [26].

4) The Simplified Hybrid method presented in OECD-NEA Workshop on Seismic Risk [26] recommends a variable factor p equal to 0.3 to estimate the plant damage seismic risk.

5) The seismic hazard curve for the FitzPatrick site, based upon EPRI NP-6395-D, "Probabilistic Seismic Hazard Evaluation at Nuclear Plant Sites in the Central and Eastern United States:

Resolution of the Charleston Issue" [27], is summarized in tabular form in Table A-1.

Table A-1 JAF EPRI Site Seismic Hazard Curve Acceleration Frequency of Exceedance (/yr)

(g) 15% 50% 85% Mean 0.01 3.50 x 10 9.80 x 103 3.10 x 10 1.50 x 10-0.05 4.70 x 104 2.10 x 104 7.30 x 104 3.60 x 104 0.10 6.60 x 104 5.00 x 105 1.30 x 10 4 7.30 x 10-'

0.26 1.90 x 107 4.00 x 104 1.50 x 105 6.90 x 104 0.51 3.10 x 109 3.00 x 107 1.60 x 104 6.90 x 107 0.71 5.30 x 101 0 5.90 x 10 3.70 x 10-7 1.80 x 10'7 1.02 4.00 x 101 0 7.60 x 10 ' 7.00 x 104' 3.60 x 104 A2.4 Seismic Analysis Method of Analysis Although quantitative risk information is not directly available from the FitzPatrick SMA IPEEE analysis, a paper presented by Robert P. Kennedy [26] provides a method called the Simplified Hybrid Method for obtaining a seismic-induced CDF estimate based on results of an SMA analysis. This methodology requires only the plant HCLPF to estimate the seismic CDF. The approach entails the following steps:

1) Determine the HCLPF seismic capacity CHCLPF from the SMA analysis

2) Estimate the 10% conditional probability of failure capacity 00%

3) Determine hazard exceedance frequency H10%that corresponds to Ci0% from the hazard curve

4) Determine seismic accident type risk PF A-3

5)' Determine the seismic core damage frequency (SCDF) t Step 1A - Determine the HCLPF seismic capacity CLPF from the SMA analysis The FitzPatrick seismic analysis examined seven seismic accident types; station blackout, structural, loss-of-offsite power, small LOCA, medium LOCA, large LOCA and transient with loss-of-containment heat removal. The respective HCLPF values are:

CHCLPF..SBO 0.22g station blackout HCLPF [Section A2.3, Input #1]

CHCLPF..SSC 0.38g structural components HCLPF [Section A2.3, Input #1]

CHCLPF.LOSP 0.27g plant loss-of-offsite HCLPF [Section A2.3, Input #1]

CHCLPF-S2 0.27g small LOCA HCLPF [Section A2.3, Input #1]

CHCLPF4SI 0.22g medium LOCA HCLPF [Section A2.3, Input #1]

CHCLPF-A 0.31g large LOCA HCLPF [Section A2.3, Input #1]

CHCLPF-T. 0.30g loss of containment heat removal HCLPF [Section A2.3, Input #1]

However, the above values do not directly consider the effect of random failures on seismic risk because it uses the HCLPF Maximum/Minimum method to approximate the seismic accident type fragility. Per the OECD-NEA Workshop on Seismic Risk [26] methodology a HCLPF reduction factor is applied to each seismic accident type to account for non-seismic failures and human errors.

Based on a examination of the seismic IPEEE cutsets, a HCLPF reduction factor of 0.7 was selected. As a result, the revised seismic accident type HCLPF values are:

CHCLPF 480 0.22g I 0.7 0.31g (station blackout HCLP)

CHCLPF-SC 0.38g I 0.7 0.54g (structural components HCLPF)

CHCLPP-LOSP 0.279 I 0.7 0.39g (plant loss-of-offsite HCLPF)

CIICLP1-S2 0.27g I 0.7 0.39g (small LOCA HCLPF)

CHCLPF..SI 0.22g I 0.7 0.31g (medium LOCA HCLPF)

CHCLPF-A 0.31g I 0.7 0.44g (large LOCA HCLPF)

C~ICLPFAW 0.30g I 0.7 0.43g (loss of containment heat removal HCLPF)

SteD 2A - Estimate the 10% conditional Drobabilitv of failure cavacitv C0o Per the work presented in OECD-NEA Workshop on Seismic Risk [26], the 10% conditional probability of failure capacity is calculated as follows:

CIO% o=

• CHCuPw Where:

10% conditional probability of failure capacity CHCLPF = seismic accident type HCLPF capacity by the CDFM (Conservative Deterministic Failure Margin) method F0 = e(NEPIO%4NEPI%)p3 Where:

NEP10% 10% NEP standard normal variable is -1.282 [Section A2.3, Input #2]

NEP1% 1% NEP standard normal variable is -2.326 [Section A2.3, Input #3]

13 variable factor 0.3 [Section A2.3, Input #4]

Therefore, Fp = e(-1 282

. )- (-2.326)j 0.3 = 1.37 A-4

Therefore, C1O%sBo 1.37

• Clcepff-seQ 1.37

• 0.31 0.43g PGA C10%Ssc 1.37

• CWCLPF-SC 1.37 0.54 0.74g PGA 1.37

• CWCLPF-LOSP 1.37 0.39 0.53g PGA CIO%S2 1.37

• CHCLPF-S2 1.37 0.39 0.53g PGA C10%sl 1.37

• C:HCLPF -Sl 1.37

• 0.31 0.43g PGA CIO%A 1.37

• C:HCLPF-A 1.37 0.44 0.61g PGA CIO%Tw 1.37

• C:HCLPF-TW 1.37

• 0.43 0.59g PGA Step 3A - Determine hazard exceedance freauencv H0 % that corresponds to C10N,from the hazard curve The seismic hazard curve for the Fitzpatrick site, as presented in Table A-I is used to determine the hazard exceedance frequency H1 0%that corresponds to C10%. These are as follows:

Acceleration Mean Frequency of Seismic Accident TvDe (la) Exceedance (Iyr)

H1o~sBo 0.31 5.70 x 104 Hio%ssc 0.54 6.71 x 104 HioLosp 0.39 3.67 x 10J 0.39 3.67 x 107 H1o0 s 0.31 5.70 x IO-Hjo~sw 0.44 2.43 x 104 0.43 2.68 x I0 4 Step 4A - Determine seismic accident type risk PF Per the OECD-NEA Workshop on Seismic Risk [26], the seismic accident type risk is calculated as follows:

PF-WSO 0.5

• 0.5

• 5.70 x 104 2.5 x 10/y 0.5

• Hio~ssc 0.5

• 6.71 x 104 3.36 x 10ry PF-LOSP 0.5

• 0.5

• 3.67 x 104 1.84 x 10ry PF-S2 0.5

• 0.5

• 3.67 x 10' 1.84 x 104 /ry PF..Sl 0.5

• HioLos 0.5

• 5.70 x 104 2.85 x 104 y PF..A 0.5

• 0.5

• 2.43 x 104 1.21 x 10:/ri PF..TW 0.5

• 0.5

• 2.68 x 104 1.34 x 10ry Step SA - Determine the seismic core damage freauency (SCDF)

The step involves the summation of the individual seismic accident types frequencies.

SCDF = PF-SBO + PFSSC + PF.LOSP + PF-S1 + PF-S2 + PF-A + PF-1.W SCDF = 2.85x104 +3.36x10 4 +1.84x10 4 +1.84x104 +2.85x10 4 +1.21x10 4 +1.34x10 4 SCDF = 1.53E-05 A-5

This information is used in Section A5.0 of this appendix to provide insight into the impact of external hazard risk on the conclusions of this ILRT risk assessment.

A3.0 FitzPatrick IPEEE Fire Analysis A3.1 Fire Analysis Methodology Selection The Fire analysis performed for the FitzPatrick IPEEE submittal 120] use the EPRI Fire PRA methodology

[23] following the guidance of NUREG-1407 [21]. The fire PRA analysis entailed the identification of critical areas of vulnerability, the calculation of fire initiation frequencies, the identification of fire-induced initiating events and their impact on systems, the disabling of critical safety functions, and potential fire-induced containment failure. The core damage frequency (CDF) contribution due to internal fires was calculated as 3.42 x 104Iry 128].

The conclusions of the FitzPatrick IPEEE fire PRA are as follows:

1. The major contribution (61.8 percent) to the fire-induced CDF comes from fires in the cable spreading room, control room, and relay room.

2. The CDF resulting from fires in the cable spreading room may be reduced significantly if the heat detector placement is changed.

3. No significant vulnerabilities were found in an evaluation to resolve unresolved safety issue USI-A45 with respect to decay heat removal fire vulnerabilities.

4. No significant vulnerabilities to water spray, flooding, and CO2 effects on safe shutdown equipment were found.

5. Except for a specific seismic vulnerability, the CDF induced by hydrogen fires and explosions falls well below the screening criterion of 107/ year.

6. No additional containment vulnerabilities resulting from fire and random equipment failures were seen.

7. A review of risk issues raised in the Fire Risk Scoping Study [10] concluded that no vulnerabilities exist at FitzPatrick with respect to these issues.

A3.4 Fire Analysis Method of AnalysIs The FitzPatrick IPEEE submittal [22, 27] for the fire induced core damage scenarios and the associated frequency results were resiewed in support of this assessment. Based on review of the critical fire areas, the approximate breakdown of the FitzPatrick fire risk profile is as follows:

Fire PRA Accident Type Frequency (try)  % CDF Cable spreading room 1.13 x 10 0.329 Control room 5.44 x 10, 0.159 Relay Room 4.75 x 104 0.139 Reactor building east crescent 1.74 x l04 0.051 Others 1.13 x 10 5 0.331 Total 3.42 x 10 5 1.0 A-6

This information is used in Section A5.0 of this appendix to provide insight into the impact of external hazard risk on the conclusions of this ILRT risk assessment.

A4.0 Other External Hazards The FitzPatrick IPEEE submittal [20], in addition to the internal fires and seismic events, examined a number of other external hazards:

• High Winds and ITornadoes

• External Flooding

• Ice, Hazardous Chemical, Transportation and Nearby Facility Incidents No risks to the plant occasioned by high winds and tornadoes, external floods, ice, and hazardous chemical, transportation and nearby facility incidents were identified that might lead to core damage with a predicted frequency in excess of 10 6 /year. Therefore, these other external event hazards are not included in this appendix and are expected not to impact the conclusions of this ILRT risk assessment.

A5.0 Effect of External Events Hazard Risk on ILRT Risk Assessment A5.1 Effect of External Hazard Risk on ILRT Risk Assessment Assumptlons

1) Because both the seismic margin analysis and fire PRA did not report any seismic-induced accident progression releases, for the purpose of this report, the percent contribution for EPRI accident classes 7b and 7c, are based on the accident class frequencies presented in Table 2-5 (see page 58 of 80) are used. These values are as follows:

CLASS 7b % = 7.37 x 107 / 2.44 x Io0 = 0.3019 CLASS_7c_% = 5.82 x 10 1 2.44 x 10i = 0.0239 The remaining ERPI Classes are determined strictly by external events core damage frequencies.

2) Both the seismic margin analysis and the fire PRA are dominated by non-recoverable accident sequences that result in large early releases. Specifically, non-recoverable station blackout accident sequences (seismic margin analysis and fire PRA) and seismic-induced containment failure sequences. From the FitzPatrick IPE submittal [6] station blackout initiated accidents dominate the occurrence of a large early release. This is attributed to drywell failures with no drywell spray operation and either dry or flooded molten core-concrete interaction.

Per the NEI Guidance Document [4M, Enclosure 1, Discussion of Conservatisms in Quantitative Guidance for Delta LERF Impact," specific accident sequences that independently cause a LERF or could never cause a LERF, are to be removed from Class 3b LERF evaluation. Therefore, for the external events impact on the ILRT risk assessment, the evaluation of LERF is performed by multiplying the Class 3b probability by only that portion of core damage frequency that is impacted by Type A ILRT.

A5.2 Effect of External Events Hazard Risk on ILRT Risk Assessment Input

1) Based on the examination in Sections A2.0 through A4.0 the FitzPatrick external event initiated CDF is approximately 3.42 x 104Iry (internal fires) + 1.53 x 10 /ry (seismic) = 4.95 x 104 /ry.

A-7

2) Based on Section A5.2, Assumption#1, the following external event accident sequences are excluded from the Class 3b frequency calculation because they cannot result in a LERF release or independently result in LERF:

0 Seismic-induced station blackout sequences, 5.70 x 104Iry 0 Seismic-induced containment failures sequences, 6.71 x 10 Iry a Seismic-induced loss-of-containment heat removal scenarios, 2.68 x 101/ry a Fire-induced station blackout sequences, 2.15 x 10e/ry (Cable spreading room, Control room, and Relay Room)

A5.3 Effect of External Events Hazard Risk on ILRT Risk Assessment Method of Analysis The FitzPatrick IPEEE external events risk information presented in Sections A2, A3 and A4 is used to calculate, in accordance with the NEI Interim Guidance [3] for the following:

1) Evaluate the risk impact fbr the New Surveillance Intervals of Interest

2) Evaluate the external hazard risk impact in terms of LERF

3) Evaluate the external hazard change in conditional containment failure probability Evaluate the risk Impact for the New Surveillance Intervals of Interest.

This step calculates the percentage of the total dose rate attributable to EPRI accident Classes 3a and 3b (those accident classes affected by change in ILRT surveillance interval) and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

The change in population dose rate is calculated as outline in Step 7 (section 2.4.7, page 46 of 80) of this report. The results of this calculations when using the information contain in Section A5.2, Assumption #1 and Section A5.3, Input #1, is presented below as follows:

For 3-in-10 years (internal fires and seismic event),

EPRI Class Person-regn Freguency43v Permsonm'reBrn 1 2.91 x 10 3.66 x 10 1.07 x 10 2 6.79 x 105 4.95 x 1O' 3.36 x 10.2 3a 2.91 x IO' 5.52 x107 1.61 x 102 3b 1.02 x 105 5.52 x 10' 5.63 x 10-3 4 N/A 0.00 0.00 5 N/A 0.00 0.00 6 N/A 0.00 0.00 7a 6.11 x105 2.44 x IO-' 1.49 x 101 7b 3.96 x 105 1.49 x 10i 5 5.91 7c 5.21 x 105 1.18 x 10I 6.16 x 10 '

7d 3.06 x 105 1.34 x 10' 4.10 x 10 1 8 4.46 x 105 3.36 x 10 ' 1.50 ji'~! !ITotal !i 4.95 x i2.3366 x 101 A-8

For 1-in-10 years (internal fires and seismic event),

EPRI Class Person-rein Freauengay/ ,Person-remlmv 1 2.91 x 103 2.25 x 10 6.55 x 10 2 6.79 x 10' 4.95 x 104 3.36 x 102 3a 2.91 x 10 4 1.84 x 10' 5.36 x 102 3b 1.02 x 10' 1.84x 10 7 1.88 x 102 4 N/A 0.00 0.00 5 N/A 0.00 0.00 6 N/A 0.00 0.00 7a 6.11 x 105 2.44 x 1O-5 1.49 x 10' 7b 3.96 x 105 1.49x 105 5.91 7c 5.21 x 105 1.18 x l0 4 6.16 x 10' 7d 3.06 x 10' 1.34 x 104 4.10 x 101-8 .4.46

..q. ~¢ex x 105 3.36 x I04 1.50 V'll,i~;ETotal:E i -- ~ 4.95 lo ~2.3412 x 10 For 1-in-15 years (internal fires and seismic event),

EPRI Class Person-rer Frequency/Ry Person-remlVy 1 2.91 x 103 1.23 x 10 3.60 x 10 4.95 x 10 4 3.36 x 10 2 6.79 x IO' 3a 2.91 X 10 4 2.76 x 104 8.04 x 102 3b 1.02 x leS 2.76 x 1077 2.81 x 10.2 4 N/A 0.00 0.00 5 N/A 0.00 0.00 6 N/A 0.00 0.00 7a 6.11 x 105 2.44x10-5 1.49x10I 7b 3.96 x IO' 1.49 x 105 5.91 7c 5.21 x 105 1.18 x 104 6.16 x 10-1 7d 3.06 x IO' 1.34 x 104 4.10 x 107' 8 i:"Till 4.46 x IO' 3.36 x 10 4 1.50 Total 4 . 72.3446 k1o

[ X_,E!.i ..... g4,95 Based on the results summarized above and those presented in Table 2-13 (see page 64 of 80), for the current FitzPatrick 1-inlO years ILRT interval, the percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b is calculated as follows:

PER...CHGCOMBINEDI10 = [CLASS-3a-DOSECOMBINED-10 + CLASS-3b-DOSE CMBINED-10 J 100 TOT- DOSEcoMBINED-10 Where:

PER-CHGCOMBINED -10 = combined internal and external events percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given an 1-in-10 years ILRT interval A-9

CLASS_3aDOSE COMBINED-10 = combined internal and external events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval

= CLASS_3aDOSE INTERNAL-10 + CLASS_3aDOSE EXTERNAL-10 CLASS 3b-DOSE COMBINED-10 = combined internal and external events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval

= CLASS_3b_DOSE INTERNAL-10 + CLASS_3bDOSE EXTERNAL -10 CLASS 3a-DOSE INTERNAL-10 = internal events EPRI accident Class 3a dose rate given a 1-in-10 years ILRT interval = 6.40 x 103 /ry [Table 2-13]

CLASS_3bDOSE INTERNAL-10 = internal events EPRI accident Class 3b dose rate given a 1-in-10 years ILRT interval = 2.24 x 104 /ry [Table 2-13]

CLASS_3aDOSE EXTERNAL -10 = external events EPRI accident Class 3a dose rate given a 1-in-I 0 years ILRT interval = 5.36 x 10-2 person-rem/ry [See 1-in-1O years table above]

CLASS_3bDOSE EXTERNAL 10 = external events EPRI accident Class 3b dose rate given a 1-in-1 0 years ILRT interval = 1.88 x 10-2 person-rern/ry [See 1-in-10 years table above]

TOT- DOSECOMBINED-10 = Total combined internal and external events dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval

= TOT- DOSEINTERNAL10 + TOT- DOSEFXTERNAs10 TOT- DOSE INTERNAL-10 = Total internal events dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval = 0.720 (person-rem/ry) [Table 2-13]

TOT- DOSE EXTERNAL-40 = Total external events dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval = 23.412 (person-rem/ry)See I-in-10 years table above]

Therefore, PERCHG COMBINED-40 = [ (6.40 x107 + 5.36x 102) + (2.24x104 +1.88x10-2 )

• 100 0.720 + 23.412 j PERCHG COMBINED-0 = 0.34%

The percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b based on the proposed 1-in-15 years ILRT interval is calculated as follows:

PER-CHGCOMBINED-15 = [CLASS.3a.DOSE COMBINED-1S + CLASS 3b-DOSE COMBINED-15 I

• 100

.. 1~~~~~~~~~~

. TOT- DOSErr.*

..........-1 D itJlillblNtlJ A-10

Where:

PER-CHGCOMBINED.-15 = combined internal and external events percentage contribution to total dose rate from EPRI's accident Classes 3a and 3b given an 1-in-15 years ILRT interval CLASS 3a DOSE COMBINED-15 = combined internal and external events EPRI accident Class 3a dose rate given a 1-in-15 years ILRT interval

= CLASS_3aDOSE INTERNAL-15 + CLASS_3aDOSE EXTRNAL-15 CLASS_3bDOSE COMBINED-15 = combined internal and external events EPRI accident Class 3b dose rate given a 1-in-15 years ILRT interval

= CLASS_3bDOSE INTERNAL-15 + CLASS_3bDOSE EXTERNAL -15 CLASS_3aDOSE INTERNAL-1S = internal events EPRI accident Class 3a dose rate given a 1-in-15 years ILRT interval = 9.60 x 103 person-rem/ry [Table 2-13]

CLASS_3bDOSE INTERNAL-15 = internal events EPRI accident Class 3b dose rate given a 1-in-1 5 years ILRT interval = 3.36 x 103 person-rem/ry [Table 2-13]

CLASS_3aDOSE EXTERNAL -15 = external events EPRI accident Class 3a dose rate given a 1-in-I 5 years ILRT interval = 8.04 x 10-2 person-rem/ry [See 1-in-15 years table above]

CLASS_3bDOSE EXTERNAL 15 = external events EPRI accident Class 3b dose rate given a 1-in-I 5 years ILRT interval = 2.81 x 10-2 person-rem/ry [See 1-in-15 years table above]

TOT- DOSECOMBINED 15 = Total combined internal and external events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval

= TOT- DOSEINTERNA-15 + TOT- DOSEEXTERNAJ.15 TOT- DOSE INTERNAL-15 = Total internal events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval = 0.724 (person-rem/ry) [Table 2-13]

TOT- DOSE EXTERNAL-15 = Total external events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval = 23.446 (person-rem/ry)[See 1-in-10 years table above]

Therefore, PER_CHG COMBINED-15 = (9.60x104 +8.04x10-) + (3.36x1073 + 2.81x10-2)1

• 100 I 0.724 + 23.446 J I

PERCHG COMBINED-15 = 0.50%

A-1l

Based on the above results, the combined internal and external events changes from the 1-in-10 years to 1-in-15 years dose rate is as follows:

INCREASECOMBINEDI0-15 = [ TOT- DOSEcoMsMNED-1s - TOT- DOSECOMBINED-10 I

• 100 TOT- DOSECOMBINED-10 Where:

INCREASECOMBINEDIO.15 = combined internal and external events percent change from Sin-10 years ILRT interval to 1-in-15 years ILRT interval TOT- DOSECOMBINED-15 = Total combined internal and external events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval

= TOT- DOSEINTERNAL-15 + TOT- DOSEEXTERN^-15 TOT- DOSEcoMsBNED-1o = Total combined internal and external events dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval

= TOT- DOSEINTERNAL-10 + TOT- DOSEEXTERNPJ-10 TOT- DOSE INTERNAL-15 = Total internal events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval = 0.724 (person-rem/ry) [Table 2-13]

TOT- DOSE EXTERNAJ-15 = Total external events dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval = 23.446 (person-rem/ry)[See 1-in-10 years table above]

TOT- DOSE INTERNAL-10 = Total internal events dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval = 0.720 (person-rem/ry) [Table 2-13]

TOT- DOSE EXTERNAI-10 = Total external events dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval = 23.412 (person-rem/ry) [See 1-in-10 years table above]

Therefore, INCREASEcOMBINEDIO-15 [ (23.446 + 0.724) - (23.412 + 0.720) 1

• 100 (23.412 + 0.720)

INCREASEcoMBaNEDo1015 = 0.16%

The above increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.16%. This value can be considered to be a negligible increase in risk.

A-12

Evaluate the External Events Hazard Risk Impact In Terms of LERF This step, per the NEI Interim Guidance [3] calculates the change in the large early release frequency with extending the ILRT interval from 1-in-10 years to 1-4I5-years.

The combined internal and external events affect on the LERF risk measure due to the proposed ILRT interval extension is calculated as follows:

,ALERFCOMBINED1O..15 = CLASS-..3bcomBINED15 - CLASS...3b COMBINED10 Where:

ALERFcOMBINED1o.15 = the combined internal and external events change in LERF from 1in-10 years ILRT interval to 1-in-15 years ILRT interval CLASS...3bcoMBINED15 = the combined internal and external frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval 3 3

= CLASS_ biNTERNAL-15 + CLASS- bEXTERNAL-15 CLASS...3biNTERNAL..15 = internal events frequency of EPRI accident Class 3b given a 1in-15 years ILRT Interval = 3.29 x 104Iry [Table 2-121 CLASS...3bEXTERNALA15 = external events frequency of EPRI accident Class 3b given a 1-in-15 years ILRT Interval = 2.76 x 1077/ry [See 1-in-15 years table above]

CLASS-3bCOMBINED10 = the combined internal and external frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval 3 3

= CLASS biNTERNAL-10 + CLASS- bEXTERNAL-1o CLASS...3bINTERNAL..10 = internal events frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval = 2.20 x 10'/ry [Table 2-12]

CLASS...3bEXTERNAL-10 - external events frequency of EPRI accident Class 3b given a 1-in-10 years ILRT Interval = 1.84 x 10 -7/ry [See 1-in-10 years table above]

Therefore,

,&LERFcOMBINED10..15 = (3.29 x104' + 2.76 x 1 ') - (2.20 x 107 + 1.84 x107)

ALERFOOMBINEDIO..15 = 1.03 x10-/ry The risk acceptance criteria of Regulatory Guide 1.174 as previously discussed in Section 2.4.8, Step 8 of this report, is used here to assess the ILRT Interval extension. Regulatory Guide 1.174, 'An Approach for Using PRA in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis' [5], provides NRC recommendations for using risk information in support of applications requesting changes to the license basis of the plant.

The ALERFCOMBINEDIO.15 of 1.03 x 10'/ry from extending the FitzPatrick LRT frequency from 1-in-10 years to 1-in-15 years is slightly above the 10 7 /yr criterion of Region IlIl, Very Small Change in Risk (Figure 2-1),

of the acceptance guidelines in NRC Regulatory Guide 1.174 [5]. Therefore, increasing the ILRT interval A-13

at FitzPatrick from the currently allowed 1-in-10 years to 1-in-1 5 years is non-risk significant from a risk perspective.

Evaluate the External Events Hazard Chanae In Conditional Containment Failure Probability This step calculates the change in conditional containment failure probability (CCFP).

Similar to Step 9 (Section 2.4.9) of this report, the change in CCFP tracts the impact of the ILRT on both early (LERF) and late radionuclide releases. Therefore, CCFP consists of all those accident sequences resulting in a radionuclide release other that the intact containment state for EPRI accident Class 1, and small failures state for EPRI accident Class 3a. In additional, the CCFP is conditional given a severe core damage accident. The change in CCFP is calculated by the following equation:

CCFP= {1-((Class I frequency + Class 3a frequencyl/CDF)}*100, %

For the combined internal and external events 1-in-10 years ILRT interval:

CCFPCOMBINED.1 tl1 - ( CLASS1 COMBINED-40 + CLASS 3aCOMBINED- 1)1

• 100%

CDFCCOMBINED Where:

CCFPCOMBINED -10 = combined internal and external events conditional containment failure probability given 1in-10 years ILRT interval CLASS_1 COMBINED-40 = combined internal and external events frequency of EPRI accident Class I given a 1-in-10 years ILRT interval

= CLASS_11NTERNAL-10 + CLASS 1 EXTERNAL-10 CLASS...INTERNA~L-1 = internal events frequency of EPRI accident Class 1given a 1-in-10 years ILRT interval = 4.49 x 1077ry [Table 2-12]

CLASS..1lEXTERNAL-j10 = external events frequency of EPRI accident Class Igiven a 1-in-10 years ILRT interval = 2.25 x lO4 /ry [See 1-in-10 years table above]

CLASS 3 aINTERNAL..10 = internal events frequency of EPRI accident Class 3a given a 1in-10 years ILRT interval = 2.20 x 10 7/ry [See 1-in-10 years table above]

CLASS- 3 aEXTERNAL..I0 = external events frequency of EPRI accident Class 3a given a 1-in-10 years ILRT interval = 1.84 x lO4 ry /ry [See 1-in-10 years table above]

CDFCOM13INED = FitzPatrick combined internal events and external events CDF

= 2.44 x 106/ry [Section 5, input#2] + 4.95 x 101/ry [Section A5.3, input#1]

= 5.19 x 104 /ry A-14

Therefore, CCFPCOMBINED -lo = 1 - ( [ (4.49 x 10-7 + 2.25 x 10) + (1.84 x 10' + 2.20 x 10-7) ))

• 100%

5.19 x lo, CCFPCOMMINED-.10 - 90.8%

For the combined internal and external events 1-in-15 years ILRT interval:

CCFPCOMBINED..15 -1(F CLASS.1 COMBINED-I 5 + CLASS 38COMBINED415 1)1} 100%

CDFcoMBINEI Where:

CCFPCOMBINED -15 = combined internal and external events conditional containment failure probability given 1in-15 years ILRT interval CLASS 1 COMBINED-15 = combined internal and external events frequency of EPRI accident Class 1 given a 1-in-15 years ILRT interval

= CLASSIINTERNAL-15 + CLASS_1 EXTERNAL-15 CLASS-11NTERNAL-15 = internal events frequency of EPRI accident Class lgiven a 1-in-15 years ILRT interval = 3.29 x 10e/ry [Table 2-12]

CLASS-1EXTERNAL-15 = external events frequency of EPRI accident Class Igiven a 1-in-15 years ILRT interval = 1.23 x lOIry [See 1-in-15 years table above]

CLASS- 3aINTERNAL-15 = internal events frequency of EPRI accident Class 3a given a 1-in-15 years ILRT interval = 3.29 x lO7 1ry [See 1-in-15 years table above]

CLASS-3a EXTERNAL-1S5 = external events frequency of EPRI accident Class 3a given a 1-in-15 years ILRT interval = 2.76 x 104 /ry /ry [See 1-in-15 years table above]

CDFCOMBINED = FitzPatrick combined internal events and external events CDF

= 2.44 x 104 /ry [Section 5, input#2] + 4.95 x 10-4 /ry [Section A5.3, input#1]

= 5.19 x 10-1ry Therefore, CCFPCOMBINM-15 ={ - ( [ (3.29 x 1077 + 1.23 x 106) + (3.29 x 0-7 + 2.76 x 1046) ] }

• 100%

I.L ' ~~~~~~5.19 x1lo, A-15

CCFPCOMBINED-15 = 91.0%

Therefore, the change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-15 years is:

ACCFPcoMBINED14s5 CCFP COMBINED15 - CCFPCOMBINED1o ACCFP COMBINEDIO-15 91.0% - 90.8%

ACCFP COMBINEDIO-15 0.20%

This change in CCFP of less than 1% is insignificant from a risk perspective.

The effects of external hazard risk on ILRT risk are shown in Table A-1. The combined internal and external events effect on the ILRT risk is shown in Table A-3. This Table combines the results of Table 11, 12, and 13 with the results depicted in Table A-2.

A6.0 Conclusions This appendix discusses the risk-implication associated with external hazards in support of the FitzPatrick Integrated Leak Rate Testing (ILRT) interval extension risk assessment. The following conclusions are derived from this evaluation

1. The ALERFCOMBINED1IO15 of 1.03 x 10/ry from extending the FitzPatrick ILRT frequency from 1-in-10 years to 1-in-15 years is slightly above the 107 /yr criterion of Region IlIl, Very Small Change in Risk (Figure 2-1), of the acceptance guidelines in NRC Regulatory Guide 1.174 [5]. Therefore, increasing the ILRT interval at FitzPatrick from the currently allowed 1-in-10 years to 1-in-15 years is non-risk significant from a risk perspective.

2. The combined internal and external events increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.16% (0.038 person-rem/ry). This value can be considered to be a negligible increase in risk.

3. The change in the combined internal and external events conditional containment failure probability from 1-in-10 years to 1-in-15 years is 0.20%. A change in CCFP of less than 1% is insignificant from a risk perspective.

A-16

Table A-2 Effect of External Events Hazard Risk on FitzPatrick ILRT Risk Assessment Base Case Extend to Extend to 3 Years _ 10 Years;_ 15 Years EPRI CDF IPer-RemI Per-Rem CDF Per-Rem Per-tem CDF I Per-Rem I Per-Rem Class (Per Ry) (Per Ry) (Per Ry) I (Per Rv) (Per RY) I (Per Rv) 1 3.66 x 10 4 2.91 x 103 1.07 x 10-2 2.25 x 104 2.91 x 10 3 6.55 x 10-3 1.23 x 106 2.91 x i03 3.60 x 104 2 4.95x10- 6.79x105 3.36x10-2 4.95x10-8 6.79x105 3.36x102 4.95x10-8 6.79x105 3.36x10-2 3a 5.52 x 10 7 2.91 x 104 1.61 x 10-2 1.84 x 104 2.91 x 104 5.36 x 10-2 2.76 x 104 2.91 x 104 8.04 x 102 3b 5.52 x 10-5 1.02 x 105 5.63 x 10-3 1.84 x 10-' 1.02 x 105 1.88 x 10-2 2.76 x 10-7 1.02 x 105 2.81 x 10-2 4 0.00 N/A 0.00 0.00 N/A 0.00 0.00 N/A 0.00 5 0.00 N/A 0.00 0.00 N/A 0.00 0.00 N/A 0.00 6 0.00 N/A 0.00 0.00 N/A 0.00 0.00 N/A 0.00 7a 2.44 x 10-5 6.11 x 105 1.49 x 10' 2.44 x 10' 6.11 x 10 1.49 x 10' 2.44 x IO' 6.11 x 1 5 1.49 x 101 7b 1.49 x 10-5 3.96 x 105 5.91 1.49 x 10-5 3.96 x 105 5.91 1.49 x 10-5 3.96 x 105 5.91 5

7c 1.18 x 104 5.21 x 10 6.16 x 101' 5 1.18 x 104 5.21 x 10 6.16 x 101' 5 1.18 x 104 5.21 x 10 6.16 x 10.1' 7d 1.34 x 104 3.06 x 105 4.10 x 10-' 1.34 x 104 3.06 x 105 4.10 x 10-' 1.34 x 104 3.06 x 105 4.10 x 101 '

8 3.36 x 10 4 4.46 x 105 1.50 3.36 x 10 4.46 x 105 1.50 3.36 x 104 4.46 x 105 1.50 Total 4.95 x 110 2.337 x 101 4.95 x 104 2.341 x 10' 4.95 x 104 2.345 x 10' ILRT Dose Rate 2.17 x 10-2 7.24 x 1072 1.09 x 10.1 from 3a and 3b

% Of Total 0.0929% 0.3091% 0.4630%

Delta Dose Rate 3.32 x 10.2 from 3a and 3b (10 to 16yr)

LERF from 3b 5.52 x 104 1.84 x 10-7 2.76 x 107 Delta LERF 9.20 x 104 (10 to 1 yr)

CCFP

• 91.48% 91.74% 91.93%

Delta CCFP % 0.186% l (10 to I6yr) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

A-17

Table A-3 Effect of Internal and External Events Risk on FitzPatrick ILRT Risk Assessment Base Case Extend to Extend to 3 Years, 10 Years 15 Years EPRI CDF I Per-Rem I Per-Rem CDF l Per-Rem l Per-Rem CDF Per-Rem J Per-Rem Class j(eY) (PerRy) (PerRy) ( e (PerRy) (Per Ry) 1 4.28 x 104 2.91 x 10 3 1.25 x 10-2 2.69 x 104 2.91 x 10 7.85 x 10-3 1.56 x 104 2.91 x 103 4.55 x 10-3 2 5.19x104 6.79x105 3.53x 10-2 5.19x104 6.79x10 5 3.53x10- 2 5.19x104 6.79x10 5 3.53x102 30 6.18 x 107 2.91 x 104 1.80 x 10-2 2.06 x 104 2.91 x 10i 6.00 x 10.2 3.09 x 104 2.91 x 10 4 9.00 X 10-2 3b 6.18x10 4 1.02x105 6.30x10-3 2.06x10 7 1.02x105 2.10x10-2 3.09x10 7 1.02x10 5 3.15x10-2 4 0.00 N/A 0.00 0.00 N/A 0.00 0.00 NIA 0.00 5 0.00 N/A 0.00 0.00 N/A 0.00 0.00 N/A 0.00 6 0.00 N/A 0.00 0.00 N/A 0.00 0.00 N/A 0.00 70 2.45 x ic0 6.11 x 105 1.50 x 10' 5 2.45 x 10 6.11 x 105 1.50 x 10' 2.45 x lo, 6.11 x 105 1.50 x 101 7b 1.57x I04 3.96x 105 6.20 1.57x 10-5 3.96x 105 6.20 1.57x 105 3.96x 05 6.20 7c 1.24 x 104 5.21 x I 6.46 x 10-' 1.24 x 104 5.21 x 10 6.46 x 10-1 1.24 x I04 5.21 x0 5 6.46 x10-7d 1.82x104 3.06x10 5 5.58x10- 1 1.82x104 3.06x10 5 5.58x10-' 1.82x104 3.06x105 5.58x10-'

8 3.65x104 4.46x105 1.63 3.65x10 I .46x105 1.63 3.65x10 4 4646x10 5 1.63 Total 5.19 x 104 2.408x10 1 5.19x105 2.413x10' 5.19x105 2.417x10' ILRT Dose Rate 2.43 x 10f2 8.10 x 10.2 1.22 x 10.'

from 3a and 3b

% Of Total 0.1009% 0.3357% 0.5027%

Delta Dose Rate 3.80 x 10.2 from 3a and 3b (10 to 15 yr) 7 7 LERF from 3b 6.18 x 104 2.06 x 10- 3.09 x 1i0 Delta LERF 1.03 x 1 o0 (10 to 15 yr)

CCFP 90.57% 90.85% 91.05%

Delta CCFP % 0.198%

(1 0 to 1 5 vr) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

A- 18

Appendix B Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval

Table of Contents Page No.

B1.0 Introduction B-2 B2.0 Method of Analysis B-2 B3.0 Assumptions B-3 B4.0 Input B-3 B5.0 Steel Shell Corrosion Analysis B-4 B6.0 Steel Shell Corrosion Sensitivity B-16 B6. Conclusions B-17 Tables Page No.

Table B Flaw Failure Rate as a Function of lime B-1 8 Table B Flaw Failure Rate as a Function of Test Interval B-18 Table B FitzPatrick Containment Failure Probability Given Containment Liner Flaw B-19 Table B FitzPatrick Containment Liner Corrosion Base Case B-20 Table B Impact of Containment Steel Liner Corrosion on FitzPatrick ILRT Intervals B-21 Table B Containment Steel Liner Corrosion Sensitivity Cases B-22 Figures Page No.

Figure B FitzPatrick Containment Failure Probability Given Containment Liner Flaw B-19 B-1

B13.0 Introduction Inspections of reinforced and steel containments at some facilities (e.g., North Anna, Brunswick D.C. Cook, and Oyster Creek) have indicated degradation from the inaccessible side of the steel shell and liner of primary containments. The major inaccessible areas of the Mark I containment are the vertical portion of the drywell shell and part of the shell located between the drywell floor and the basemat. As a result of these inaccessible areas, a potential increase in risk due to liner leakage, caused by age-related degradation mechanisms may occur when extending the current 1-in-10 years to 1-in-15 years Type A Integrated Leak Rate Testing (ILRT) interval.

Therefore, this appendix evaluates the likelihood and risk-implication associated with containment liner corrosion going undetected in visual examinations during the proposed extension of the ILRT interval.

B2.0 Method of Analysis The analysis utilizes the referenced Calvert Cliffs Nuclear Power Plant assessment 124] to estimate the risk impact from containment liner corrosion during an extension of the ILRT interval.

Consistent with the Calvert Cliffs analysis, the following issues are addressed:

• Differences between the containment basemat and the drywell and torus liner

• The historical drywell/torus steel shell flaw likelihood due to concealed corrosion

• The impact of aging

• The corrosion leakage dependency on containment pressure

• The likelihood that visual inspections will be effective at detecting a flaw The method determines the total likelihood of non-detected containment leakage given a change in the likelihood given that a flaw exists (i.e., increase in flaw likelihood due to the ILRT extension), that the flaw is not detected and that flaw results in a breach.

Consistent with Calvert Cliffs analysis 124], the following six steps are performed:

1) Determine the historical liner flaw likelihood.

2) Determine aged adjusted liner flaw likelihood.

3) Determine the increase in flaw likelihood between 3, 10 and 15 years.

4) Determine the likelihood of containment breach given liner flaw.

5) Determine the visual inspection detection failure.

6) Determine the likelihood of non-detected containment leakage.

B-2

In additions to these steps, the following three additional steps are added to evaluated risk-implication of containment liner corrosion:

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

8) Evaluate the risk impact in terms of LERF.

9) Evaluate the change in conditional containment failure probability.

B3.0 Assumptions

1) Consistent with the Calvert Cliffs methodology [24], a half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures.

2) Consistent with the Calvert Cliffs methodology [24], the leakage potential via the drywell floor (due to crack formation) is considered less likely than other sections of the containment structure.

3) Consistent with the Calvert Cliffs methodology [24], the likelihood of the containment atmosphere reaching the outside atmosphere given a liner flaw exists was estimated as a function of the pressure inside the containment.

4) Consistent with the Calvert Cliffs methodology [24], the containment liner flaw likelihood doubles every five years. This is based solely on judgment and is included in this analysis to address the increase likelihood of corrosion as the containment liner ages.

5) Consistent with the Calvert Cliffs methodology [24], the probability of a concurrent containment breach given a flaw in the containment liner is depicted as an exponential function.

6) Consistent with the Calvert Cliffs methodology [24], a 0.05 (5%) visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 0.10 (10%) is used 15

7) Consistent with the Calvert Cliffs methodology 124], 1.0 (100%) visual inspection detection failure likelihood given the flaw is located in an inaccessible area of either the drywell or torus.

8) Consistent with the Calvert Cliffs methodology [24], all non-detectable containment failures are considered to result in large early releases.

B4.0 Input

1) The containment liner failure rate is based on two industry events:

1. On September 22, 1999, North Anna Unit 2 experienced through-wall corrosion of the metal liner.

The corrosion appeared to have been initiated from a piece of lumber imbedded in the concrete behind the liner plate.

2. On April 27, 1999, inspection at Brunswick 2 discovered two through-wall holes and pitting in the drywell shell. The through-wall condition was believed to have originated from the coated (visible) side.

I5 Note: to date, all liner corrosion events have been detected through visual inspect ion.

B-3

2) The number of steel-lined containments is 70 [24].

3) The exposure time in detecting a containment flaw is 5.5 years. This is consistent with the Calvert Cliffs methodology 124] and reflects the time period since IOCFR 50.55a starting requiring visual inspection. This is deemed conservative, since the exposure time period is bounding as no additional failures have been identified in the nuclear industry since March 2002 and no failures were identified prior to September 1996 (the date when 10CFR 50.55a was implemented).

4) Consistent with the Calvert Cliffs methodology [241, leakage through the drywell floor is 10 times less likely than through other sections of the containment structure.

5) The probability of a concurrent containment breach given a flaw in the containment liner is depicted as an exponential function. This curve is used to interpolate the containment failure probability at the pressure at which the ILRT is to be performed for the accessible and inaccessible areas of containment. Consistent with the Calvert Cliffs methodology, the lower bound limit was assigned a failure probability of 0.1% at a pressure of 20 psia and the upper bound was assigned a failure probability of 100% at the ultimate containment failure pressure of 155psia psia [6].

B5.0 Steel Shell Corrosion Analysis SteI IB - Determine the Historical Liner Flaw Likelihood.

This step calculates historical liner flaw likelihood consistent wit the Calvert Cliffs mythology [24]. This value, for FitzPatrick's consists of the accessible potion of the drywell and torus, the inaccessible portion of the drywell and submergence area of the torus, and the inaccessible area of the drywell floor.

The accessible portion of the drywell and torus liner flaw likelihood is determined as follows:

AHLFDT = NFAIL, I (NPLANTS

• TEXPO)

The inaccessible portion of the drywell and submergence area of the torus liner flaw likelihood is determined as follows:

IAHLFDT = NFAIL, / (NPLANTS

• TEXPO)

The inaccessible area of the drywell floor IAHLFDF = NFAILI I (NPLANTS

• TEXPO)

Where:

AHLFDT = accessible portion of the drywell and torus liner flaw IAHLFDT = inaccessible portion of the drywell and submergence area of the torus liner flaw likelihood IAHLFDF = inaccessible area of the drywell floor liner flaw NFAIL, = number of industry events due to liner corrosion = 2 [Section B4.0, Input #1]

NFAILR, = number of industry events due basemat corrosion = 0.5 [Section B3.0, Input *1]

NPLANTS = number of steel-lined containments = 70 [Section B4.0, Input #2]

TEXPO = time exposure since issuing of 10CFR5O.55a = 5.5 years [Section B4.0, Input #3]

B-4

Therefore, AHLFDT = 2 (70

• 6.5) = 5.19 x lO 4 yr IAHLFDT = 2 I (70

• 5.5) = 5.19 xlIO4 Iyr IAHLFDF = 0.5 I (70

• 5.5) = 1.30OxlO 4I-yr The above results are documented in Table B-4.

Step 2B - Determine Aged Adjusted Liner Flaw Likelihood.

Per the Calvert Cliffs methodology [24], the aged adjustment liner flaw likelihood is calculated for a 15-year interval given that the failure rate doubles every 5 years (Section B3.0, assumption #4) or increases 14.9 % per year. In addition, the average for the 5th to 1 0th year was set to the historical failure calculated in Step 1B.

The results, based on an iterative process that satisfies the above conditions are presented in Table B-1.

Step 3B - Determine the Increase In flaw likelihood between 3. 10 and 15 years! 6 .

This step calculates the increase in flaw likelihood at 3-in-10 years interval (or 1in-3 years), 1-in-10 years interval, and 1-in-15 years interval, per the Calvert Cliffs methodology [241. The results of Step 2B are use to generate these values as follows:

Accessible portion of the drywell and torus, ADTFLAW3-1 0 I Z ADTFRTEM i=1,3 ADTFLAW 1..Io = X ADTFRATEB i=1 ,1O ADTFLAW 1.1 s = E ADTFRATEII i=1 ,15 Inaccessible portion of the drywell and submergence area of the torus, IDTFLAW- 10 = Z IDTFRATEII i=1,3 IDTFLAW- 1. 0 = z IDTFRA i=1,10 IDTFLAWI 15 = E IDTFRTMI i=1,15 la (Note: the Calvert ClifMs analysis presents the delta between 3 and 15 years of 8.7% to utilize inthe estimation of the delta-LERF value. For this analysis, however, the values are calculated based on the 3-i-10 years, 1-in-10years, and -in-15 years intervals consistent with the evaluation inthis calculation, and then the delta-LERF values are determined from there.

B-5

Inaccessible area of the drywell floor DFFLAW-io = Z DFFRAEa i=1,3 DFFLAWiio = v DFFRATE.

i=1,10 DFFLAWI-Is = I DFFRATE.

i=1,15 Where:

ADTFLAW3-io increase in flaw likelihood at 3-in-10 years test interval given accessible portion of the drywell and torus ADTFLAW1..1o = increase in flaw likelihood at 1-in-10 years test interval given accessible portion of the drywell and torus ADTFLAWi 1 5.I = increase in flaw likelihood at 1-in-15 years test interval given accessible portion of the drywell and torus IDTFLAW3-10 = increase in flaw likelihood at 3-in-10 years test interval given inaccessible portion of the drywell and submergence area of the torus IDTF LAW1..1o = increase in flaw likelihood at 1-in-10 years test interval given inaccessible portion of the drywell and submergence area of the torus IDTFLAW1..15 = increase in flaw likelihood at 1-in-15 years test interval given inaccessible portion of the drywell and submergence area of the torus DFFLAW3-10 = increase in flaw likelihood at 3-in-1 0 years test interval given inaccessible area of the drywell floor DFF LAW1 -10 = increase in flaw likelihood at 1-in-10 years test interval given inaccessible area of the drywell floor DFFLAWi..15 = increase in flaw likelihood at 1-in-15 years test interval given inaccessible area of the drywell floor ADTFPRATEI = aged adjusted liner flaw likelihood, given accessible portion of the drywell and torus (Table B-1)

IDTFPATEi = aged adjusted liner flaw likelihood, given inaccessible portion of the drywell and submergence area of the torus (Table B-1)

DFFRATEI = aged adjusted liner flaw likelihood, given inaccessible area of the drywell floor (Table B-1)

B-6

Therefore, ADTFLAW310 = 0.71%, ADTFLAWI-10 = 4.14%, ADTFLAW 1..15 = 9.68%

IDTFLAW3 10 = 0.71%, IDTFLAW1 -1 0 = 4.14%, IDTFLAW1 -15 = 9.68%

DFFLAW310 = 0.18%, DFFLAW 1. 10 = 1.04%, DFFLAWi. 15 = 2.42%

The above results are documented in Table B-2.

Step 4B - Determine the Likelihood of Containment Breach Given Liner Flaw.

The likelihood of a breach in containment given a liner flaw is based on the Calvert Cliffs methodology (241 with a FitzPatrick specific value for the upper-end pressure failure (100% likelihood) taken from Section 4.5 of the IPE []. A containment pressure of 155 psia corresponds with the 100% probability of failure. The lower-end pressure failure (0.1% likelihood) is set at 20 psia, consistent with Calvert Cliffs

[24]. Per the Calvert Cliffs methodology [24], the containment failure probability (FP) versus containment pressure (P) is assumed to be an equation of the form:

FP(P) = b*enmP Where:

FP (P) = containment failure probability given containment liner breach m = slope of the containment failure probability b = intercept of the containment failure probability p = containment pressure, psia The two anchor points of 0.1% at 20 psia and 100% at 155 psia provide sufficient information to solve for the slope m, and the intercept b, as follows:

Slope m, m = LN (FP(100%) - LN (0.1%) I (Upper Pressure -Lower Pressure) m = LN (1.0) - LN (0.001) / (155-20) m = 5.12 x 10-2 Intercept b, b = FP (100%) I em'P b = I 1 e6.12 x 10-r155 b = 3.56x 104 B-7

The FitzPatrick March 7, 1995 ILRT used a test pressure of 46.5 psig (or 61.2 psia) [25]. Based on this pressure the likelihood of containment breach in the liner is:

FP (61.2 psia) = 3.56 x 10 4 eS.1 2X 10-2'61.2 FP (61.2 psia) = 0.0082 or 0.82%

For the Drywell floor, the failure probability is set to one-tenth of the failure probability for Drywell walls, or 0.082%. (See Section B3.0, Assumption #4 and Section B4.0, Input #2).

Based on the above equation, containment liner breach and drywell floor intermediate values for FP are calculated and presented in Table B-3 and Figure B-1.

Step 6B - Determine the visual Insnection detection failure.

This step examines the visual inspection detection failure likelihood for FitzPatrick. The three areas of interest are the accessible portion of the drywell and torus, the inaccessible portion of the drywell and submergence area of the torus, and the inaccessible portion of the drywell floor.

The visual inspection detection failure likelihood for the accessible area of the drywell (100 percent internal and 75 percent external) [29] and torus (100 percent external and 100 percent of the area above the water line) [29] is set to 10%, consistent with the Calvert Cliffs analysis [241. This represents a 5%

(0.05) failure to identify a visual flaw and 5% (0.05) likelihood that the flaw is not visible.

The inaccessible portion of the drywell and submergence area of the torus is assigned a 100% (1.0) visual detection failure likelihood. This is bounding, as the submerged area of the Torus may be examined.

Because the liner under the Drywell floor cannot be visually inspected, a visual detection failure likelihood of 100 % (1.0) is assigned, consistent with the Calvert Cliffs method.

The above results are documented in Table B-4.

Step 6B - Determine the likelihood of non-detected containment leakage Per the Calvert Cliffs methodology [24], the likelihood of a non-detected containment leakage is calculated by multiplying the results of Steps 3B, 4B, and 5B. This yields the following:

Accessible portion of the drywell and torus, ADTLEAK-io = ADTFLAW3 1o

• ADTFPILRT

• ADTVISUAL ADTLEAKI..0 = ADTFLAW .o 1

• ADTFPILRT

• ADTVISUAL ADTLEAK-I. 5 = ADTFLAW1. 15 I ADTFPILRT

• ADTVISUAL Where:

ADTLEAK3 10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and accessible portion of the drywell and torus B-8

ADTLEAKI- 10 = likelihood of non-detected containment leakage, given 1-in-1 0 years test interval and accessible portion of the drywell and torus ADTLEAKI- 15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and accessible portion of the drywell and torus ADTFLAWs1 0 = increase in flaw likelihood at 3-in-10 years test interval given accessible portion of the drywell and torus = 0.71% (0.0071) [Table B-2]

ADTFLAW 1 ,0 = increase in flaw likelihood at 1-in-10 years test interval given accessible portion of the drywell and torus = 4.14% (0.0414) [Table B-2]

ADTFLAW-..1 = increase in flaw likelihood at 1-in-15 years test interval given accessible portion of the drywell and torus = 9.68% (0.0968) [Table B-2]

ADTFPILRT = likelihood of containment breach at ILRT test pressure (61.2 psia) given liner flaw and accessible portion of the drywell and torus = 0.0082 (0.82%) [Step 41]

ADTVISUAL = visual inspection detection failure accessible portion of the drywell and torus

= 0.1 (10%) [Step 58]

Therefore, ADTLEAK3I0 = 0.0071

• 0.0082

• 0.1 = 5.822 x 10' (0.0005822%)

ADTLEAK1 -1 o = 0.0414

• 0.0082

• 0.1 = 3.395 x 10-5 (0.0033948%)

ADTLEAK1.1 5 = 0.0968

• 0.0082

• 0.1 = 7.938 x 10-5 (0.0079376%)

Inaccessible portion of the drywell and submergence area of the torus, IDTLEAK3 10 = IDTFLAW3 10

• ADTFPILRT IDTVISUAL IDTLEAK1 -10 = IDTFLAW1 -1 0

• ADTFPILRT IDTVISUAL IDTLEAK1. 15 = IDTFLAWi 15

• ADTFPILRT

• IDTVISUAL Where:

IDTLEAK3-10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTLEAKI-1 o = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTLEAK1.-15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTFLAW3-io = increase in flaw likelihood at 3-in-10 years test interval given inaccessible portion of the drywell and submergence area of the torus = 0.71% (0.0071) [Table B-2]

B-9

IDTFLAW1 .so = increase in flaw likelihood at 1-in-10 years test interval given inaccessible portion of the drywell and submergence area of the torus = 4.14% (0.0414) [Table B-2]

IDTFLAW1 -15 = increase in flaw likelihood at 1-in-15 years test interval given inaccessible portion of the drywell and submergence area of the torus = 9.68% (0.0968) [Table B-2]

ADTFPILRT = likelihood of containment breach at ILRT test pressure (61.2 psia) given liner flaw and inaccessible portion of the drywell and submergence area of the torus

= 0.0082 (0.82%) [Step 4B]

IDTVISUAL = visual inspection detection failure inaccessible portion of the drywell and submergence area of the torus = 1.0 (100%) [Step 5Bj Therefore, IDTLEAK31 0 = 0.0071

• 0.0082

• 1.0 = 5.822 x 105 (0.005822%)

IDTLEAKi.i 0 = 0.0414

• 0.0082

• 1.0 = 3.395 x 10 (0.033948%)

IDTLEAKm1 5 = 0.0968

• 0.0082

• 1.0 = 7.938 x 10' (0.079376%)

Inaccessible portion of the drywell floor, DFLEAK3 10 = DFTFLAW3 10

• DFTFPILRT

• DFTVISUAL DFTLEAK-.1 0 = DFTFLAW.,l 0

• DFTFPILRT

• DFTVISUAL DFTLEAKl.Is = DFTFLAW1 -15

• DFTFPILRT

• DFTVISUAL Where:

DFLEAK3-1 o = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the drywell floor DFLEAK1..1 o = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the drywell floor DFLEAK 1.15S = likelihood of non-detected containment leakage, given 14in-15 years test interval and inaccessible portion of the drywell floor DFFLAW3-1 0 = increase in flaw likelihood at 3-in-I0 years test interval given inaccessible portion of the drywell floor = 0.18% (0.0018) [Table B-2]

DFFLAW1..Io = increase in flaw likelihood at 1-in- 10 years test interval given inaccessible portion of the drywell floor = 1.04% (0.0104) [Table B-2]

DFFLAWI..15 = increase in flaw likelihood at 1-in-15 years test interval given inaccessible portion of the drywell floor = 2.42% (0.0242) [Table B-2J DFTFPILRT = likelihood of containment breach at ILRT test pressure (61.2 psia) given liner flaw and inaccessible portion of the drywell floor = 0.00082 (0.082%) [Step 4B]

B-10

DFVISUAL = visual inspection detection failure inaccessible portion of the diIywell floor

= 1.0 (100%) [Step SB]

Therefore, DFTLEAK3.I0 = 0.0018

• 0.00082

• 1.0 = 1.476x le4 (0.0001476%)

DFTLEAK 1.. 0 = 0.0104

• 0.00082 1.0 = 8.528 x 104 (0.000853%)

DFTLEAK,- 15 = 0.0242

• 0.00082

• 1.0 = 1.984 x 105 (0.001984%)

Total Likelihood of Non-Detected Containment Leakage due to Corrosion is, TOTAL3. = ADTLEAKr31 o + IDTLEAK 3. l0 + DFTLEAK3. io TOTAL,1 0 = ADTLEAKI- 10 + IDTLEAK1 -10 + DFTLEAK1 -10 TOTAL1,-If = ADTLEAKI. 15 + IDTLEAK 1 -15 + DFTLEAK 1 15 Where:

TOTAL 10 = total likelihood of non-detected containment leakage due to corrosion, given 34in-10 years test interval TOTAL,., 0 = total likelihood of non-detected containment leakage due to corrosion, given 1in-10 years test TOTAL,-I5 = total likelihood of non-detected containment leakage due to corrosion, given 1-in-I 5 years test interval ADTLEAK3 10 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and accessible portion of the drywell and torus ADTLEAK1 -10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and accessible portion of the drywell and torus ADTLEAK 1 _15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and accessible portion of the drywell and torus IDTLEAK3 -10 = likelihood of non-detected containment leakage, given 3-in-I 0 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTLEAKI. 10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the drywell and submergence area of the torus IDTLEAK 1. 15 = likelihood of non-detected containment leakage, given 1-in-I5 years test interval and inaccessible portion of the drywell and submergence area of the torus DFLEAK 3., 0 = likelihood of non-detected containment leakage, given 3-in-10 years test interval and inaccessible portion of the drywell floor DFLEAK,- 10 = likelihood of non-detected containment leakage, given 1-in-10 years test interval and inaccessible portion of the drywell floor B-11

DFLEAK1 .15 = likelihood of non-detected containment leakage, given 1-in-15 years test interval and inaccessible portion of the drywell floor Therefore, TOTAL31 o = 0.00058% + 0.00582% + 0.00016% = 0.00655%

TOTAL 1 10o = 0.00340% + 0.03390% + 0.00085% = 0.0385%

TOTAL1, 1 s = 0.00794% + 0.07938% + 0.00198% = 0.08975%

The above results are documented in Table B-4.

Step 7B - Evaluate the Risk Impact In Terms of Population Dose Rate and Percentile Change for the Interval Cases.

This step calculates the change in population dose rate for EPRI accident Class 3b (all non-detectable containment failures are considered to result in large early releases), the change in percentage of the total dose rate attributable to liner corrosion and the change in this result dose rate from the base dose rate attributable to changes in ILRT surveillance interval.

The change in population dose rate is calculated as outline in Step 7 (Section 2.4.7, page 46 of 80), of this report. The results of this calculations, is presented below as follows:

For 3-in-10 years, EPRI Class PersonWrepi Freguency Rv Person-rerrf 1 2.91 x 10 6.18 x 10 1.80 x lo, 2 6.79 x 10 5 2.44 x 10e Corrosion Addition 1.66 x 107 3a 2.91 x 10 4 6.59 x 10 1.92 x 10-3b 1.02 x 10 5 6.75 x 10 9 1.60 x 10 1 6.88 x 10-4 4 N/A 0.0 0.0 S N/A 0.0 0.0 6 N/A 0.0 0.0 7a 6.11 x 105 1.74 x 107 1.06 x 10-7b 3.96 x 105 7.37 x 10 7 2.91 x 10-7c 5.21 x 105 5.82 x 10 3.04 x 10-2 7d 3.06 x I05 4.83 x 10-7 1.48 x 10-8 4.46 x 105 2.97 x 10i 7 1.32 x 10-1 To ital 2.44 x'I0 1 -IL- : 0.7147T - --

ILRT Dose Rate from 3a and 3b = 1.92 x I0' + 6.88 x 10-4 = 2.61 x 10-3 person-rern/ry

%Of Total = 100 *[ 1.92_x 10-3 + 6.88 x 104 ] 1 0.7147 = 0.3650%

LERF from 3b = 6.75 x 10-9/ry CCFP%LINER310 = 1 - [6.18 x 10 + 6.59 x 10 l 1 2.44 x 10 4 = 72.0%

B-12

For 1-in-10 years, EPRI Class Personre~M Freguencvly Person-remJy 1 2.91 x 10 4.49 x 10 1.31 x 10'3 2 6.79 x I05 2.44 x le-9 CorrosicmnA dddition 1.66 x 103 3a 2.91 x l04 2.20 x le-7 6.40 x 107' 3b 1.02 x 105 2.29 x 10 9.29 x 11)1o 2.34x 10'3 4 NJA 0.0 0.0 5 NIA 0.0 0.0 6 NIA 0.0 0.0 7a 6.11 x 105 1.74 x 10' 1.06 x 107' 7b 3.96 x 105 7.37 x IO' 2.91 x 10-1 7c 5.21 x 105 5.82 x IO' 3.04 x 10-2 7d 3.06 x I05 4.83 x 1iO 7 1.48 x 101-8 4.46 x I05 2.97 x IO'7 1.32 x 10' T

-o-t-a-l Y! 2A4 x ,10 .7203 ILRT Dose Rate from 3a and 3b = 6.40 x10 4' + 2.34 x 10 43 = 8.74 x 10'3 person-rem/ry

%OfTotal = 100 * [6.4x10 3 + 2.34 x 10 4 / 0.7203 = 1.213%

LERF from 3b = 2.29 x 104 /ry CCFPLINER-110 = I - [4.49 x 10 + 2.20 x10-7 1 I 2.44 x104 = 72.6%

For 1-in-15 years, EPRI Class Person-rem Frequen PersonrernifL 1 2.91 x 10 3.26 x 10 9.52 x 10-3 2 6.79 x 10 5 2.44 x I0 9 Corrosion Addition 1.66x 103 3a 2.91 x 104 3.29 x 10' 9.60 x 10-3 3b 1.02x10 6 3.51 x 104 2.17 x 109 3.58 x 10-3 4 NIA 0.0 0.0 5 NIA 0.0 0.0 6 NIA 0.0 0.0 7a 6.11 x 105 1.74 x I0-7 1.06 x 10.1 7b 3.96 x 105 7.37 x 10-7 2.91 x 10-1 7c 5.21 x 105 5.82 x 104 3.04 x 10-2 7d 3.06 x I05 4.83 x 10 7 1.48 x 10-8 4.46 x 105 2.97 x10' 1.32 x 10-1 Total t : ;.44 x w;::-IO 7,; : ;:E;L Q 70J244.

ILRT Dose Rate from 3a and 3b = 9.60 x 10J + 3.58 x 104 = 1.32 x 1O02 person-rem/ry

%OfTotal = 100 * [9.60x10 + 3.58x1043 1 0.7244 = 1.819%

LERF from 3b = 3.51 x 104 Iry CCFP%LINERI.15 = I - [3.26 x 10- + 3.29 x 10 7 1 1 2.44 x 10 = 73.2%

B-13

Based on the above results, the changes from the 1in-10 years to 1-in-15 years dose rate is as follows:

INCREASEUNERIO-1 5 = [ TOT- DOSERATE..NERg5 - TOT- DOSERATE-LINERgO]

• 100 TOT- DOSERATE-UNERIG Where:

INCREASELINER1IO15 = percent change from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval TOT- DOSE RATE-UNER15 = Total dose rate for all EPRI's Classes given a 1-in-15 years ILRT interval

= 0.7244 (person-rem/ry) [See 1-in-15 years table above]

TOT- DOSE RATE-UNERIS = Total dose rate for all EPRI's Classes given a 1-in-10 years ILRT interval

= 0.7203 (person-rem/ry) [See 1-in-10years table above]

Therefore, INCREASEUNERlO-1s = [ 0.7244 - 0.7203 ] 100 = 0.57%

0.7203 The above increase in risk on the total integrated plant risk for those accident sequences influenced by Type A testing, given the change from a 1-in-10 years test interval to a 1-in-15 years test interval, is found to be 0.57%. This value can be considered to be a negligible increase in risk.

Step 8B - Evaluate the risk Impact In terms of LERF This step calculates the change in the large early release frequency with extending the ILRT intervals from 1-in-10 years to 1-mI5-years given the inclusion of a postulated liner corrosion flaw failure.

The affect on the LERF risk measure due to liner corrosion flaw is calculated as follows:

ALERFLNER1o..15 = CLASS_3bFREQUENC YLNER15 - CLASS_3b.FREQUENC Y LNERIO Where:

ALERFLNER1O-15 = the change in LERF from 1-in-10 years ILRT interval to 1-in-15 years ILRT interval CLASS-3b4FREQUENC YLNER15 = frequency of EPRI accident Class 3b given a 1-in-15 years ILRT interval = 3.51 xi04 try [Step 7B]

CLASS_3bFREQUENC Y LNERIO = frequency of EPRI accident Class 3b given a 1-in-10 years ILRT interval = 2.29 x 105/ry [Step 7BJ Therefore, ALERFLNERIO..15 = 3.51 XI0 - 2.29 x1 07 ALERFLNER10..15 = 1.22 x10-8/ry B-14

Based on this result, the inclusion of corrosion effects in the ILRT assessment would not change the previous conclusions of this report (See Section 2.4). That is, the change in LERF from extending the interval to 15 years from the current 10 years requirement is estimated to be about 1.22 x 1I0/ry. This value is below the NRC Regulatory Guide 1.174 [5] of 10 7Ir. Therefore, because Regulatory Guide 1.174 (5] defines very small changes in LERF as below 10 /yr, increasing the ILRT interval at FitzPatrick from the currently allowed 1-in-10 years to 1-in-15 years and taking into consideration the likelihood of a containment liner flaw due to corrosion is non-risk significant from a risk perspective.

Similarly, the change in LERF from the original 3-in-10-year interval is calculated as follows:

ALERFLNER3-15 = CLASS_3bFREQUENC YLNER15 CLASS..3bJFREQUENC Y IJ4ER3 Where:

ALERFLNER3-15 = the change in LERF from 3-in-10 years ILRT interval to 1-in-15 years ILRT interval CLASS_3b_FREQUENC YLNER15 = frequency of EPRI accident Class 3b given a 1-in-15 years ILRT interval = 3.51 x I01 Iry [Step 7BJ CLASS_3b-FREQUENC Y LNER3 = frequency of EPRI accident Class 3b given a 1-in-10 years ILRT interval = 6.75 x 10 9 /ry [Step 7B]

Therefore, ALERFLNER3-15 = 3.51 XIO-' - 6.75 x IO-ALERFLNER3-15 = 2.84 x I0-8ry Step 98 - Evaluate the change In conditional containment failure probability This step calculates the change in conditional containment failure probability (CCFP). Similar to Step 9 (Section 2.4.9) of this report, the change in CCFP tracts the impact of the ILRT on both early (LERF) and late radionuclide releases. Therefore, CCFP consists of all those accident sequences resulting in a radionuclide release other that the intact containment state for EPRI accident Class 1, and small failures state for EPRI accident Class 3a. In additional, the CCFP is conditional given a severe core damage accident. Therefore, the change in the conditional containment failure probability from 1-in-10 years to 1-in-15 years is:

ACCFPUNER1O0.15 - CCFPLINERII15 - CCFPLINER140 Where:

ACCFPLINER10-15 = the change in conditional containment failure probability from 1-in-10 years to 1-in-I5 years given non-detected containment leakage CCFPUNER10 = conditional containment failure probability given 1-in-10 years ILRT interval and potential non-detected containment leakage [Step 7B]

CCFPUNER15 = conditional containment failure probability given 1-in-15 years ILRT interval and potential non-detected containment leakage [Step 7B]

Therefore, 3-15

ACCFP UNER10-15 73.2% - 72.6%

ACCFP UNERIO-15 - 0.6%

This change in ACCFP uNER1o-15 of less than 1% is insignificant from a risk perspective.

The results of Steps 7B, 88, and 9B of the updated ILRT assessment including the potential impact from non-detected containment leakage scenarios assuming that 100% of the leakages result in EPRI Class 3b are show in Table B-5.

B6.0 Steel Shell Corrosion Sensitivity Additional sensitivity cases were also developed to gain an understanding of the sensitivity of this analysis to the various key parameters. The sensitivity cases are as follows:

• Sensitivity Case 1 - Flaw rate doubles every 2 years

• Sensitivity Case 2 - Flaw rate doubles every 10 years

• Sensitivity Case 3 - 5% Visual inspection failures

• Sensitivity Case 4 - 15% Visual inspection failures

• Sensitivity Case 5 - Containment breach base point 10 times lower

• Sensitivity Case 6 - Containment breach base point 10 times higher

• Sensitivity Case 7 - Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF (Lower bound)

• Sensitivity Case 8 - Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF (upper bound)

The above sensitivities cases used the calculational methodology presented in Steps 2B to 9B. These steps were developed in an EMEL spreadsheet. They are reproduced in Attachment B.

These results are summarized in Table B-6.

B-16

B7.0 Conclusions This appendix provides a sensitivity evaluation of considering potential containment liner corrosion impacts within the structure of the ILRT interval extension risk assessment. The evaluation yields the following conclusions:

1. The impact of including age-adjusted corrosion effects in the ILRT assessment has minimal impact on plant risk and is therefore acceptable.

2. The change in LERF, taking into consideration the likelihood of a containment liner flaw due to age-adjusted corrosion is non-risk significant from a risk perspective. Specifically, extending the interval to 15 years from the current 10 years requirement is estimated to be about 1.22 x 10'Iry. This is below the Regulatory Guide 1.174 [51 acceptance criteria threshold of I 7 lyr.

3. The age-adjusted corrosion impact in dose increase is estimated to be 4.1 x 103 person-rem/ry or 0.57% from the baseline ILRT 10 year's interval.

4. The age-adjusted corrosion impact on the conditional containment failure probability increase is estimated to be 0.6%.

5. A series of parametric sensitivity studies regarding potential age related corrosion effects on the containment steel liner also demonstrated minimal impact on plant risk.

B17

Table B-I Flaw Failure Rate as a Function of Time Accessible Area Inaccessible Area Drywell Floor Year DrJwell and Torus Drywell and Torus Failure Rate Success Rate Failure Rate Success Rate Failure Rate Success Rate 0 1.79x10- 9.98x10-1 1.79x10- 9.98x10-' 4.46x104 1.00 1 2.05 x 10- 9.98 x 10- 2.05 x 1V 9.98 x 10-1 5.13 x 104 9.99 x10" 2 2.36 x 10- . 9.98 x 10-1 2.36 x 10- 9.98 x 10-1 5.89 x 10 9.99x 10-T 3 2.71 x le 9.97 x 0-' 2.71 x 10 9.97 x 10-' 6.77 x 104 9.99 x I-'

4 3.11x10 9.97x10- 3.11x109 9.97x10-1 7.78x10 4 9.99x10-'

5 3.57 x 104 9.96 x 10-' 3.57 x 10 9.96 x 10-' 8.94 x 104 9.99 x 10-1 6 4.11x10-3 9.96x10-' 4.11x10-3 9.96x10' 1.03x104 9.99x10-'

7 4.72 x 104 9.95 x 101 4.72 x 104 9.95 x 101 1.18 x10 9.99 x 10-8 5.42x10-3 9.95x107- 5.42x104 9.95x10- 1.36x104 9.99x10-9 6.23x10- 9.94x1071 6.23x10- 9.94x10- 1.56x1073 9.98x10-'

10 7.16x10 9.93x10-' 7.16x10- 9.93x10- 1.79x10- 9.98x10-'

11 8.23x104 9.92x10-' 8.23x 10- 9.92x10-1 2.06x 10-3 9.98x10-'

12 9.45 x 10 9.91 x 10-1 9.45 x 10-3 9.91 x 10-1 2.36 x 104 9.98 x 10-'

13 1.09 x 10 9.89 x 10-' 1.09 x 10 9.89 x10-' 2.71 x 104 9.97 x 10' 14 1.25x10- 9.88x10-' 1.25x10I- 9.88x 10-' 3.12x10-3 9.97x10' 15 1.43x10- 9.86x10-' 1.43x10-2 9.86x10-1 3.58x10- 9.96x10-Table B-2 Flaw Failure Rate as a Function of Test Interval Accessible Area Inaccessible Area Drywell Floor Years Dnrwell and Torus Drvwell and Torus

(

Failure Rate Success Rate Failure Rate Success Rate Failure Rate Success Rate 3-in-I0 0.71% 9.93 x 10-' 0.71% 9.93x10- 0.18% l 9.98x10-1-in-10 4.14% 9.59 x 10-, 4.14% 1 9.59 x10-, 1.04% 9.900x10-1-n-I 15 9.68% 9.03 x 10-' 9.68% 9.03x10' 2.42% 9.76x10O-'

B-18

Table B-3 FitzPatrick Containment Failure Probability Given Containment Liner Flaw Pressure (psla) 1 Containment Liner 1 Drywell Floor Failure Probability Failure Probability 0 0.0000 0.0000 10 0.0006 0.0001 15 0.0008 0.0001 20 0.0010 0.0001 30 0.0017 0.0002 40 0.0028 0.0003 50 0.0046 0.0005 60 0.0077 0.0008 70 0.0128 0.0013 80 0.0214 0.0021 90 0.0357 0.0036 100 0.0596 0.0060 110 0.0994 0.0099 120 0.1659 0.0166 130 0.2768 0.0277 140 0.4618 0.0462 150 0.7706 0.0771 155 0.9954 0.0995 Figure O FitzPatrick Containment Failure Probability Given Containment Liner Flaw 1.00 0.90 0.80 0.70 c 0.60

.0 0.50

= 0.40 IL.0 0.30 0.20 0.10 0.00 0 20 40 60 80 100 120 140 160 Containment Pressure, psia B-19

Table B-4 FitzPatrick Containment Liner Corrosion Base Case Accessible Area Inaccessible Area Drywell Floor Description Drywell and Torus Drewell and Torus I Historical Steel Shell Flaw Likelihood 5.19x 104 5.19 x 10-3 1.30x 10-3 2 Age Adjusted Steel Shell Flaw Year Failure Year Failure Year Failure Rate Likelihood Rate Rate 1 2.05 x 10o 1 2.05 x 104 1 4.46 x 104 5-15 5.19 x 10 4 5-15 5.19 x 10 3 5-15 1.30 x 10-3 15 1.43 x 102 15 1.43 x 10-2 15 3.58 x 10-3 Increase in Flaw Likelihood at 0.71% (3-to-i0 years) 0.71% (3-to-10 years) 0.18% (3-to-10 years)

3. 10, and 15 years 4.14% (14-t-10 years) 4.14% (1-to-3 years) 1.04% (1-to-3 years) 9.68% (14o-15 vears) 9.68% (1-toI vears) 2A2% (1-to-3 tears) 4 Likelihood of Breach in Pressure Likelihood Pressure Likelihood Pressure Likelihood Containment Given Steel (psia) of Breach (psia) of Breach (psia) of Breach Shell Flaw 20 0.001 20 0.001 20 0.000 61.2 (ILRT) 0.008 61.2 (ILRT) 0.008 61.2 (ILRT) 0.0008 100 0.060 100 0.060 100 0.006 120 0.166 120 0.166 120 0.017 155 0.995 155 0.995 155 0.100 5 Visual Inspection Detection 0.1 (10%) 1.0 (100%) 1.0 (100%)

Failure Likelihood 6 Likelihood of Non-Detected 0.00058% (3-to-1 0 years) 0.00582% (3-to-10 years) 0.00015% (3-to-1 0 years)

Containment Leakage (Steps 0.00340% (1-to-10 years) 0.03390% (1-to-10 years) 0.00085% (1-to10 years) 3*4*5) 0.00794% (1-to-15 years) 0.07938% (1-to-I5 years) 0.00198% (14o-15 vears)

Total Likelihood of Non-Detected 0.00655% (3-to-10 years)

Containment Leakage 0.03815% (1-to-10 years) 0.08975% (1-to-15 years)

B-20

Table B-5 Impact of Containment Steel Liner Corrosion on FitzPatrick ILRT Intervals Base Case Extend to Extend to 3 Years 10 Years I1 _ 5Years EPRI CDF I Per-Rem I Per-Rem CDF l Per-Rem Per-Rem CDF l Per-Rem I Per-Rem Class (PerRy) ll(Per I (PerRy) (PerRy) l (PerRy) 1 6.18 x I0V 2.91 x 103 1.80 x 10-3 4.49 x 10V 2.91 x 103 1.31 x 10-3 3.26 x 10-7 2.91 x 103 9.52 x 104 9 5 2 2.44 x 10-9 6.79 x 105 1.66 x 10-3 2.44 x 10- 6.79 x 10 1.66 x 10-3 2.44 x 1P 6.79 x 105 1.66 x 10'3 3a 6.59 x 104 2.91 x 104 1.92 x 10'3 2.20 x 1 0 -7 2.91 x 10 4 6.40 x 10-3 3.29 x 10-7 2.91 x 104 9.60 x 10-3 3b 6.75 x 10-9 1.02 x 105 6.88 x 104 2.29 x 104 1.02 x 105 2.34 x 10-3 3.51 x 104 1.02 x 105 3.58 x 10-3 4 0.0 N/A 0.0 0.0 NIA 0.0 0.0 N/A 0.0 5 0.0 N/A 0.0 0.0 NIA 0.0 0.0 N/A 0.0 6 0.0 N/A 0.0 0.0 NIA 0.0 0.0 N/A 0.0 7b 1.74 x l0e 6.11 x 105 1.06 x 10-' 1.74 x 10-7 6.11 x 105 1.06 x 1071 1.74 x l0e 6.11 x 105 1.06 x 10-1 7b 7.37 x 10e 3.96 x 105 2.91 x 10-' 7.37 x 1o-7 3.96 x 10 5 2.91 x 10-' 7.37 x 10-7 3.96 x 10 5 2.91 x l 1o' 7c 5.82 x 10 4 5.21 x 10 5 3.04 x 10.2 5.82 x 104 5.21 x 105 3.04 x 102 5.82 x 104 5.21 x 105 3.04 x 10-2 7d 4.83 x 10-7 3.06 x 105 1.48 x 10' 4.83 x 10-7 3.06 x 105 1.48 x 10-' 4.83 x 10-7 3.06 x 10 5 1.48 x 10- 1 8 2.97 x 10-7 4.46 x 10 5 1.32 x 10-' 2.97 x 10-7 4.46 x 105 1.32 x101 7 2.97X10 4.46 x 10 5 1.32 x 101' Total 2.44 x 10 7.1 x 10' 2.44 x 104 7.20 x 10' 2.44 x 104 7.24 x 10-'

ILRT Dose Rate 2.61 x 10'3 8.74 x 10-3 1.32 x 102 from 3a and 3b (+1.63 x 10Or) (+9.48 x 10GT (+2.21E-04)

% Of Total 0.365% 1.2128% 1.8199%

(+0.0023%)r (+0.0130%) (+0.0300%r Delta Dose Rate 4.09 x 10 4 from 3a and 3b (+0.0123°h)

(10to 15yr)

LERF from 3b 6.75 x 10-9 2.29 x 104 3.51 x 103

(+1.60 x 10f')' (+9.29 x 101'y (+2.17 x 10')

Delta LERF 1.22 x 104 (10 to 15 yr) (+1.24 x 10-)

CCFP 71.99% 72.65% 73.15%

(+0.0065%j) (+0.0381 %) (+0.0889%)

Delta CCFP % 0.60%

(10 to 15 vrl (0.0508%)

• Denotes increase from original values presented in Steps 7,8, and 9 (Section 2.4.7,2.4.8 and 2.4.9) of this report.

B-21

Table B-6 Containment Steel Liner Corrosion Sensitivity Cases Visual LERF LERF LERF Total LERF Drywell! Inspection Likelihood Increase Increase Increase Increase Age Torus & Non- Flaw is LERF From From From From ILRT (Step 2) Breach Visual (EPRI Class Corrosion Corrosion Corrosion Extension (Step 4) Flaws 3b) (3-in-10 (1-in-10 (1 to 15 (10 to 15 (Step 5) years) years) years) years)

Base Case Base Case Base Case Base Case Base Case Base Case Base Case Base Case Doubles .8171%liner 10% 100% 1.60 x 10'10 9.29 x 1010 2.17 x 10-9 1.22 x 106 every 5 yrs 0.0817%floor Doubles Base Base Base 4.57 x 10-" 7.75 x 10-'0 4.50 x io-9 1.47 x 1o every 2 yrs Doubles Base Base Base 2.37 x 10-'1 3.25 x 10-10 4.20 x 10-10 1.11 x 104 every 10 yrs Base Base 5% Base 1.53 x 10-' 6.20 x 10-" 1.45 x 10.'1 1.11 x 104 Base Base 15% Base 1.67 x 10-10 1.45 x 10.'° 3.38 x 1010 1.12 x 104 Base .166%liner'7 Base Base 3.25 x 10." 1.89 x 10.10 4.42 x 101'0 1.12 x 104

.017%floorI3 Base .07%liner1 Base Base 7.95 x 10-'° 4.63 x 109 1.08 x 10- 1.72 x I04 0.411 %floor" I I ___ -- I Lower Bound Doubles Base 5% 10% I 4.62 x 10-12 2.00 x 101 " 3.66 x 10-" .lOE-08 every IO yrs I I I I I I Upper Bound Doubles 2 Base 15% 100% 2.38 x 10- 0 1 4.03 x 10 9 2.34 x 10-8 3.04E08 every 2yrs I I II I I 7

1 Base point 10Otimes lowevrthan bas caseof 0.00O at20psia.

'aBasetpoit I imshigherthanbbase caseofO.O1I at20psia.

B-22

Attachment A FitzPatrick IPE Level 2 Model Results to Support One-Time Extension

Table of Contents Page No.

A1.0 Introduction A2 A2.0 Mean Internal Core Damage Frequency Contributions By Plant Damage States Results A2 A3.0 Summary of FitzPatrick IPE LEVEL 2 Release Categories A2 A4.0 Summary of FitzPatrick IPE LEVEL 2 Containment Failures A5 Figure Al - JAF ILRT Evaluation PDS -1 Base Results A9 Figure A2 - JAF ILRT Evaluation PDS -2 Base Results A10 Figure A3 - JAF ILRT Evaluation PDS-3 Base Results All Figure A4 - JAF ILRT Evaluation PDS-4 Base Results A12 Figure A5 - JAF ILRT Evaluation PDS-5 Base Results A13 Figure A6 - JAF ILRT Evaluation PDS -6 Base Results A14 Figure A7 - JAF ILRT Evaluation PDS -7 Base Results A15 FigureA8 - JAF ILRT Evaluation PDS-8 Base Results A16 Figure A9 - JAF ILRT Evaluation PDS -9 Base Results A17 Figure A10 - JAF ILRT Evaluation PDS-l0 Base Results A18 Figure All - JAF ILRT Evaluation PDS-11 Base Results A19 Al

A1.0 Introduction This attachment presents the results of the FitzPatrick Level 2 IPE model to support the one-time exemption to the ten-year frequency of the performance-based Type A containment leakage-testing program for the Fitzpatrick plant. Specifically, the values of Tables 2-1, 2-2 and 2-3 are presented.

A2.0 Mean Internal Core Damage Frequency Contributions by Plant Damage States Results Table 2-1 of this report presents the mean internal core damage frequency contributions by plant damage states. The table uses the plant damage states frequencies as reported in Table 4.4.2.2 (Section 4.4) of the FitzPatrick IPE [6J. The mean internal plant damage states frequencies are computed by taking the individual plant damage states frequencies and dividing by the point estimate core damage frequency value of 2.17 x lO6 Iry to obtain the percent contribution to core damage frequency for each plant damage state (column label % of CDF). The mean internal core damage frequency is then obtained by multiplying this result by the mean core damage frequency of 2.44 x 1Oe/ry (FitzPatrick IPE, Revision 1)

[6].

These results are use in Table 2-2 of this report (the mean internal core damage frequency contributions by plant damage states)

The EXCEL spreadsheet data is as follows:

Mean Internal Core Damage Frequency Contributions by Plant Damage States Results From Table 4.4.2.2 Rev.1 Frequency (Iry) %CDF-PDS PDS Frequency TOTAL PDS1 FREQUENCY = 1.28E-08 0.0059 1.44E-08 TOTAL PDS2 FREQUENCY = 1.69E-07 0.0778 1.90E-07 TOTAL PDS3 FREQUENCY = 1.11E-06 0.5110 1.25E-06 TOTAL PDS4 FREQUENCY = 1.90E-07 0.0875 2.13E-07 TOTAL PDS5 FREQUENCY = 4.08E-08 0.0188 4.58E-08 TOTAL PDS6 FREQUENCY = 7.71 E-08 0.0355 8.66E-08 TOTAL PDS7 FREQUENCY = 5.68E-09 0.0026 6.38E-09 TOTAL PDS8 FREQUENCY = 3.56E-08 0.0164 4.OOE-08 TOTAL PDS9 FREQUENCY = 1.44E-08 0.0066 1.62E-08 TOTAL PDS10 FREQUENCY = 2.50E-07 0.1151 2.81 E-07 TOTAL PDS1 I FREQUENCY = 2.67E-07 0.1229 3.OOE-07

=

k" I. ,,Total= :1 '2.117EW:',Zll:i-7 1.00E+00 0 - 2.44E-06 Aft A3.0 Summary of FitzPatrick IPE LEVEL 2 Release Categories Table 2-2 of this report presents a summary of FitzPatrick's IPE Level 2 release categories based on the mean internal core damage frequency of 2.44 x 104 /ry 16]. The results presented in Table 2-2 are computed in a two-step process. Step one multiplies the individual plant damage states frequencies by the release magnitudes split fractions generated by the Event Progress Analysis Code (EVNTRE) 130].

Step two, takes these results and sums the value for each of the nine-release category found in the Fitzpatrick IPE analysis.

A2

Shown below are the EXCEL spreadsheets for each plant damage state release bins, followed by the summary of FitzPatrick's IPE Level 2 release categories based on the mean internal core damage frequency.

Individual Plant Damage States Release Categories Release Category -Split Fractions IPS1 PSD-2IPDS3 PDS-41 PDS-5 I PDS-6 Late Low 0.000 0.000 0.000 0.040 0.013 0.006 Late Medium Low 0.313 0.313 0.279 0.085 0.293 0.180 L Medium High 0.000 0.000 0.000 0.000 0.000 0.000 Late High 0.013 0.013 0.010 0.095 0.024 0.242 Early Low 0.000 0.000 0.000 0.000 0.000 0.000 Early Medium Low 0.000 0.000 0.000 0.021 0.000 0.000 Early Medium High 0.245 0.245 0.279 0.145 0.239 0.107 Early High 0.000 0.000 0.000 0.522 0.047 0.374 No Containment Failure 0.429 0.429 0.432 0.094 0.384 0.091

~Total 10001.0 ~1000 1.00100 I 1.000 Individual Plant Damage States Release Categories Release Catepory - Split Fractions IPDS-71 PSD-8 I PDS-9 .PDS-1 01 PDS-1 I Late Low 0.007 0.000 0.000 0.000 0.000 Late Medium Low 0.316 0.313 0.000 0.000 0.000 L Medium High 0.000 0.000 0.000 0.000 0.000 Late High 0.016 0.013 0.000 0.000 0.000 Early Low 0.000 0.000 0.000 0.000 0.000 Early Medium Low 0.000 0.000 0.885 0.887 0.788 Early Medium High 0.245 0.245 0.115 0.113 0.118 Earlv High 0.000 0.000 0.000 0.000 0.094 No Containment Failure 0.416 0.429 0.000 0.000 0.000 Total 1.0001 1.000 1.000 '1.000 -1.000 A3

Individual Plant Damage States Release Categories Release Category - Frequencylry PDS-1 I PSD-2 PDS-3 PDS-4 Late Low O.OOE+OO O.OOE+OO O.OOE+OO 8.62E-09 Late Medium Low 4.50E-09 5.94E-08 3.48E-07 1.81 E-08 L Medium High O.OOE+OO O.OOE+OO O.OOE+OO 8.37E-12 Late High 1.88E-10 2.49E-09 1.26E-08 2.03E-08 Early Low O.OOE+OO O.OOE+OO O.OOE+OO O.OOE+OO Early Medium Low O.OOE+OO O.OOE+OO O.OOE+OO 4.46E-09 Early Medium High 3.52E-09 4.65E-08 3.48E-07 3.09E-08 Early High O.OOE+OO O.OOE+OO O.OOE+OO 1.11E-07 No Containment Failure 6.17E-09 8.14E-08 5.39E-07 2.OOE-08 Total= 41.4-08 I 90E-07 1.5E- 21E-07 Individual Plant Damage States Release Categories Release Category - Frequencylry PDS-6 PSD-6 I PDS-7 PDS-8 Late Low 5.91E-10 5.01E-10 4.68E-11 O.OOE+00 Late Medium Low 1.34E-08 1.56E-08 2.02E-09 1.25E-08 L Medium High 9.44E-12 7.59E-12 1.67E-12 O.OOE+OO Late High 1.08E-09 2.1 OE-08 1.03E-10 5.20E-10 Early Low O.OOE+00 O.OOE+O0 O.OOE+0O O.OOE+O0 Early Medium Low 2.07E-11 O.OOE+0O O.OOE+OO O.OOE+O0 Early Medium High 1.10E-08 9.27E-09 1.56E-09 9.80E-09 Early High 2.16E-09 3.24E-08 O.OOE+00 O.OOE+00 No Containment Failure 1.76E-08 7.88E-09 2.65E-09 1.72E-08 Total- 4.58E-08 8.66E-081 6.38E-09 4.OOE-08 Individual Plant Damage States Release Categories Release Category - Frequencylry PDS-9 I PSD-1 0 I PDS41* TOTALi-Late Low 0.OOE+00 0.OOE+0O O.OOE+O0 9.76E-09 Late Medium Low O.OOE+O0 O.OOE+O0 O.OOE+00 4.73E-07 L Medium High O.OOE+0O O.OOE+O0 O.OOE+00 2.71 E-11 Late High O.OOE+00 O.OOE+OO O.OOE+OO 6.82E-08 Early Low O.OOE+00 O.OOE+0O O.OOE+O0 O.OOE+OO Early Medium Low 1.43E-08 2.49E-07 2.36E-07 5.04E-07 Early Medium High 1.86E-09 3.17E-08 3.54E-08 6.29E-07 Early High O.OOE+0O O.OOE+00 2.82E-08 1.74E-07 No Containment Failure O.OOE+0O O.OOE+00 O.OOE+0O 6.91 E-07'1

-Total 1.62E-08 2.81E-07 3.OE-07 2.44E-O6 9 These are the values used in Table 2.

A4

A4.0 Summary of FitzPatrick IPE LEVEL 2 Containment Failures Table 2-3 of this report presents a summary of FitzPatrick's IPE Level 2 containment failures bins frequencies based on the mean internal core damage frequency of 2.44 x 104Sry [6J. The results presented in Table 2-3 are computed as follows:

No Containment Failure Frequency = Frequency of No Containment Failure Release Category No Containment Failure = 6.91 E-07/ry Early Drywell Failure Frequency = Frequency of Early High Release Category Early Drywell Failure = 1.74E-07/ry Early Wetwell Failure Frequency = (Frequency of Early Medium Low Release Category

+ Frequency of Early Medium High Release Category)

- (Frequency of ATWS Early Medium Low Release Category

+ Frequency of ATWS Early Medium High Release Category)

Early Wetwell Failure Frequency = (5.04E-07 + 5.29E-07)

- (1.43E-08 + 2.49E-07 + 1.86E-09 + 3.17E-08)

Early Wetwell Failure Frequency = 7.37E-07Iry Late Drywell Failure Frequency = Frequency of Late High Release Category Late Drywell Failure = 5.82E-08/ry Late Wetwell Failure Frequency = Frequency of Late Low Release Category

+ Frequency of Late Medium Low Release Category

+ Frequency of Late Medium High Release Category Late Wetwell Failure Frequency = 9.76E-09 + 4.73E-07 + 2.71 E-1 1 Late Wetwell Failure Frequency = 4.83E-07 Bypass Failure Frequency = Frequency of ATWS Early Medium Low Release Category A5

+ Frequency of ATWS Early Medium High Release Category Bypass Failure Frequency = 1.43E-08 + 2.49E-07 + 1.86E-09 + 3.17E-08 Bypass Failure Frequency = 2.97E-071ry A5.0 FitzPatrick Collapsed Accident Progression Bins Frequencies Table 2-8 of this report presents the associated collapsed accident progression bins frequencies for FitzPatrick.

The results presented in Table 2-8 are determined by running the EVNTRE 130]. This code is used to compute the ten collapsed accident progression bins for use in mapping the FitzPatrick person-rem bins to those found in the NUREG/CR-4551 Peach Bottom, Unit 2 analysis [9 and 12]. The EXCEL spreadsheet results are summarized below and depicted in Figures Al to A1.

Legend:

CD = core damage VB = vessel breach CF = containment failure DW = drywell WW = torus RPV = reactor pressure vessel Collapsed Accident Progression Bins -Split FactionsjPDS-1 PSD-2 I PDS-3 I PDS-4 1 PDS-6 I PDS-6 VB, Early CF, WW, RPV pressure >200 psig at VB 0.000 0.000 0.0000 0.016 0.000 0.000 V Early CF, WW, RPV pressure <200 psig at VB 0.000 0.000 0.0000 0.000 0.000 0.000 VB, Early CF, DW, RPV pressure >200 psig at VB 0.000 0.000 0.0830 0.616 0.066 0.000 Early CF, DW, RPV pressure <200 psig at VB 0.245 0.245 0.1961 0.053 0.221 0.481

• Late CF, WW 0.000 0.000 0.0000 0.001 0.003 0.000

• Late CF, DW 0.067 0.067 0.0534 0.103 0.071 0.252 VENT 0.259 0.259 0.2352 0.117 0.255 0.175

• No CF 0.139 0.139 0.1950 0.028 0.115 0.027 No VB, No CF 0.290 0.290 0.2373 0.066 0.269 0.064 No CD, No CF 0.000 0.000 0.0000 0.000 0.000 0.000 Total 1.000 1.00 I 1.000 I1.000 1.000 1.000 11 A6

Collapsed Accident Progression Bins -Split FactlonsIPDS-71 PSD-8 IPDS-9 PDS-1 01 PDS-11 B, Early CF,WW, RPV pressure >200 psig at VB 0.000 0.000 0.000 0.081 0.389 B. Early CF. WW. RPV pressure <200 psib at VB 0.000 0.000 0.082 0.000 0.003 Earl CF. DW, RPV pressure >200 Psi at VB 0.000 0.000 0.000 0.494 0.548 Early CF, DW. RPV pressure <200 psig at VB 0.245 0.245 0.461 0.000 0.032 VB_Late CF, VW 0.001 0.000 0.000 0.000 0.000 VB, Late CF, DW 0.072 0.067 0.000 0.000 0.000 VB. VENT 0.267 0.259 0.457 0.425 0.027 YB No CF 0.125 0.139 0.000 0.000 0.000 No VB, No CF 0.290 0.290 0.000 0.000 0.000 No CD No CF 0.000 0.000 0.000 0.000 0.000

-Total 1.000 1.000 -1.000 1.000 - 1.000 Collapsed Accident Progression Bins - Frequency/rv PDS-1 I PSD-2 I PDS-3 PDS-4 VB, Early CF,WW, RPV pressure >200 psig at VB O.OOE+00 O.OOE+00 O.OOE+00 3.38E-09 lB, Early CF,VWV, RPV pressure <200 psig at VB O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 lB,Early CF. DW, RPV pressure >200 psig at VB O.OOE+00 O.OOE+00 1.03E-07 1.31 E-07 IB,Early CF, DW, RPV pressure <200 psig at VB 3.53E-09 4.65E-08 2.44E-07 1.14E-08 VB, Late CF. WW O.OOE+00 O.OOE+00 O.OOE+00 1.47E-10 VB. Late CF, DW 9.62E-10 1.27E-08 6.66E-08 2.21E-08 VB, VENT 3.72E-09 4.92E-08 2.93E-07 2.50E-08 B,No CF 2.OOE-09 2.64E-08 2.43E-07 5.92E-09 No VB,No CF 4.17E-09 5.50E-08 2.96E-07 1.41E-08 NO CD, No CF O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 Total- 1.44E-08 1.90E-07 1.25E-06 2.13E-07 Collapsed Accident Progression Bins - Frequencylryl PDS-6 PSD-6 I PDS-7 1 PDS-8 VB, Early CF. WW, RPV pressure >200 psig at VB O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 VB, Early CF, WW, RPV pressure <200 psig at VB O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 VB, Early CF, DW, RPV pressure >200 psig at VB 3.01 E-09 O.OOE+00 O.OOE+00 0.OOE+00 VB Early CF. DW. RPV pressure <200 psig at VB 1.01 E-08 4.17E-08 1.56E-09 9.80E-09 VB, Late CF, WW 1.35E-10 2.27E-11 4.78E-12 O.OOE+00 VB Late CF,DW 3.27E-09 2.18E-08 4.60E-10 2.67E-09 VB VENT 1.17E-08 1.52E-08 1.70E-09 1.04E-08 B, No CF 5.29E-09 2.36E-09 8.OOE-10 5.56E-09 NOVB, No CF 1.23E-08 5.52E-09 1.85E-09 1.16E-08 NO CD, No CF O.OOE+00 O.OOE+00 O.OOE+00 0.OOE+00 Total- 04.68E-080 8.66E-080 l:(06.38E-09 : O4.OE-08 Collapsed Accident Progression Bins - Frequencylry I PDS-9 I PSD-10 PDS-11 I Total B,Early CF,VWW, RPV pressure >200 psig at VB IO.OOE+00I 2.27E-081 1.17E-07 1.43E-07 B Early CF. WW RPV pressure <200 psig at VB 1.33E-09 0.OOE+00 9.40E-10 2.27E-09 A7

B, Early CF, DW, RPV pressure >200 psig at VB O.OOE+0O 1.39E-07 1.64E-07 5.41E-07 B, Early CF, DW, RPV pressure <200 psig at VB 7.46E-09 O.OOE+00 9.71E-09 3.86E-07 B, Late CF. WW O.OOE+00 O.OOE+O0 O.OOE+0 3.09E-10 Late CF, DW O.OOE+00 O.OOE+00 O.OOE+00 1.31E-07

• VENT 7.39E-09 1.19E-07 8.20E-09 S.45E-07

• No CF O.OOE+00 O.OOE+00 O.OOE+00 2.92E-07 No VB, No CF O.OOE+00 O.OOE+00 O.OOE+00 4.OOE-07 No CD, No CF O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+O0

-~ft- :Lt i :iS: EtiE; :i;-:E

---F tl=1I .62E-08 2 Total 62-04 0E 2.81 E-07 .0E-0 l 3.OOE- - .T4.8-062.06 A8

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Figure A5 - JAF ILRT Evaluation PDS-5 Base Results A13

Figure A6 - JAF ILRT Evaluation PDS-6 Base Results A14

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• U A19

Attachment B FitzPatrick Risk Impact of Containment Liner Corrosion During an Extension of the ILRT Interval Results

Table of Contents Page No.

B13.0 Introduction B1 B2.0 Sensitivity Case 1 - Flaw rate doubles every 2 years B3 B3.0 Sensitivity Case 2 - Flaw rate doubles every 10 years B6 B4.0 Sensitivity Case 3 - 5% Visual inspection failures B9 B5.0 Sensitivity Case 4 - 15% Visual inspection failures B12 B6.0 Sensitivity Case 5 - Containment breach base point 10 times lower B15 B7.0 Sensitivity Case 6 - Containment breach base point 10 times higher B18 B8.0 Sensitivity Case 7 - Lower bound B21 B9.0 Sensitivity Case 8 - Upper bound) B24 BI

B1.0 Introduction This attachment presents the results of the FitzPatrick risk impact of containment liner corrosion during an extension of the ILRT interval. Seven sensitivity cases were examined. These are:

• Sensitivity Case I - Flaw rate doubles every 2 years

• Sensitivity Case 2 - Flaw rate doubles every 10 years

• Sensitivity Case 3 - 5% Visual inspection failures

• Sensitivity Case 4 - 15% Visual inspection failures

• Sensitivity Case 5 - Containment breach base point 10 times lower

• Sensitivity Case 6 - Containment breach base point 10 times higher

• Sensitivity Case 7 - Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF (Lower bound)

• Sensitivity Case 8 - Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF (upper bound)

The EXCEL spreadsheet results are presented in the following sections.

B2

B2.0 Sensitivity Case I - Flaw Rate Doubles Every 2 Years 3-in-1 0 years From Estimated Change Inaccessible Drywell/Torus DWlTorus Drywell Floor 1 to 3 years 0.20% 0.20% 0.05%

1 to 10 years 3.46% 3.46% 0.86%

I to 15 years 20.07% 20.07% 5.02%

Other Assumptlons:

Containment Breach 0.8171% 0.8171% 0.0817%

Visual Inspection Failures 10.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies DrywellrTorus DWITorus Drywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00017% 0.00167% 0.00004% 0.00187%

0.00187%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry 3 in 101 yrs Person-rem/ry 1 2.91 E+03 6.18E-07 1.80E-03 2 6.79E+05 2.44E-09 Corrosion Aiddition 1.66E-03 3a 2.91 E+04 6.59E-08 O.OOE+C 10 1.92E-03 3b 1.02E+05 6.63E-09 4.57E-1 11 6.77E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 NIA O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 To0l~',ta

',',>^ 24,4E-06, g Risk Cont :ribution: 0.36%

From 3a and 3b: 2.60E-03 3 b LERF: 6.63E-09 CCFP: 71.98%

B3

1-in-10 years Increases to 3a and 3b Frequencies Drywell/Torus Inaccessible DWJTorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.00000%

0.0028% 0.0282% 0.0007% 0.03177%

0.03177%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 10 yrs Person-remiry 1 2.91 E+03 4.49E-07 1.31 E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 2.20E-07 0.OOE+00 6.40E-03 3b 1.02E+05 2.27E-08 7.75E-10 2.32E-03 4 N/A O.OOE+00 O.OOE+00 5 NIA O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01

[ Total 2.44E-06

,a 0.720 Risk Contribution: 1.21%

From 3a and 3b: 8.72E-03 3b LERF: 2.27E-08 1-in-1 5 years Increases to 3a and 3b FrequenclesDrywell/Torusinaccessible DWlTorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.00000%

0.0164% 0.1640% 0.0041% 0.18447%

0.18447%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 15 yrs Person-remlry I 2.91 E+03 3.24E-07 9.45E-04 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 3.29E-07 O.OOE+00 9.60E-03 3b 1.02E+05 3.74E-08 4.50E-09 3.82E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 Total 2.44E-06 0.-24' Risk Contribution: 1.85%

From 3a and 3b: 1.34E-02 3b LERF: 3.74E-08 B4

Other Pertinent Risk Metrics 10 to 15 Increase (Person-remlry).1  : i; -4.34E-07 3 to 15 Increase (Person-rem/ry): 9.97E-03 10 to 15 Delta-LERF:1 1.47E-8 3 to 15 Delta-LERF: 3.08E-08 10 to 15 Delta-CCFP- 0.160 3 to 15 Delta-CCFP: 1.26%

3 to 15 Delta-LERF from Corrosion: 3.08E-08 10 to 15 Delta-LERF from Corrosion: 1.47E-08 Increase in LERF (ILRT 3-to-15 years) 5.32E-09 B5

B3.0 Sensitivity Case 2 - Flaw Rate Doubles Every 10 Years 3-in-10 years From Estimated Change Inaccessible Drywell/Torus DW/Torus Drywell Floor 1 to 3 years 1.06% 1.06% 0.26%

1 to IO years 4.58% 1.06% 1.15%

I to 15 years 8.38% 1.06% 2.10%

Other Assumptions:

Containment Breach 0.8171% 0.8171% 0.0817%

Visual Inspection Failures 10.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Frequencies Total 0.00000% 0.00000% 0.00000% 0.00000%

0.000&6% 0.00865% 0.00022% 0.00973%

0.00973%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry 3 In 10 yrs Person-remJry 1 2.91 E+03 6.18E-07 1.80E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 6.59E-08 0.OOE+00 1.92E-03 3b 1.02E+05 6.83E-09 2.37E-10 6.96E-04 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 6.11E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 rF7 T d. 'RY ,~2.ME-'06 II 7,

}

ta 59.

I . .. ..... '!0. 147 Risk Contribution: 0.37%

From 3a and 3b: 2.62E-03 3b LERF: 6.83E-09 CCFP: 71.99%

B6

1-in-10 years Increases to 3a and 3b FrequenciesDrywelllTorusinaccessible DWlTorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.00000%

0.0037% 0.0086% 0.0009% 0.01333%

0.01333%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I in 10 yrs Person-remiry I 2.91 E+03 4.49E-07 1.31 E-03 2 6.79E+05 2.44E-09 Corrosion Addi tion 1.66E-03 3a 2.91 E+04 2.20E-07 O.OOE+00 6.40E-03 3b 1.02E+05 2.23E-08 3.25E-10 2.27E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 -,z r' T6t I ~, 2.44E-06 ^ 0.72OZ' Risk Contribution: 1.20%

From 3a and 3b: 8.67E-03 3b LERF: 2.23E-08 1-in-15 years Increases to 3a and 3b FrequenclesDrywelllTorusinaccessible DW/TorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.00000%

0.0069% 0.0086% 0.0017% 0.01721%

0.01721%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 15 yrs Person-remiry 1 2.91 E+03 3.28E-07 9.57E-04 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 3.29E-07 O.OOE+00 9.60E-03 3b 1.02E+05 3.34E-08 4.20E-10 3AOE-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 Total 2.44 -6 Risk Contribution: 1.80%

From 3a and 3b: 1.30E-02 3b LERF: 3.34E-08 B7

Other Pertinent Risk Metrics 10 to 15 Increase (Person-rem/ry):r, ,77=98,E-038 3 to 15 Increase (Person-rem/ry): 9.54E-03 10 to 15 Delta-LERF:I 1.11E-8 3 to 15 Delta-LERF: 2.65E-08 10 to 15 Delta-CCFP: '~7] ,¢-Q.,A45%1 0tj77 3 to 15 Delta-CCFP: 1.09%

3 to 15 Defta-LERF from Corrosion: 2.65E-08 10 to 15 Delta-LERF from Corrosion: 1.IME-08 Increase in LERF (ILRT 3-to-15 years) 9.83E-10 B8

B4.0 Sensitivity Case 3 - 5% Visual Inspection Failures 34n-10 years From Estimated Change Inaccessible Drywellrlorus DWlTorus Drywell Floor I to 3 years 0.71% 0.71% 0.18%

1 to 10 years 4.14% 4.14% 1.04%

1 to 15 years 9.68% 9.68% 2.42%

Other Assumptions:

Containment Breach 0.8171% 0.8171% 0.0817%

Visual Inspection Failures 5.0%1 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies Drywell/Torus DWITorus Drywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0003% 0.0058% 0.0001% 0.0063%

0.0063%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequency/ry 3 In 10 yrs Person-rem/ry 1 2.91 E+03 6.18E-07 1.80E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 6.59E-08 0.OOE+00 1.92E-03 3b 1.02E+05 6.74E-09 1.53E-10 6.88E-04 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 ,,,t-4.46E+05

+ -.t I 2.97E-07 1.32E-01 S tal 2:44E-06 Risk Contribution: 0.36%

From 3a and 3b: 2.61 E-03 3b LERF: 6.74E-09 CCFP: 71.99%

B9

1-ln-1O years Increases to 3a and 3b FrequenciesDrywellJTorusinaccessible DW/TorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0017% 0.0339% 0.0008% 0.0025%

0.0025%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 10 yrs Person -remiry I 2.91 E+03 4.49E-07 1.31E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 2.20E-07 O.OOE+00 6.40E-03 3b 1.02E+05 2.20E-08 6.20E-1 I 2.25E-03 4 NIA O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 NIA O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 i Totoal 2A4E-06_

iWz. 0.72..

l ,&

-z 4o , ..........

AL, i...

Risk Contribution: 1.20%

From 3a and 3b: 8.65E-03 3b LERF: 2.20E-08 1-ln-15 years Increases to 3a and 3b FrequenclesDryweiilTorusinaccessible DW/rorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0040% 0.0791% 0.0020% 0.0059%

0.0059%

Release type FItzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 15 yrs Person-remlry I 2.91 E+03 3.29E-07 9.58E-04 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 3.29E-07 O.OOE+00 9.60E-03 3b 1.02E+05 3.31 E-08 1A5E-10 3.38E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01

- w,44-r-6 T TotalI ~0.7242 Risk Contribution: 1.79%

From 3a and 3b: 1.30E-02 3b LERF: 3.31 E-08 BlO

Other Pertinent Risk Metrics 10 to 15 Increase (Person-rem/ry):'i 3798E-03 3 to 15 Increase (Person-remfry): 9.52E-03 10 to 15 Delta-LERF:7 1IE0 3 to 15 Delta-LERF: 2.63E-08 10 to 15 Delta-CCFP:Lr,,,,,,, 0.45%

3 to 15 Deita-CCFP: 1.08%

3 to 15 Delta-LERF from Corrosion: 2.63E-08 10 to 15 Delta-LERF from Corrosion: 1.IIE-08 Increase in LERF (ILRT 3-to15 years) 3.59E-10 Bll

B5.0 Sensitivity Case 4 - 15% Visual Inspection Failures 3-ln-10 years From Estimated Change Inaccessible Drywell/Torus DW/Torus Drywell Floor I to 3 years 0.71% 0.71% 0.18%

I to 10 years 4.14% 4.14% 1.04%

1 to 15 years 9.68% 9.68% 2.42%

Other Assumptions:

Containment Breach 0.8171% 0.8171% 0.0817%

Visual Inspection Failures 15.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.O%

Increases to 3a and 3b Inaccessible Frequencies DrywelilTorus DW/Torus Drywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0009% 0.0058% 0.0001% 0.0068%

0.0068%

Release ty; pe FitzPatrick Dos ae CDF Case Dose Person-rem Frequencylry 3 In 10 yrs Person-remiry 1 2.91 E+03 6.18E-07 1.80E-03 2 6.79E+05 2.44E-09 Corrosion Additiclon 1.66E-03 3a 2.91 E+04 6.59E-08 0.OOE+00 1.92E-03 3b 1.02E+05 6.75E-09 1.67E-10 6.89E-04 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 NIA 0.00E+00 0.00E+00 7a 6.11E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01

[11,- Total,

' 2'11 E...... ....

Risk Contribution: 0.37%

From 3a and 3b: 2.61 E-03 3b LERF: 6.75E-09 CCFP: 71.99%

B12

1-In-10 years Increases to 3a and 3b FrequenciesDrywelllTorusinaccessible DWITorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0051% 0.0339% 0.0008% 0.0059%

0.0059%

Release tyFle FitzPatrick Dose CDF Case Dose Person-rem IFrequencylry I In I10 yrs Person-remlry 1 2.91 E+03 4.49E-07 1.31 E-03 2 6.79E+05 2.44E-09 Corrosloi nAddition 1.66E-03 3a 2.91 E+04 2.20E-07 0.001 E+00 6.40E-03 3b 1.02E+05 2.21 E-08 1.45 E-10 2.25E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 NIA O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 r Total'-- i:-;: 771.44E40

,j't: 6'r 0.7202 Risk Contribution: 1.20%

From 3a and 3b: 8.66E-03 3b LERF: 2.21 E-08 1-In-1 5 years Increases to 3a and 3b FrequenciesDrywellrTorusinaccessible DWITorusDrywell Floor Total 0.0000% 0.0000% 0.0000% 0.0000%

0.0119% 0.0791% 0.0020% 0.0138%

0.0138%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 15 yrs Person-remlry 1 2.91 E+03 3.28E-07 9.57E-04 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 3.29E-07 O.OOE+00 9.60E-03 3b 1.02E+05 3.33E-08 3.38E-10 3.39E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 2.4 -0 6:

Total O;l.~07242; ;

Risk Contribution: 1.79%

From 3a and 3b: 1.30E-02 3b LERF: 3.33E-08 B13

Other Pertinent Risk Metrics; 10 to 15 Increase (Person-rem/ry):V-, .99E- 3 3 to 15 Increase (Person-rem/ry): 9.54E-03 10 to 15 Deita-LERF: R', it2Ed O8 3 to 15 Delta-LERF: 2.65E-08 10 to 15 Delta-CCFP:. 0A6%

3 to 15 Delta-CCFP: 1.09%

3 to 15 Delta-LERF from Corrosion: 2.65E-08 10 to 15 Delta-LERF from Corrosion: 1.12E-08 Increase in LERF (ILRT 3-to-15 years) 6.49E-10 B14

B6.0 Sensitivity Case 5 - Containment Breach Base Point 10 Times Lower 3-in-1 0 years From Estimated Change Inaccessible DrywelU/Torus DWITorus Drywell Floor I to 3 years 0.71% 0.71% 0.18%

1 to 10 years 4.14% 4.14% 1.04%

I to 15 years 9.68% 9.68% 2.42%

Other Assumptions:

Containment Breach 0.1663% 0.1663% 0.0166%

Visual Inspection Failures 10.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies Drywell/Torus DWITorus Drywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00012% 0.00118% 0.00003% 0.00133%

0.00133%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry 3 In 10 yrs Person-remlry 1 2.91 E+03 6.18E-07 1.80E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 6.59E-08 O.OOE+00 1.92E-03 3b 1.02E+05 6.62E-09 3.25E-11 6.75E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 I 'l

...ota

. -'LTT. -0. 6

2. 44EH- .i 0.7146 Risk Contribution: 0.3632%

From 3a and 3b: 2.60E-03 3b LERF: 6.62E-09 CCFP: 71.98%

B15

14n-10 years Increases to 3a and 3b FrequenclesDrywell/TorusInaccessible DWJTorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00069% 0.00689% 0.00017% 0.00775%

0.00775%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry 1 In 10 yrs Person-remlry 1 2.91 E+03 4.49E-07 1.31E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 2.20E-07 O.OOE+00 6.40E-03 3b 1.02E+05 2.21E-08 1.89E-10 2.26E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91E-01 7c 5.21E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 To~+-

tjal* X~.s ~ e i 0.7202 1 Risk Contribution: 1.2024%

From 3a and 3b: 8.66E-03 3b LERF: 2.21 E-08 1-in-15 years Increases to 3a and 3b FrequenclesDrywelllTorusinaccessible DWrrorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00161% 0.01610% 0.00040% 0.01811%

0.01811%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I I n 15 yrs Person-remlry 1 2.91 E+03 3.28E-07 9.57E-04 2 6.79E+05 2.44E-09 Corros Ion Addition 1.66E-03 3a 2.91 E+04 3.29E-07 0. OOE+00 9.60E-03 3b 1.02E+05 3.34E-08 4 .42E-10 3.41 E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 Total 2.44E-06 *.7242 Risi k Contribution: 1.7959%

From 3a and 3b: 1.30E-02 3b LERF: 3.34E-08 B16

Other Pertinent Risk Metrics 10 to 15 Increase (Person-remIry):t3.99E, 3 to 15 Increase (Person-rem/ry): 9.56E-03 10 to 15 Delta-LERF:j:,12E-08 3 to 15 Delta-LERF: 2.68E-08 10 to 15 Delta-CCFPPr:¶.i t 3 to 15 Delta-CCFP: 1.10%

3 to 15 Delta-LERF from Corrosion: 2.68E-08 10 to 15 Delta-LERF from Corrosion: 1.12E-08 Increase in LERF (ILRT 3-to-15 years) 6.64E-10 B17

B7.0 Sensitivity Case 6 - Containment Breach Base Point 10 Times Higher 3-ln-10 years From Estimated Change Inaccessible DrywellUTorus DWITorus Drywell Floor I to 3 years 0.71% 0.71% 0.18%

I to 10 years 4.14% 4.14% 1.04%

1 to 15 years 9.68% 9.68% 2.42%

Other Assumptions:

Containment Breach 4.0703% 4.0703% 0.4070%

Visual Inspection Failures 10.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies DrywellUTorus DWITorus Drywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00290% 0.02896% 0.00072% 0.03258%

0.03258%

Release type FItzPatrick Dose CDF Case Dose Person-rem Frequencylry 3 In 10 yrs Person-remlry I 2.91 E+03 6.18E-07 1.80E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 6.592-08 0.00E+00 1.92E-03 3b 1.02E+05 7.38E-09 7.95E-10 7.53E-04 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.002+00 0.00E+00 6 N/A 0.00E+00 0.00E+00 7a 6.11E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 Tf otal 0il,~ - 444 -06

' 1 2. -  ;:-:0.7147-.:: -~

Risk Contribution: 0.3740%

From 3a and 3b: 2.67E-03 3b LERF: 7.38E-09 CCFP: 72.01%

B18

14in-10 years Increases to 3a and 3b FrequenciesDrywellrTorusinaccessible DW/TorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.01687% 0.16867% 0.00422% 0.18976%

0.18976%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 10 yrs Person-remiry I 2.91 E+03 4.45E-07 1.30E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 2.20E-07 O.OOE+00 6.40E-03 3b 1.02E+05 2.66E-08 4.63E-09 2.71 E-03 4 N/A 0.OOE.00 O.OOE+00 5 N/A 0.OOE.00 O.OOE+00 6 N/A 0.00E400 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01

[, TotalF I 2.4E Risk Contribution: 1.2646%

From 3a and 3b: 9.11E-03 3b LERF: 2.66E-08 1-ln-15 years Increases to 3a and 3b FrequenciesDrywellrroruslnaccessible DWlTorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.03940% 0.39396% 0.00985% 0.44321%

0.44321%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 15yrs Person-remiry 1 2.91 E+03 3.18E-07 9.26E-04 2 6.79E+05 2.44E-09 Corrosion Addi ibon 1.66E-03 3a 2.91 E+04 3.29E-07 O.OOE+00 9.60E-03 3b 1.02E+05 4.38E-08 1.08E-08 4.46E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 r-to-a iF, F 7 2 A, E-06

r. giz,<,4] X0.7252 =

Risk Contribtition: 1.9393%

From 3a an d 3b: 1.41E-02 3b LERF: 4.38E-08 B19

Other Pertinent Risk Metrics 10 to 15 Increase (Person-reml/ry):L 4iE3 3 to 15 Increase (Person-rem/ry): 1.05E-02 10 to 15 DeIta-LERF:17, t 172 v 8 3 to 15 Delta-LERF: 3.64E-08 10 to 15 Defa-CCFP - E 3 to 15 Delta-CCFP: 1.49%

3 to 15 Delta-LERF from Corrosion: 3.64E-08 10 to 15 Delta-LERF from Corrosion: 1.72E-08 Increase in LERF (ILRT 34o-15 years) 1.62E-08 B20

B8.0 Sensitivity Case 7 - Lower bound (Flaw rate doubles every 10 years, containment breach base point 10 times lower, 5% visual inspection failures and 10% EPRI accident Class 3b are LERF) 3-In-10 years From Estimated Change Inaccessible Drywell/Torus DWlTorus Drywell Floor I to 3 years 1.06% 1.06% 0.26%

1 to 10 years 4.58% 4.58% 1.15%

I to 15 years 8.38% 8.38% 2.10%

Other Assumptions:

Containment Breach 0.1663% 0.1663% 0.0166%

Visual Inspection Failures 5.0% 100.0% 100.0%

EPRI Class 3a Fraction 90.0% 90.0% 90.0%

EPRI Class 3b Fraction 10.0% 10.0% 10.0%

Increases to 3a and 3b Inaccessible Frequencies Drywell/Torus DWITorus Drywell Floor Total 0.00008% 0.00158% 0.00004% 0.00170%

0.00001% 0.00018% 0.00000% 0.00019%

0.00189%

Release type FItzPatrick Doi se CDF Case Dose Person-rem Frequency/ry 3 In 10 yrs Person-remiry I 2.91 E+03 6.18E-07 1.80E-03 2 6.79E+05 2.44E-09 C 0 rroslon Addition 1.66E-03 3a 2.91 E+04 6.59E-08 4.16E-11 1.92E-03 3b 1.02E+05 6.59E-09 4.62E-12 6.73E-04 4 N/A 0.OOE+00 0.OOE+00 5 N/A 0.OOE+00 0.OOE+00 6 N/A 0.OOE+00 0.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 illTotal - 2i.4o 06 1~0.7146 Risk Contribution: 0.3630%

From 3a and 3b: 2.59E-03 3b LERF: 6.59E-09 CCFP: 71.98%

B21

1-In-10 years Increases to 3a and 3b FrequenciesDrywelllforusinaccessible DWITorusDrywell Floor Total 0.00034% 0.00686% 0.00017% 0.00737%

0.00004% 0.00076% 0.00002% 0.00082%

0.00819%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 10 yrs Person-remiry 1 2.91 E+03 4.49E-07 1.31E-03 2 6.79E+05 2.44E-09 C orroslon Addition 1.66E-03 3a 2.91 E+04 2.20E-07 1.80E-10 6.41 E-03 3b 1.02E+05 2.20E-08 2.OOE-1 1 2.24E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05

1. -1^

2.97E-07 1.32E-01 Tota-t-77 2.44E-O&

Risk Contribution: 1.2008%

From 3a and 3b: 8.65E-03 3b LERF: 2.20E-08 1-in-15 years Increases to 3a and 3b FrequencIesDryweliTorusinaccessible DWITorusDrywell Floor Total 0.00063% 0.01255% 0.00031% 0.00094%

0.00007% 0.00139% 0.00003% 0.00150%

0.00244%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry 1 Inn5 yrs Person-remiry I 2.91 E+03 3.29E-07 9.58E-04 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 3.29E-07 2.30E-1 1 9.60E-03 3b 1.02E+05 3.30E-08 3.66E-1 1 3.36E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 V1 Toutal'~:! ^ 2.441 ~--...- :7 -.;- 7 71 7XW

. . .7242 Risk Contribution: 1.7904%

From 3a and 3b: 1.30E-02 3b LERF: 3.30E-08 B22

Other Pertinent Risk Metrics 10 5Increae' . n.

10 to 15 Increase (Person-rem/ry):E7; 3 to 15 Increase (Person-rem/ry): 9.53E-03 10 to 15 Delta-LERF:..1.1OE-08 3 to 15 Delta-LERF: 2.64E-08 10 to 15 Delta-CCFP:.=.-. -0.450%'

3 to 15 Delta-CCFP: 1.08%

3 to 15 Delta-LERF from Corrosion: 2.64E-08 10 to 15 Delta-LERF from Corrosion: 1.10E-08 Increase in LERF (ILRT 3-to-15 years) 3.06E-10 B23

B9.0 Sensitivity Case 8 - Upper Bound (Flaw rate doubles every 2 years, containment breach base point 10 times higher, 15% visual inspection failures and 100% EPRI accident Class 3b are LERF) 3-in-10 vears From Estimated Change Inaccessible DrywelltTorus DWlTorus Drywell Floor I to 3 years 0.20% 0.20% 0.05%

1 to IO years 3.46% 3.46% 0.86%

I to 15 years 20.07% 20.07% 5.02%

Other Assumptions:

Containment Breach 4.0703% 4.0703% 0.4070%

Visual Inspection Failures 15.0% 100.0% 100.0%

EPRI Class 3a Fraction 0.0% 0.0% 0.0%

EPRI Class 3b Fraction 100.0% 100.0% 100.0%

Increases to 3a and 3b Inaccessible Frequencies Drywell/Torus DWJTorus Drywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.00124% 0.00830% 0.00021% 0.00975%

0.00975%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry 3 In 10 yrs Person-remtry 2.91 E+03 6.18E-07 1.80E-03 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 6.59E-08 O.OOE+00 1.92E-03 3b 1.02E+05 6.83E-09 2.38E-10 6.96E-04 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01

_.rprm--w 244E=t!^!-_

6 X>t Tb otal I Risk Contribution: 0.3661%

From 3a and 3b: 2.62E-03 3b LERF: 6.83E-09 CCFP: 71.99%

B24

1-in-10 years Increases to 3a and 3b FrequenciesDrywell/Torusinaccessible DWlTorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.02110% 0.14066% 0.00352% 0.16527%

0.16527%

Release type FitzPatrick Dose CDF Case Dose Person-rem Frequencylry I In 10 yrs Person-remiry I 2.91 E+03 4.45E-07 1.30E-03 2 6.79E+05 2.44E-09 Corrosion ALdd itlon 1.66E-03 3a 2.91 E+04 2.20E-07 O.OOE+ 00 6.40E-03 3b 1.02E+05 2.60E-08 4.03E- 09 2.65E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 F7-e r-T l r-jIt +'7 Total -t224E 06 -r-7 w 4s..:E.._

_.. 4. -7 -F hfHY.7n!itet O.206!17 Risk Contribiution: 1.2562%

From 3a arid 3b: 9.05E-03 3b ILERF: 2.60E-08 1-in-1 years Increases to 3a and 3b FrequenclesDrywelllTorusInaccessible DWJTorusDrywell Floor Total 0.00000% 0.00000% 0.00000% 0.00000%

0.12252% 0.81681% 0.02042% 0.95975%

0.95975%

Release type FitzPatrick Dcose CDF Case Dose Person-ren Frequencylry I In 15 yrs Person-remlry 1 2.91 E+03 3.05E-07 8.90E-04 2 6.79E+05 2.44E-09 Corrosion Addition 1.66E-03 3a 2.91 E+04 3.29E-07 O.OOE+00 9.60E-03 3b 1.02E+05 5.64E-08 2.34E-08 5.75E-03 4 N/A O.OOE+00 O.OOE+00 5 N/A O.OOE+00 O.OOE+00 6 N/A O.OOE+00 O.OOE+00 7a 6.11 E+05 1.74E-07 1.06E-01 7b 3.96E+05 7.37E-07 2.91 E-01 7c 5.21 E+05 5.82E-08 3.04E-02 7d 3.06E+05 4.83E-07 1.48E-01 8 4.46E+05 2.97E-07 1.32E-01 TotaV Z.4 4E-0B-Risk Contribution 2.1129%

From 3a and 3b 1.54E-02 3b LERF 5.64E-08 B25

Other Pertinent Risk Metrics 10 to 15 Increase (Person-rem/ry): . .89E-03 3 to 15 Increase (Person-rem/ry): 1.18E-02 10 to 15 Delta-LERF{ -. i4E-Q8 3 to 15 Delta-LERF: 4.95E-08 10 to 15 Delta-CCFP: ' 71 3 to 15 Delta-CCFP: 2.03%

3 to 15 Delta-LERF from Corrosion: 4.95E-08 10 to 15 Delta-LERF from Corrosion: 3.04E-08 Increase in LERF (ILRT 3-to-15 years) 2.77E-08 B26

ATTACHMENT 5 to JAFP-03-0108 CONTAINMENT INSERVICE INSPECTION PROGRAM

SUMMARY

Proposed License Amendment to Provide a One-time Integrated Leak Rate Test (ILRT)

Interval Extension Entergy Nuclear Operations, Inc.

JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

Attachment 5 to JAFP-03-0108 CONTAINMENT INSERVICE INSPECTION PROGRAM

SUMMARY

Page 1 of 7 CONTAINMENT DESIGN James A. Fitzpatrick N.P.P. is a power uprated 881 Mwe, General Electric Boiling Water Reactor (BWR4). The reactor is contained in a Chicago Bridge and Iron Works supplied Mark 1, Free Standing Steel Containment Building. The Mark 1 Containment Building was designed by Stone and Webster Engineering Corporation (Contract No. AP-4).

The design, fabrication, inspection, and testing of the containment conforms to the requirements for Class B vessels in the 1968 Edition of Section III of the ASME Boiler and Pressure Vessel Code for Nuclear Vessels, including the 1968 Summer Addenda.

CONTAINMENT DESCRIPTION The Drywell is a free standing steel pressure vessel, with a spherical lower portion 65 feet in diameter, and a cylindrical upper portion 35 feet 7 inches in diameter. The overall height of the Drywell is approximately 111 feet 5 % inches. The Drywell is closed at the top by a curved head 30 feet 2 inches in diameter attached at the Drywell neck by a bolted flanged assembly. It is enclosed in reinforced concrete for shielding. There is a sand filled transition zone to accommodate for radial expansion of the spherical section of the freestanding vessel. Above the sand filled transition zone, the Drywell is separated from the reinforced concrete by an air gap of approximately 2 inches. Below the sand filled transition zone the reinforced concrete is in direct contact with the outside surface of the vessel. In addition to the Drywell head, there are two double-door Air Locks and two double-door Equipment Hatches and one Control Rod Drive Removal Hatch to provide access for people and equipment The suppression chamber (Torus) is a steel pressure vessel in the shape of a torus encircling the Drywell, with a major diameter of approximately 108 feet and a cross-sectional diameter of 29 feet 6 inches. The Torus has three hatch openings, one-48 inch diameter and two-24 inch diameter for personnel and equipment access. The shell is stiffened by 16 internal ring girders located at the miter joints. The Torus is supported by 16 pairs of reinforced W12x161 columns at the ring girder locations. As a result of the Mark I Containment Program, a vent header deflector, saddle plates under the Torus, and anchor bolts with tie-down structures were added.

A system of large and small pipes connects the Drywell to the Torus/water. Eight large circular pipes (Vent Pipes), each with a diameter of 6 feet 9 inches, connect the Drywell vessel to the Torus vessel. These eight large circular vent pipes ultimately connect to a circular header vessel (Ring Header) inside the Torus vessel airspace. The eight large circular vent pipes direct steam and vapor following a design bases accident to the Torus ring header. The circular ring header, with a diameter of 4 feet 9 inches directs the steam and vapor through ninety-six, 2 foot diameter pipes, which terminate 4 foot below the Torus normal water level in the Torus vessel.

There are eight bellows type expansion joints located inside the Torus vessel that accommodate thermal expansion and vibrational movement between the Drywell vessel and the Torus vessel.

Attachment 5 to JAFP-03-0108 CONTAINMENT INSERVICE INSPECTION PROGRAM

SUMMARY

Page 2 of 7 CONTAINMENT INSPECTIONS A. Background Structural degradation of containment is a gradual process that occurs due to the effects of pressure, temperature, radiation, chemical or other such effects. Such effects would be identified and corrected when the containment structure is periodically tested and inspected to verify structural integrity.

Effective September, 1996, the NRC endorsed Subsections IWE and IWL of ASME Section XI, 1992 Edition including 1992 Addenda. These subsections contain inservice inspection and repair/replacement rules for Class MC and Class CC components. The reactor containment is a free-standing steel containment, to which only the requirements of Subsection IWE apply.

The First Period (September 28, 1997-March 27, 2001) scheduled IWE examinations were performed in accordance with the requirements of the 1992 Edition, 1992 Addenda, and 10CFR50.55a. The Second Period (March 28, 2001 - September 27, 2004) and Third Period (September 28, 2004 - September 27, 2006), are under the requirements of the 1998 Edition (no Addenda) and 10CFR50.55a.

Relief Requests in effect are:

1. TAC No. MA5399 granted relief from the visual inspection of seals and gaskets, Examination Category E-D, Item No. E5.10 and E5.20. A pressure test in accordance with 10CFR50, Appendix J is substituted for these inspections.

2. TAC No. MA9112 granted to continue initial certification and re-certification of ultrasonic testing personnel in accordance with the requirements contained in the 1989 Edition of ASME,Section XI until June 30, 2001.

3. TAC No. MB2946 granted the use of ASME Section XI, Subsection Xl, 1998 Edition (no Addenda).

The First Containment Inservice Inspection Period required a General Visual examination be performed to satisfy the requirements stated in I OCFR50.55a (b)(2)(ix)(E). The General Visual examination was performed either directly or remotely by persons qualified in IWE Containment Inspection. The First Containment Inservice Inspection Period satisfied the required expedited examination of containment, outlined in IOCFR50.55a (g)(6)(ii) to be completed prior to September 9, 2001.

General Visual examinations are performed by engineering personnel knowledgeable in the requirements for design, service, inservice inspection , and/or testing of Class MC components.

Detailed Visual examinations are performed by personnel meeting the applicable requirements of IWA-2300 of the 1992 Addenda, in accordance with CP-189, 1991 Edition.

Attachment 5 to JAFP-03-0108 CONTAINMENT INSERVICE INSPECTION PROGRAM

SUMMARY

Page 3 of 7 Acceptance of components for continued service is subject to the rules of IWE-3000. Work Requests were used to document the IWE examinations. "Pre-Evaluated IWE Inspection Criteria for R013 Outage' (Ref. 1) and "ASME Section Xi, Subsection IWE, General Visual Containment Inspection Procedure and Pre-Evaluated IWE Inspection Criteria" (Ref. 2) were used to document the IWE method of examination, including acceptance criteria, for the First and Second Periods respectively.

Pre-evaluated Inspection criteria for Ref. (1)

Paint/Coating distress

a. Chipping, chalking, checking, minor pinpoint rust, and minor surface rust are acceptable

b. Flaking/peeling, blistering, major pinpoint rust, undercutting, medium surface rust, and major surface rust require evaluation Pitting / Gouges / Arc Strikes / Wear / Surface Discontinuities (base metal loss)

a. Minor surface, less than 10% of base metal, is acceptable

b. Medium surface, greater than 10% of base metal, requires evaluation

c. Majorsurface, greater than 10% of base material, requires evaluation

d. Cracking, requires evaluation Dents (no signs of base metal stress or paint/coating distress)

a. Minor dent, less than 10% of base metal thickness, is acceptable

b. Major dent, greater than 10% of base metal thickness, requires evaluation Pre-evaluated Inspection criteria for Ref. (2)

Uncoated surface areas - If any of the relevant conditions listed below are present, further evaluation may be required:

a. Cracking in base metal

b. Discoloration resulting from age, heat, or corrosion

c. Wear exceeding 10% of the design wall thickness

d. Pits, dents, or gouges of the base metal with depth exceeding 10% of the design wall thickness

e. Corrosion which results in discemable base metal loss exceeding 10% of the design wall thickness

Attachment 6 to JAFP-03-0108 CONTAINMENT INSERVICE INSPECTION PROGRAM

SUMMARY

Page 4 of 7

f. Discemable bulges

g. Arc strikes Coated surface areas

a. Any of the conditions listed above for uncoated surfaces

b. Absence of coating

c. Blisters equal to or greater than size No. 6 as specified in ASTM D 714

d. Checking equal to or greater than standard No. 2 as specified in ASTM D 660

e. Cracking equal to or greater than standard No. 6 as specified in ASTM D 661

f. Flaking equal to or greater than standard No. 6 as specified in ASTM D 772

g. Rusting equal to or greater than Rust Grade 7 as specified in ASTM D 610 Bolting assemblies

a. Bending, twisting, stretching or deforming of bolts or studs

b. Missing or loose bolts, studs, nuts, or washers

c. Fractured bolts, studs, or nuts

d. Degradation of protective coatings on bolting surfaces

e. Evidence of coolant leakage near bolting

f. Localized excessive corrosion

g. Misalignment of connection or bolting Containment supports

a. Any signs of surface irregularities

b. Deformations or structural degradations of fasteners, clamps or other support items, and loss of integrity at bolted or welded connections

c. Missing, detached, or loose support parts and bolting

d. Arc strikes, weld spatter, paint, scoring, roughness, or general corrosion on close tolerance, machined or sliding surfaces

e. Misalignment of supports

Attachment 5 to JAFP-03-0108 CONTAINMENT INSERVICE INSPECTION PROGRAM

SUMMARY

Page 5 of 7

f. Improper clearances of guides and stops

g. Wear which visibly reduces the cross-sectional area of the support

h. Abnormal corrosion which reduces the load bearing capacity of the support

i. Cracklike or linear surface flaws

j. Evidence of clamp or non-integral attachment movement, damage, or movement of component insulation due to support movement.

Using industry experience as a guide, evaluations were done of possible suspect surfaces to identify areas which would require augmented examinations (IN 84-01, NRC Generic Letter 87-05, etc.) Currently there are no identified/scheduled augmented inspections.

No conditions existed in the accessible areas which could indicate the presence of, or result in degradation to such inaccessible areas.

B. Drywell The most recent IWE inspections of the Drywell were performed during R013, R014 (First Period), and RO15 (Second Period). The Second Period Drywell inspections will be completed during RO16, October, 2004. These inspections provide a high degree of assurance that any degradation of the containment will be detected and corrected before it may provide a containment leakage path. The inspections to date have not identified degradation that threatens the structural integrity of the containment. All levels of the Drywell were inspected for coating degradation during R013 (Nov. 1998) and R014 (Oct.

2000). Areas showing peeling paint or blistering were videotaped. Peeling paint was scraped off of the located areas. In all these areas, evidence of well adhering carbozinc-1 I coating was apparent and no further repairs were deemed necessary.

In response to NRC Generic Letter 87-05, it was decided to inspect both the sand cushion drain lines and the bellow seal drain lines. These inspections were to ensure that the drain lines were unplugged and operational as designed.

Three of the four upper sand cushion drain lines were found to be operational . Only one is needed to perform the function that they were designed for, the draining of condensate which may form in the air gap.

Five of the six outer bellows drain lines were verified operational as designed. One line could not be inspected due to the line's inaccessibility. But, because of the redundancy involved in the design of the outer bellows drain lines, it was not deemed necessary to further pursue the inspection of the remaining liner drain.

Visual examination did not indicate present or past entry of moisture in these areas. The sand cushion at JAF is sealed from the air gap by a stainless steel plate and adhesive seal which would have directed any water into the air gap drain lines and prevented collection in the sand cushion.

Attachment 5 to JAFP-03-01 08 CONTAINMENT INSERVICE INSPECTION PROGRAM

SUMMARY

Page 6 of 7 C. Torus In 1981 extensive maintenance on Torus coating was done in conjunction with the Torus modifications. The Torus was desludged and inspected in 1989 and 1992. It was inspected by a contracted vendor in April, 1995 and in January, 1997 under the Torus Preservation Program.

In 1996, the NRC endorsed the use of Subsection IWE to ASME Section Xl which provided detailed requirements for inservice inspection of containment structure. IWE General Visual and VT-3 examinations of the Torus were done in October/November, 1998 by the ISI Containment engineer and a contracted vendor. This was done at a time when the Torus was drained for the installation of new strainers. Indications found were coating related and did not show degradation of structural integrity. An additional inspection in the area of the most severe coating loss (Bay B) was ultrasonically inspected to determine the actual thickness of the Torus shell plate. Results of the UT inspections found the shell thickness acceptable (readings ranged between 0.620" to 0.690" where the nominal plate thickness is 0.632").

The existing material condition of the Torus is acceptable and does not compromise the structural integrity of the primary pressure boundary. Underwater coating applications were done in April, 1995 and January, 1997. General Visual IWE inspections of the Torus were performed during R013, R014, and R015. The Second Period Drywell inspections will be completed during RO16, October, 2004. The inspection results indicate that no significant corrosion effects have been experienced on the containment vessel. The inspections conducted to date have not identified degradation that threatens the structural integrity of the containment. Various coating issues have been identified, evaluated and recoated as found necessary.

In addition to the [WE requirements, "Suppression Chamber and Drywell Deterioration Inspection, ST-15B, performed as part of overall Primary Containment Leakage Rate Testing Program, was last done during R014, October 2000, and concluded that the Drywell and Torus interior/exterior surfaces, including Torus supports, showed no evidence of degradation. The Maintenance Rule walk-downs of the interior / exterior surfaces of the primary containment are performed to insure structural integrity for both the Drywell and Torus. These inspections, also have not identified any structural degradation.

D. Pressure retaining bolting Per the 1998 Edition, pressure retaining bolting only requires General Visual examinations performed in place, Category E-A, per the 1998 Edition. This corresponds to an examination of all bolted connections three times per inspection interval (10 years). Bolted connections need not be disassembled for performance of examinations. But, JAF has committed (TAC NO MB2946, dated 5/1/2002) to perform an additional detailed visual examination on all accessible surface areas of pressure retaining bolted connections, including bolts, studs, nuts, bushings, washers, threads in base material and flange ligaments between fastener holes, using the following guidelines:

1. A detailed visual examination (VT-1) will be performed for areas where flaws or degradations, which exceed ASME Section Xl code requirements, are indicated.

Attachment 5 to JAFP-03-0108 CONTAINMENT INSERVICE INSPECTION PROGRAM

SUMMARY

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2. Pressure retaining bolting indications will be evaluated by the responsible individual for continuous services. If disassembly is required for further evaluation, then a detailed visual examination (VT-1) shall be performed.

3. A detailed visual examination (VT-1) will be performed is a bolted connection is disassembled at the time of a scheduled General Visual examination.

4. A detailed visual examination (VT-1) will be performed when a bolted connection is disassembled at times other than a scheduled visual examination. Procedures are in place to ensure that the integrity of the reassembled bolted connection is maintained.

The bolting examinations are performed during preventive maintenance activities of certain components. These maintenance activities are scheduled to support replacement of the seals and gaskets used in the component connections. Additionally, some of these connections are routinely used during outages, and the examination and testing of these connections is performed to re-establish containment integrity at the end of the outage. Any parts (except seals and gaskets which are exempt) that are replaced are subject to compliance with our Repair/Replacement Program and receive the appropriate inspections at that time.