JAFP-16-0170, Response to Request for Additional Information (RAI) Regarding License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type a and Type C Leak Rate Test Frequencies - Supplement 1

Response to Request for Additional Information (RAI) Regarding License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type a and Type C Leak Rate Test Frequencies - Supplement 1

ML16326A444
Person / Time
Site:	FitzPatrick  (DPR-059)
Issue date:	11/21/2016
From:	Brian Sullivan Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF8305, JAFP-16-0170
Download: ML16326A444 (91)

Text

JAFP-16-0170 November 21, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Response to Request for Additional Information (RAI) Regarding License Amendment Request to revise Technical Specifications Section 5.5.6 for Extension of Type A and Type C Leak Rate Test Frequencies - Supplement 1 (CAC No. MF8305)

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59

References:

1. ENOI letter, License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type A and Type C Leak Rate Test Frequencies, JAFP-16-0129, dated August 29, 2016 (ML16242A332)

2. NRC letter, James A. FitzPatrick Nuclear Power Plant-Request for Additional Information Regarding: License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type A and Type C Leak Rate Test Frequencies, dated October 21, 2016 (ML16287A650)

Dear Sir or Madam:

By letter dated August 29, 2016 [Reference 1], Entergy Nuclear Operations, Inc. (ENOI) requested NRC approval of a License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type A and Type C Leak Rate Test Frequencies. In processing the submittal, the NRC determined that additional information was required to complete the review [Reference 2]. The specific questions provided to JAF in the NRC request for additional information (RAI) are addressed in the Attachment and in Enclosure 1 of this letter. Revised pages to the Ref. 1 submittal are included in Enclosure 2 of this letter.

Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.

James A. FitzPatrick NPP P.O. Box 110 Lycoming, NY 13093 Tel 315-342-3840 Brian R. Sullivan Site Vice President - JAF

JAFP-16-0170 Page 2 of 2 There are no new regulatory commitments in this submittal. Should you have any questions please contact Mr. William G. Drews, Regulatory Assurance Manager at 315-349-6562.

I declare under penalty of perjury that the foregoing is true and correct. Executed on November 21, 2016.

rian

. Sullivan Site Vice President BRS/WGD/dc

Attachment:

Response to Request for Additional Information

Enclosures:

1. Evaluation of Risk Significance of Permanent ILRT Extension

2. Revised Pages cc:

NRG Region 1 Regional Administrator NRG Project Manager NRG Resident Inspector NYSPSG NYSERDA

JAFP-16-0170 Attachment Response to Request for Additional Information (12 Pages)

JAFP-16-0170 Attachment Response to Request for Additional Information Page 1 of 12 REQUEST FOR ADDITIONAL INFORMATION (RAI)

REGARDING LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS SECTION 5.5.6 ROGRAM JAMES A. FITZPATRICK NUCLEAR POWER PLANT ENTERGY NUCLEAR OPERATIONS, INC DOCKET NO. 50-333 RENEWED FACILITY OPERATING LICENSE NO. DPR-59 In an application dated August 29, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16242A332), Entergy Nuclear Operations, Inc. (Entergy),

submitted an application for a proposed amendment to the Technical Specifications (TSs) for James A. FitzPatrick Nuclear Power Plant (FitzPatrick). The proposed amendment would revise the TS 5.5.6 Primary Containment Leak Rate Testing Program to allow a permanent extension of the Type A Primary Containment Integrated Leak Rate Test interval to 15 years and allow an extension of Type C Local Leak Rate testing interval up to 75 months. The Nuclear Regulatory Commission staff is reviewing the submittal and has the following questions:

SBPB-RAI-1 The license amendment request (LAR) page 32 states In accordance with TS 5.5.6, the allowable maximum pathway total Types B and C leakage is 0.6 La where La equals 320 SLM. Thus the performance criterion of 0.6 La for evaluating combined type B and C leakage would be 192 SLM. However, TS 5.5.6 states During plant startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and Type C tests. NEI 94-01 Section 8.0 states:

ANSI/ANS 56.8-1994 also specifies surveillance acceptance criteria for Type B and Type C tests. The ANSI/ANS 56.8-1994 definition is that the combined leakage rate for all penetrations subject to Type B or Type C tests is limited to less than or equal to 0.6 La, when determined on a MNPLR basis from As-found LLRT results; and limited to less than or equal to 0.6 La, as determined on a Maximum Pathway Leakage Rate (MXPLR) basis from the As-left LLRT results.

Due to the performance-based nature of Option B to Appendix J and this guideline, it is recommended that acceptance criteria for the combined As-found leakage rate for all penetrations subject to Type B or Type C testing be the same as that defined in ANSI/ANS 56.8-1994 with the following additions. The combined As-left leakage rates determined on a MXPLR basis for all penetrations shall be verified to be less than 0.6 La prior to entering a mode where containment integrity is required following an outage or shutdown that included Type B and Type C testing only. The combined As-found leakage rates determined on a MNPLR basis for all penetrations shall be less than 0.6 a at all times when containment integrity is required.

TS 5.5.6 indicates that the leakage rate testing program will be in accordance with Regulatory Guide (RG) 1.163 which endorsed NEI 94-01. Determining the as-found minimum pathway combined Type B and C test totals verifies that the requirement had been met at all times when

JAFP-16-0170 Attachment Response to Request for Additional Information Page 2 of 12 containment integrity was required. As left values are a permissive and can always be achieved through more maintenance while as-found values demonstrate acceptable performance is being maintained throughout the plant operating cycles.

Table 3.4.5-1 shows a row for the outage totals as Fraction of La and that for 2012 RO20 as found minimum pathway that fraction was 0.9315. There was no explicit indication in the LAR that this represented a failure to meet the surveillance performance criterion or that past operability of the primary containment had been evaluated given how close the Type B and C total was to the overall primary containment performance criterion of La and that the combined Type B and C leakage did not necessarily account for all containment leakage potential.

a. Was there a failure to meet the surveillance acceptance criterion (0.6 La) for as-found Type B and C total in 2012?

Response

a. No. Technical Specification (TS) 5.5.6, Primary Containment Leakage Rate Testing Program, describes the leakage rate acceptance criteria as follows:

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 La for the Type B and Type C tests, and 0.75 La for the Type A tests.

NEI 94-01 (all revisions), Section 10.2, Type B and Type C Testing Frequencies and ANSI/ANS 56.8 (1994 and 2002 revisions), Section 6.4.4, Acceptance Criteria require that:

The combined leakage rate for all penetrations subject to Type B and Type C tests shall be less than 0.60 La when evaluated on a minimum pathway (MNPLR) basis from the as-found local leak rate test results, at all times when containment operability is required.

The combined leakage rate for all penetrations subject to Type B and Type C tests shall be less than 0.60 La as determined on a maximum pathway (MXPLR) basis from the as-left local leak rate test results. This criterion is only required to be met prior to entering a mode where containment integrity is required following a refueling outage or following a shutdown that included Type B or Type C testing.

The Containment Leakage Rate testing Program incorporates these requirements in order to minimize the effects of possible post-loss of coolant accident (LOCA) releases.

The MNPLR summation total is not required to meet a TS limit or associated limiting condition for operation (LCO) action statement unless it results in exceeding the value of 1.0 La. However, whenever the MNPLR summation total exceeds 0.60 La, the Containment Leakage Rate Testing Program Manager will initiate a Corrective Action Program document to address the reduction in available margin to the value of 1.0 La.

Every reasonable effort will be made to reduce the minimum pathway summation total to less than 0.60 La, or whatever is achievable, in a timely manner. However, there are no pre-imposed restrictions on continued plant operation while corrective action is being planned and implemented. Plant operation through and up to the beginning of the start of the next outage is permissible with the MNPLR summation total in excess of 0.60 La

JAFP-16-0170 Attachment Response to Request for Additional Information Page 3 of 12 provided the overall leakage does not, nor is forecasted, to exceed 1.0 La as discussed in Technical Specification 5.5.6.

If the as-found MNPLR summation total at the beginning of a refueling outage exceeds 0.60 La, probable cause determination and corrective actions should be addressed in the Corrective Action Program.

The as-left MXPLR summation total is acceptable per the TS if it does not exceed 0.60 La at the end of an outage during which Type B and/or Type C tests were performed.

The MXPLR summation total must meet this acceptance criterion prior to entry into a mode requiring containment integrity following the end of the outage.

b. Was the possibility for overall containment leakage potential to have exceed the La criterion evaluated? If not, at what value of as-found minimum pathway combined Type B and C leakage would past containment operability be evaluated?

Response

b. Yes, CR-JAF-2012-07378 was initiated in response to R20 ST-39B, "Type B and C LLRT of Containment Penetrations," does not meet Level 1 Acceptance Criteria per Step 10.1.1. The total As-Found Min Pathway leakage of 298.104 SLM is above the allowable Level 1 limit of 192 SLM.

The basis for operability was that TS only refers to the acceptance criteria of type B and type C to be less than 0.6 La (192 SLM) prior to start-up, stating "primary containment leakage rate acceptance criteria is less than or equal to 1.0 La. During plant startup following testing in accordance with this program, the leakage rate acceptance criteria are less than or equal to 0.60 La for Type B and Type C tests." While Appendix J program supporting documents, ANSI 56.8 and NEI 94-01, do suggest acceptance criteria levels, there is no specific commitment made in the program plan.

c. Justify why the requested test interval extensions should be considered prudent given the 2012 RO20 and 2014 RO21 combined Type B and C minimum pathway test results of 0.9315 La and 0.503 La respectively?

Response

c. The associated JAF "License Amendment Request to Revise Technical Specifications Section 5.5.6 for Extension of Type A and Type C Leak Rate Test Frequencies," was submitted to:

Increase in the existing integrated leak rate test (ILRT) program test interval to 15 years.

Allow an extension from the 60-month frequency currently permitted by Option B to a 75-month frequency for Type C leakage rate testing of selected components.

This will be accomplished by revising JAF TS 5.5.6 by replacing the reference to Regulatory Guide (RG) 1.163 (Reference 1) with a reference to Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A, dated July 2012 and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, as the

JAFP-16-0170 Attachment Response to Request for Additional Information Page 4 of 12 implementation documents used by JAF to implement the JAF performance-based leakage testing program in accordance with Option B of 10 CFR Part 50, Appendix J.

NEI 94-01 R3-A, discusses the performance factors that Licensees must consider in determining test intervals. NEI 94-01 R3-A, Section 11.3 states the following regarding the acceptability of extended surveillance frequencies:

A licensee may adopt specific surveillance frequencies from Section 10.0 provided that plant-specific test performance history is acceptable as discussed in Section 10.0, and certain performance factors and controls are reviewed and applied as appropriate in the determination of test intervals. Each licensee should demonstrate by quantitative or qualitative review that plant-specific performance is adequate to support the extended test interval.

Prior to determining and implementing extended test intervals for Type B and Type C components, an assessment of the plants containment penetration and valve performance should be performed and documented. The following are some factors that have been identified as important and should be considered in establishing testing intervals:

Performance Factors Past Performance Service Design Safety Impact Cause Determination Programmatic Controls As-found Tests Schedule Review When routinely scheduling any LLRT interval beyond 60-months, the primary containment leakage rate testing program trending or monitoring (Section 12.1) shall include an estimate of the amount of understatement in the minimum pathway Type B & C summation. The estimate must be included in the post-outage report of Section 12.1 and include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

NEI 94-01 R3-A, Section 12.1 states the following regarding programmatic reporting requirements:

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

JAFP-16-0170 Attachment Response to Request for Additional Information Page 5 of 12 the report and a corrective action plan developed to restore the margin to an acceptable level.

In the NRC Safety Evaluation Report (SER) enclosed within NEI 94-01, R3-A, the Staff found that the guidance in NEI 94-01, Revision 3, was acceptable for referencing by licensees in the implementation for the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J. However, the NRC staff identified two conditions on the use of NEI 94-01, Revision 3. The JAF response to the Limitations and Conditions identified in Section 4.0 of the NEI 94-01, R3-A, NRC SER, were provided in Section 3.7.2, Limitations and Conditions Applicable to NEI 94-01 Revision 3-A, of the LAR.

JAF is not requesting the approval of test interval extensions for Type C components.

However, we are requesting approval to amend TS 5.5.6 to incorporate NEI 94-01 R3-A and the associated Limitations and Conditions. Upon receipt of the requested TS amendment, JAF will initially be required to address and update the following reporting requirement prior to placing any components on an extended interval greater than 60 months:

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

The placement of Type C components on an extended interval greater than 60 months will not occur until all of the associated requirements of NEI 94-01 R3-A and the limitations and conditions as discussed in section 3.7.2 of the LAR have been satisfied, including the restoration of Type B and Type C margin to the TS limitation as committed to in LAR Section 3.7.2.

SBPB-RAI-2 The LAR page 66 indicates that an adjustment will be applied for those Type C tested components on a 75-month interval. The NEI 94-01 Revision 3 Condition 2 indicated that an estimate of the test total understatement be made for LLRT valve testing intervals beyond 60-months.

a. When the LAR refers to Type C tested components on a 75-month interval, does this include all Type C tested components scheduled for intervals exceeding 60-months, which would include those that may be scheduled for intervals between 60-months and 75-months?

JAFP-16-0170 Attachment Response to Request for Additional Information Page 6 of 12

Response

The following is a clarification to the "Response to Condition 2, Issue 1" that addresses the above concern.

The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25% in the LLRT periodicity. As such, JAF will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate for each Type C component currently on greater than a 60-month test interval up to the 75-month extended test interval. This will result in a combined conservative Type C total for all 75-month LLRTs being "carried forward" and will be included whenever the total leakage summation is required to be updated (either while on-line or following an outage).

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

APLA-RAl-1 Section 4.2.7 of Electric Power Research Institute (EPRI) Topical Report (TR)-1009325, Rev. 2-A, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325" states that "[w]here possible, the analysis should include a quantitative assessment of the contribution of external events (for example, fire and seismic) in the risk impact assessment for extended [Integrated Leak Rate Test] ILRT intervals. For example, where a licensee possesses a quantitative fire analysis and that analysis is of sufficient quality and detail to assess the impact, the methods used to obtain the impact from internal events should be applied for the external event."

The "Fire Analysis" section of Attachment 1 to the LAR states that "although the JAF Individual Plant Examination of External Events (IPEEE) fire risk model has not been updated since its original issuance, use of the IPEEE model would tend to give conservative results."

A fire PRA model with sufficient quality and detail that has undergone a peer review under American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) RA-SA-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications", as clarified/qualified by Rev. 2 of Regulatory Guide 1.200 (RG 1.200), "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," is not utilized for this application. The IPEEE model was issued 20 years ago in June 1996, and there have been significant changes to Fire PRA methodology since then.

a. Identify any gaps between the FitzPatrick IPEEE fire risk model and the "Internal Fire Technical Elements" required by Rev. 2 of RG 1.200 that are relevant to this submittal

JAFP-16-0170 Attachment Response to Request for Additional Information Page 7 of 12 and explain why these gaps do not have significant impact on the risk assessment results for this application.

OR Demonstrate that addressing the requirements would have no impact on the final results in the LAR by providing a sensitivity evaluation to show that the use of the IPEEE results remains conservative-replace the IPEEE results, including at least the following:

i.

Reanalysis of all containment bypass scenarios leading to LERF from the internal events PRA, replacing any random failure probabilities with appropriate fire-induced values (e.g., NUREG/CR-7150 values for MOVs for ISLOCAs) - this should include any ISLOCA pathways previously screened out due to low random failure probability, as the probability could be significantly higher if fire induced.

ii.

Reanalysis of the IPEEE fire LERF using updated fire frequencies and, if suppression was credited, updated non-suppression probabilities from NUREG-2169.

The "Seismic Analysis" section of Attachment 1 to the LAR states that "In the IPEEE, JAF

[FitzPatrick] performed a Seismic Margin Assessment (SMA), which is a deterministic and conservative evaluation that does not calculate risk on a probabilistic basis." This section does not cite the use of the results from Generic Issue (Gl)-199, "Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment." It implies that there was no seismic quantification or sensitivity analysis performed for the application. However, Section 4.0 of Attachment 4 to the LAR indicates that a sensitivity analysis was performed based on Gl-199.

b. Provide additional details in the "Seismic Analysis" section of the LAR to clarify that an estimate from Gl-199 based on the United States Geological Survey (USGS) seismic hazard curves was used for quantification (bounding CDF = 6.1 E-6/y from "Weakest Link Model" stated in Gl-199) performed in Attachment 4 of the LAR.

Response

See Attachment 2 of Enclosure 1.

APLA RAI-2 Section 1.0 of Attachment 4 to the LAR states that "the purpose of this analysis is to provide a risk assessment of permanently extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) to fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for the James A FitzPatrick Nuclear Power Plant (JAF). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1]."

Section 2.0 of Attachment 4 to the LAR states that "the basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, "Performance-Based Containment Leak Test Program," September 1995, provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessment of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. This section does not contain the basis for extending the Type A Primary Containment ILRT interval to 15 years.

JAFP-16-0170 Attachment Response to Request for Additional Information Page 8 of 12 Provide the basis for extending the ILRT to a 15-year test interval per NEI 94-01, Rev. 3-A.

Response

See Attachment 2 of Enclosure 1.

APLA RAl-3 Section 5.1.2, Table 5-1 and Table 5-2 of Attachment 4 to the LAR contain the summary of the Internal Events core damage frequency (CDF) and large early release frequency (LERF) results, respectively, for the FitzPatrick PRA Model when quantified using the Level 2 model top gates. For the loss-of-coolant accident "LOCA Outside Containment" internal event, the LERF (1.09E-08) is lower than the CDF (1.78E-08). If there is no release mitigation credited for that internal event, the CDF and LERF values would be expected to be the same.

What was credited in the Probabilistic Risk Assessment (PRA) model that contributed to a lower LERF value for the "LOCA Outside Containment" internal event? Verify the credits taken reflect the "as-designed, as-built, as-operated" plant configurations per RG 1200, Rev. 2.

Response

See Attachment 2 of Enclosure 1.

APLA RAl-4 Section 4.2.2 of EPRI TR-1009325, Rev. 2-A states that "The most relevant plant-specific information should be used to develop population dose information. The order of preference shall be plant-specific best estimate, Severe Accident Mitigation Alternative (SAMA) for license renewal, and scaling of a reference plant population dose."

Section 5.2.2 of Attachment 4 to the LAR states that the population doses were calculated using the methodology of scaling Peach Bottom population doses to JAF and the scaling produced more conservative population dose values. JAF's plant-specific data is not used in calculating the population doses, while more recent plant-specific data exists.

Provide justification of (1) why plant-specific data is not used and (2) why using Peach Bottom's data would produce a more conservative value. Is the data of the two sites sufficiently comparable to render this scaling approach based on power level, leakage, and population conservative?

Response

See Attachment 2 of Enclosure 1.

APLA RAl-5 In Section 5.2.2 of Attachment 4 to the LAR, Table 5-9 lists the population dose for Class 7a, "Severe accident phenomena induced failure (early)" is "1.28E+06 person-rem," while the population dose for Class 7b, "Severe accident phenomena induced failure (late)" is "6.81 E+05 person rem."

When used in conjunction with the release class frequencies for Class 7a and Class 7b in Table 5-10, the annual population dose resulting from late containment failure (1.01 person-rem per year) is higher than that resulting from early containment failure (0.687 person-rem per year).

In NUREG-1437, "Generic Environment Impact Statement for License Renewal of Nuclear Plants",

Table 5-4 and Table G-2 show that the annual population doses due to early containment failure and late containment failure are 0.87 person-rem per year and 0.76 person rem per year, respectively.

JAFP-16-0170 Attachment Response to Request for Additional Information Page 9 of 12 That is, the value for early containment failure exceeds that for late containment failure-a trend opposite to the one in the LAR.

Explain the differences between the NUREG-1437 and the LAR results; provide references for supporting assumptions and calculations.

Response

See Attachment 2 of Enclosure 1.

ALRA RAl-6 In Section 5.3.1.of Attachment 4 to the LAR, the fire LERF result was calculated based on the fire CDF (2.12E-5) taken from the original FitzPatrick IPEEE, "James A FitzPatrick Nuclear Power Plant Individual Plant Examination of External Events", which was published in June, 1996. However, an updated fire CDF (3.42E-5) that is higher than the one listed in the original IPEEE is available in Section 2.1. 7 of "James A FitzPatrick Nuclear Power Plant - Review of FitzPatrick Individual Plant Examination of External Events (IPEEE) Submittal (TAC NO. M83622)" dated September 21, 2000 ADAMS Accession No. ML003743771). Using a smaller fire CDF value from the original IPEEE to obtain the fire LERF produced a less conservative fire LERF value in the LAR.

Subsequent to resolving the question in RAI 1.a above regarding the conservatism when using the FitzPatrick Fire IPEEE risk, confirm that the total change in LERF is still within the "very small" and "small" risk increase regions of Regulatory Guide 1.17 4. That is, perform the calculations in Section 5.3 using the updated fire CDF value and correct all subsequent calculations in the LAR that are affected by using the non-conservative old fire CDF value (2.12E-5).

Response

See Attachment 2 of Enclosure 1.

APLA RAl-7 In Table 5-21, Section 5.3.2 of Attachment 4 to the LAR, the dose rate for dose rate increase from 3-per-10 years to 1-per-10 years is 3.76E-03; the dose rate for dose rate increase from 3-per-10 years to 1-per-15 years is 6.45E-03; the dose rate for dose rate increase from 1-per-10 years to 1-per-15 years is 2.69E-03. However, based on Table 5-13, these three increases should be 5.08E-3, 8. 71 E-3 and 3.63E-3, respectively, which are higher than those listed in Table 5-21.

Correct the dose rates in Table 5-21 and all affected subsequent calculations in the LAR based on the correct dose rates listed in Table 5-13 to ensure the affected calculations do not change the final results of this LAR.

Response

See Attachment 2 of Enclosure 1.

APLA RAl-8 Section 5.3.3 of Attachment 4 to the LAR states that "Taking the baseline analysis and using the values provided in Tables 5-10 and 5-10 for the expert elicitation sensitivity yields the results in Table 5-23."

Confirm that the 2nd "5-10" in the above sentence is a typo and it should be replaced by "5-22".

JAFP-16-0170 Attachment Response to Request for Additional Information Page 10 of 12

Response

See Attachment 2 of Enclosure 1.

ALRA RAl-9 In Table 6-1, Section 6.0 of Attachment 4 to the LAR lists the dose values for Class 1, 3a and 3b are 9.15E+02, 9.15E+03 and 9.15E+04, respectively. In Table 5-10, Section 5.2.3 of Attachment 4 to the LAR, the dose values for Class 1, 3a and 3b are 2.97E+03, 2.97E+04 and 2.97E+05,respectively.

The values in Table 6-1 and Table 5-10 are inconsistent. Justify the lower dose values used in Table 6-1, Section 6.0 of Attachment 4 to the LAR; or if the dose values in Table 5-10 are correct, replace Table 6-10 dose values with the values from Table 5-10 and recalculate all subsequent results as necessary.

Response

See Attachment 2 of Enclosure 1.

ALRA RAl-10 Address the following in the F&Os section in Table A-1 of Attachment 1, "PRA Technical Adequacy" to the LAR:

a. Finding (Page 43 of 53 in the LAR, Attachment 4): "Appendix H4 states that when applying the dependency model, the dependency was generally applied to the Human Error Probability (HEP) with the higher HEP (versus assigning the dependency to the action which occurs later in time). This does not conform to the established calculation method. In general, it is more appropriate to choose the event that would occur first in the accident sequence... "

To ensure the final CDF and LERF results remain conservative and are not affected by some inadequately low joint HEP values, JAF is expected to follow the requirement of HR-G7 listed in Table 2-2.5-8(g) in ASME/ANS RA-Sa-2009 to assess the HEP dependencies in the cutsets that contain multiple HEPs. Although NRC has not endorsed any standards regarding the floor value of a joint HEP, Section 2.4.3.5 of NUREG-1792, "Good Practices for Implementing Human Reliability Analysis (HRA)" states that "The total combined probability of all the HFEs in the same accident sequence/cut set should not be less than a justified value. It is suggested that the value not be below -0.00001 since it is typically hard to defend that other dependent failure modes that are not usually treated (e.g., random events such as even a heart attack) cannot occur. Depending on the independent HFE values, the combined probability may need to be higher." EPRI TR-1021081, "Establishing Minimum Acceptable Values for Probabilities of Human Failure Events," provides a basis for extending this lowest limit by an order of magnitude to 1 E-6, which may be characteristic only for internal events. The status of this finding is still open because "the new HRA guidance has not yet been incorporated into the JAF PRA model." The LAR states that this finding is "not significant", since "the assigned dependencies are often conservative because they are based on the overall HEP value (whereas dependency between the execution portions of the HEPs can frequently be justified as being low or zero). Re-analysis of the combined HEPs to address this Fact and Observation (F&O) resulted in a CDF increase of 1.73% and a LERF increase of 0.02%... "

JAFP-16-0170 Attachment Response to Request for Additional Information Page 11 of 12 Regarding the re-analysis of the combined HEPs, was there any joint HEP floor value applied to the cutsets that contain multiple HEPs? If so, and if the floor value is significantly less than the value stated in NUREG-1792 or EPRI TR-1021081, provide justification.

If there is no floor value applied to those cutsets that contain multiple HEPs, provide justification and a sensitivity evaluation using a joint HEP floor value of <::1 E-6 to show that these low joint HEP values have no significant impact on the overall results stated in the LAR.

b. Finding (Page 44 of 53 in the LAR, Attachment 4): 'The JAF PRA determined a point estimate for damage and documented an analysis for parametric uncertainty. However, the state of knowledge correlation was not fully accounted for due to the manner in which the JAF basic event data base is constructed."

The SR description for this finding did not include LERF. However, the QU-A3 requirement listed in Table A-2 of RG 1.200 states that "The state-of-knowledge correlation should be accounted for all event probabilities. Left to the analyst to determine the extent of the events to be correlated. Need to also acknowledge LERF quantification."

Provide the quantitative results from the sensitivity evaluation performed at the time of the RG 1.200, Rev. 2 peer review-the results should include the effect on both CDF and LERF due to parametric/state of knowledge uncertainty.

c. Finding (Page 45 of 53 in the LAR, Attachment 4): "The level 2 model uses point estimates for success branches while the failure logic is modeled for the failure branch."

The LAR states that "no impact on the application is expected" because the "use of point estimates is reasonable for top events that include phenomena, and system related success probabilities are approximately equal to 1.0.

Provide the values of all success probabilities to show that they are sufficiently close to 1.0 and have minimal impact on the quantification.

d. Finding (Page 46 of 53 in the LAR, Attachment 4): "Spray-induced and submergence induced failures appear to have been addressed in the analysis. No documentation of a systematic assessment of the effects of jet impingement, pipe whip, humidity, temperature, etc., on structure, system or components (SSCs) could be identified. No evaluation of the specific equipment evaluated in the PRA compared to equipment considered in the design analyses, e.g., EQ lists was documented. Since PRA can credit non-safety-related equipment, relying on design basis evaluations to dismiss these dynamic effects may credit equipment that cannot withstand the effects considered in the design analysis. In addition, failure in a system containing high temperature fluid can actuate fire systems and impact additional equipment. Also, the PRA models may evaluate breaks beyond those of the design basis."

The status of this finding is open and the licensee states that "The only requirement to meet Capability Category II (CC-II) is to document that those mechanisms are not included in the scope."

However, Table A-3 in Rev. 2 of RG 1.200, IFSN-A6 requires that "For Cat II, it is not acceptable to just note that a flood-induced failure mechanism is not included in the scope of the internal flooding analysis. Some level of assessment is required." Also, "For the SSCs identified in IFSN-A5, IDENTIFY the susceptibility of each SSC in a flood area to flood-induced failure mechanisms. INCLUDE failure by submergence and spray in the identification process. ASSESS qualitatively the impact of flood-induced mechanisms

JAFP-16-0170 Attachment Response to Request for Additional Information Page 12 of 12 that are not formally addressed (e.g., using the mechanisms listed under Capability Category Ill of this requirement), by using conservative assumptions."

Provide the results from the analysis related to IFSN-A6 to show that CC-II requirements are met; or explain how meeting only CC-I requirements does not have impact on the final conclusions of this application.

Response

See Attachment 2 of Enclosure 1.

JAFP-16-0170 Evaluation of Risk Significance of Permanent ILRT Extension (69 Pages)

JENSEN HUGHES Advancing the Science of Safety James A. FitzPatrick Nuclear Power Plant:

Evaluation of Risk Significance of Permanent ILRT Extension 3301 O-CALC-01 Prepared for:

James A. FitzPatrick Nuclear Power Plant Project

Title:

Permanent ILRT Extension Revision: 2 Preparer: Justin Sattler Reviewer: Matthew Johnson Reviewer: Kelly Wright Review Method Approved by: Richard Anoba (Signed by Matt Johnson per Telecon)

Revision 2 Name and Date Design Review ~ Alternate Calculation D Page 1of69

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 2 of 69 REVISION RECORD

SUMMARY

Revision Revision Summary 0

Initial Issue 1

Revised PRA quality statement in Appendix A to include statement that the JAF PRA model is outside of the periodic update frequency and that any model changes that would affect the PRA model results used in this calculation have already been evaluated by Entergy personnel via MCR prioritization, and that revisions to the PRA data would not significantly affect the PRA model results.

2 Added RAIs and RAI responses as Attachment 2. Incorporated changes from the RAI responses.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 3 of 69 TABLE OF CONTENTS 1.0 PURPOSE...................................................................................................................... 4 2.0 SCOPE........................................................................................................................... 4

3.0 REFERENCES

............................................................................................................... 6 4.0 ASSUMPTIONS AND LIMITATIONS.............................................................................. 8 5.0 METHODOLOGY AND ANALYSIS................................................................................ 9 5.1 Inputs........................................................................................................................... 9 5.1.1 General Resources Available................................................................................ 9 5.1.2 Plant Specific Inputs............................................................................................12 5.1.3 Impact of Extension on Detection of Component Failures that Lead to Leakage (Small and Large)..............................................................................................14 5.2 Analysis......................................................................................................................15 5.2.1 Step 1 - Quantify the Baseline Risk in Terms of Frequency per Reactor Year.....16 5.2.2 Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose)................19 5.2.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Years.................................................................................................................21 5.2.4 Step 4 - Determine the Change in Risk in Terms of LERF and Dose..................23 5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability

..........................................................................................................................24 5.2.6 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage

..........................................................................................................................25 5.3 Sensitivities.................................................................................................................27 5.3.1 Potential Impact from External Events Contribution.............................................27 5.3.1.1 Calculating Class 3b Contribution without Subtracting LERF.....................29 5.3.2 Potential Impact from Steel Liner Corrosion Likelihood........................................30 5.3.3 Expert Elicitation Sensitivity.................................................................................31 5.3.4 Control Rod Drive Failure Sensitivity....................................................................33 6.0 RESULTS......................................................................................................................35

7.0 CONCLUSION

S AND RECOMMENDATIONS..............................................................36 A. - PRA Technical Adequacy.....................................................................38 B. - PRA RAI Responses............................................................................53

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 4 of 69 1.0 PURPOSE The purpose of this analysis is to provide a risk assessment of permanently extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) to fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for the James A. FitzPatrick Nuclear Power Plant (JAF). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1], the methodology used in EPRI TR-104285 [Reference 2], the NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals from November 2001 [Reference 3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 as applied to ILRT interval extensions, risk insights in support of a request for a plants licensing basis as outlined in Regulatory Guide (RG) 1.174 [Reference 4], the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval [Reference 5],

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

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

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

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

NEI 94-01 Revision 3-A supports using EPRI Report No. 1009325 Revision 2-A (EPRI 1018243), Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, for performing risk impact assessments in support of ILRT extensions [Reference 24]. The Guidance provided in Appendix H of EPRI Report No. 1009325 Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

It should be noted that containment leak-tight integrity is also verified through periodic in-service inspections conducted in accordance with the requirements of the American Society of

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 5 of 69 Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.

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

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

In addition, the total annual risk (person rem/year population dose) is examined to demonstrate the relative change in this parameter. While no acceptance guidelines for these additional figures of merit are published, examinations of NUREG-1493 and Safety Evaluation Reports (SER) for one-time interval extension (summarized in Appendix G of Reference 24) indicate a range of incremental increases in population dose that have been accepted by the NRC. The range of incremental population dose increases is from 0.01 to 0.2 person-rem/year and/or 0.002% to 0.46% of the total accident dose. The total doses for the spectrum of all accidents (NUREG-1493 [Reference 6], Figure 7-2) result in health effects that are at least two orders of magnitude less than the NRC Safety Goal Risk. Given these perspectives, a very small population dose is defined as an increase from the baseline interval (3 tests per 10 years) dose of 1.0 person-rem per year or 1% of the total baseline dose, whichever is less restrictive for the risk impact assessment of the proposed extended ILRT interval.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 6 of 69

3.0 REFERENCES

The following references were used in this calculation:

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

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

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

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

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

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

50-317, March 27, 2002.

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

7. Evaluation of Severe Accident Risks: Methodology for the Containment, Source Term, Consequence, and Risk Integration Analyses, NUREG/CR-4551, SAND86-1309, Volume 4, Revision 1, Part 1, December 1990.

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

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

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

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

12. Technical Findings and Regulatory Analysis for Generic Safety Issue II.E.4.3 Containment Integrity Check, NUREG-1273, April 1988.

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

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

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

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

17. Engineering Report No. JAF-NE-09-00001, Appendix G, Revision 0, James A.

FitzPatrick Nuclear Power Plant, Accident Sequence Quantification, August 2009.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 7 of 69

18. Engineering Report No. JAF-NE-09-00001, Appendix L, Revision 0, James A. FitzPatrick Nuclear Power Plant, Large Early Release Frequency Quantification, August 2009.

19. James A. FitzPatrick Nuclear Power Plant, License Renewal Application, 2008.

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

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

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

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

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

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

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

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

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

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

28. ML112070867, Containment Liner Corrosion Operating Experience Summary Technical Letter Report, U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research, August 2011.

29. Technical Letter Report ML112070867, Containment Liner Corrosion Operating Experience Summary, Revision 1, August 2011.

30. Engineering Report No. JAF-RPT-03-00007, Revision 0, James A. FitzPatrick Nuclear Power Plant, Risk Impact Assessment of Extending Containment Type A Test Interval, June 2003.

31. Engineering Report No. JAF-NE-09-00001, Revision 0, Appendix J, James A. FitzPatrick Nuclear Power Plant, Containment Performance Analysis, August 2009.

32. Calculation No. JAF-CALC-05-00134, Revision 0, James A. FitzPatrick Nuclear Power Plant, MACCS2 Model for JAFNPP, December 2005.

33. Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, NEI 05-04, Revision 2, November 2008.

34. Engineering Report No. JAF-RPT-05-00158, Revision 1, James A. FitzPatrick Nuclear Power Plant, Cost-Benefit Analysis of Severe Accident Mitigation Alternatives, June 2006.

35. Engineering Report No. JAF-RPT-MISC-02211, Revision 0, James A. FitzPatrick Nuclear Power Plant, Individual Plant Examination of External Events, June 1996.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 8 of 69

36. Generic Issue 199 (GI-199), ML100270582, September 2010, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment.

37. JAF Internal Events PRA Model, JAFSTG.caf and JAF2011.rr, Rev 2, 2011.

38. An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 2, March 2009.

39. NRC Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities -10 CFR 50.54(f), Supplement 4," June 28, 1991.

40. Parkinson, W. J., EPRI Fire PRA Implementation Guide, prepared by Science Applications International Corporation for Electric Power Research Institute, EPRI TR-105928, December 1995.

41. Professional Loss Control, Inc., Fire-Induced Vulnerability Evaluation (FIVE)

Methodology Plant Screening Guide, EPRI TR-100370, Electric Power Research Institute, Final Report, April 1992.

42. U.S. Nuclear Regulatory Commission, NUREG-1742 Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, Volume 1 & 2, Final Report, April 2002.

43. NRC letter to James A. FitzPatrick Nuclear Power Plant issuing Technical Specification Amendment 234 to implement the requirements of 10 CFR 50, Appendix J, Option B, dated October 4, 1996.

44. Entergy Nuclear Northeast, James A. FitzPatrick Nuclear Power Plant Updated Final Safety Analysis, Revision 4, April 2013.

45. Engineering Report No. JAF-NE-09-00001, Revision 0, Fitzpatrick Probabilistic Safety Assessment (PSA), Rev 4, August 2009.

4.0 ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculation:

The technical adequacy of the JAF PRA is consistent with the requirements of Regulatory Guide 1.200 [Reference 38] as is relevant to this ILRT interval extension, as detailed in Attachment 1.

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

It is appropriate to use the JAF internal events PRA model to effectively describe the risk change attributable to the ILRT extension. An extensive sensitivity study is done in Section 5.3.1 to show the effect of including external event models for the ILRT extension. A Seismic PRA [Reference 36] and Fire PRA from the IPEEE [Reference 35]

are used for this sensitivity analysis. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose) will not substantially differ if detailed analysis of high wind events were to be included in the calculations.

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

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

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 9 of 69

The representative containment leakage for Class 3a sequences is 10La based on the previously approved methodology performed for Indian Point Unit 3 [Reference 8, Reference 9].

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

[Reference 24].

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

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

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

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

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

1.

NUREG/CR-3539 [Reference 10]

2.

NUREG/CR-4220 [Reference 11]

3.

NUREG-1273 [Reference 12]

4.

NUREG/CR-4330 [Reference 13]

5.

EPRI TR-105189 [Reference 14]

6.

NUREG-1493 [Reference 6]

7.

EPRI TR-104285 [Reference 2]

8.

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

9.

NEI Interim Guidance [Reference 3, Reference 20]

10.

Calvert Cliffs liner corrosion analysis [Reference 5]

11.

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

This first study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and is to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different containment leakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRCs cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh study is an EPRI study of the impact of extending ILRT and local leak rate test (LLRT) intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 10 of 69 used as the basis for the consequence analysis of the ILRT interval extension for JAF. The ninth study includes the NEI recommended methodology (promulgated in two letters) for evaluating the risk associated with obtaining a one-time extension of the ILRT interval. The tenth study addresses the impact of age-related degradation of the containment liners on ILRT evaluations.

Finally, the eleventh study builds on the previous work and includes a recommended methodology and template for evaluating the risk associated with a permanent 15-year extension of the ILRT interval.

NUREG/CR-3539 [Reference 10]

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

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

NUREG/CR-4220 [Reference 11]

NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985.

The study reviewed over two thousand LERs, ILRT reports and other related records to calculate the unavailability of containment due to leakage.

NUREG-1273 [Reference 12]

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

NUREG/CR-4330 [Reference 13]

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

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

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

EPRI TR-105189 [Reference 14]

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

NUREG-1493 [Reference 6]

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

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

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 11 of 69 Given the insensitivity of risk to the containment leak rate and the small fraction of leak paths detected solely by Type A testing, increasing the interval between integrated leak rate tests is possible with minimal impact on public risk.

EPRI TR-104285 [Reference 2]

Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study),

the EPRI TR-104285 study is a quantitative evaluation of the impact of extending ILRT and LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with NUREG-1150 Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-104285 uses a simplified Containment Event Tree to subdivide representative core damage frequencies into eight classes of containment response to a core damage accident:

1.

Containment intact and isolated

2.

Containment isolation failures dependent upon the core damage accident

3.

Type A (ILRT) related containment isolation failures

4.

Type B (LLRT) related containment isolation failures

5.

Type C (LLRT) related containment isolation failures

6.

Other penetration related containment isolation failures

7.

Containment failures due to core damage accident phenomena

8.

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

Rthe proposed CLRT (Containment Leak Rate Tests) frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms. For example, for the PWR analyzed, the change is about 0.04 person-rem per yearR NUREG-1150 [Reference 15] and NUREG/CR-4551 [Reference 7]

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

NEI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals [Reference 3, Reference 20]

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

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

Calvert Cliffs Response to Request for Additional Information Concerning the License Amendment for a One-Time Integrated Leakage Rate Test Extension [Reference 5]

This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting liner corrosion, and the corresponding change in risk. The

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 12 of 69 methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms was factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete basemat, each with a steel liner.

EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals [Reference 24]

This report provides a generally applicable assessment of the risk involved in extension of ILRT test intervals to permanent 15-year intervals. Appendix H of this document provides guidance for performing plant-specific supplemental risk impact assessments and builds on the previous EPRI risk impact assessment methodology [Reference 2] and the NRC performance-based containment leakage test program [Reference 6], and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

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

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

Level 1 Model results [Reference 37]

Level 2 Model results [Reference 37, Reference 18, Reference 19, Reference 34]

Release category definitions used in the Level 2 Model [Reference 18, Reference 19, Reference 34]

Dose within a 50-mile radius [Reference 19, Reference 32]

ILRT results to demonstrate adequacy of the administrative and hardware issues

[Reference 30]

Containment failure probability data [Reference 31]

JAF Model The Internal Events PRA Model that is used for JAF is characteristic of the as-built plant. The current Level 2 model (JAF PRA Model JAFSTG, which was updated in 2011) [Reference 37] is a linked fault tree model (with supporting information in Reference 17). The total CDF is 2.40E-6/year.

For the purposes of the ILRT extension analysis, the Level 2 model is used to facilitate mapping the JAF PRA results to the EPRI Accident Classes. Thus, each Level 2 release state top gate was quantified and the resulting cutsets combined to produce a single Level 2 PRA cutset file.

When each Level 2 top gate is quantified and the cutsets combined, the total CDF is 2.63E-6/year, which is slightly larger than the single top model total CDF. This slight difference is expected when quantifying separate tops and combining cutsets when compared to a single top quantification. Using the combined Level 2 results of this analysis is conservative with respect to the ILRT extension application. Table 5-1 and Table 5-2 provide a summary of the Internal Events CDF and LERF results for the JAF PRA Model when quantified using the Level 2 model top gates. The total LERF is 2.69E-7/year.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 13 of 69 The Fire CDF from the IPEEE [Reference 35] is 2.12E-5/year. The Seismic PRA results from Generic Issue 199 yields a CDF of 6.1E-6/year [Reference 36]. Refer to Section 5.3.1 for further details on external events as they pertain to this analysis.

Table 5 Internal Events CDF Internal Events Frequency (per year)

Loss of AC Bus 6.15E-07 Loss of Offsite Power 2.35E-07 Transient 1.20E-06 Loss of DC Bus 2.58E-07 LOCA 1.62E-07 Internal Flooding 1.32E-07 Interfacing Systems LOCA 9.64E-09 LOCA Outside Containment 1.09E-08 Total Internal Events CDF 2.63E-06 Table 5 Internal Events LERF Internal Events Frequency (per year)

Loss of AC Bus 3.20E-09 Loss of Offsite Power 2.76E-08 Transient 1.94E-07 Loss of DC Bus 1.24E-08 LOCA 9.08E-09 Internal Flooding 2.91E-09 Interfacing Systems LOCA 9.64E-09 LOCA Outside Containment 1.09E-08 Total Internal Events LERF 2.69E-07 Release Category Definitions Table 5-3 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology [Reference 2]. These containment failure classifications are used in this analysis to determine the risk impact of extending the Containment Type A test interval, as described in Section 5.2 of this report.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 14 of 69 Table 5 EPRI Containment Failure Classification [Reference 2]

Class Description 1

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

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

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

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

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

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

7 Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J testing requirements do not impact these accidents.

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

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

The ILRT can detect a number of component failures such as liner breach, failure of certain bellows arrangements, and failure of some sealing surfaces, which can lead to leakage. The proposed ILRT test interval extension may influence the conditional probability of detecting these types of failures. To ensure that this effect is properly addressed, the EPRI Class 3 accident class, as defined in Table 5-3, is divided into two sub-classes, Class 3a and Class 3b, representing small and large leakage failures respectively.

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

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

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

The supplemental information states:

The methodology employed for determining LERF (Class 3b frequency) involves conservatively multiplying the CDF by the failure probability for this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, some plant-specific accident classes leading to core damage are likely to include individual sequences that either may already (independently) cause a LERF or could never cause a LERF, and are thus not associated with a

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 15 of 69 postulated large Type A containment leakage path (LERF). These contributors can be removed from Class 3b in the evaluation of LERF by multiplying the Class 3b probability by only that portion of CDF that may be impacted by Type A leakage.

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

The LERF sequences, which includes Class 8 sequences, are subtracted from the CDF that is applied to Class 3b. To be consistent, the same change is made to the Class 3a CDF, even though these events are not considered LERF. Class 2 events refer to sequences with large pre-existing containment isolation failures; Class 8 events refer to sequences with containment bypass events. These sequences are already considered to contribute to LERF in the JAF Level 2 PRA analysis.

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

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

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

5.2 Analysis The application of the approach based on the guidance contained in EPRI Report No. 1009325, Revision 2-A, Appendix H [Reference 24], EPRI TR-104285 [Reference 2] and previous risk assessment submittals on this subject [References 5, 8, 21, 22, and 23] have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report, as described in Table 5-4.

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

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

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

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

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 16 of 69

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

Table 5 EPRI Accident Class Definitions Accident Classes (Containment Release Type)

Description 1

No Containment Failure 2

Large Isolation Failures (Failure to Close) 3a Small Isolation Failures (Liner Breach) 3b Large Isolation Failures (Liner Breach) 4 Small Isolation Failures (Failure to Seal - Type B) 5 Small Isolation Failures (Failure to Seal - Type C) 6 Other Isolation Failures (e.g., Dependent Failures) 7a Failures Induced by Phenomena (Early) 7b Failures Induced by Phenomena (Late) 8 Bypass (Interfacing System LOCA)

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

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

Step 1 - Quantify the baseline risk in terms of frequency per reactor year for each of the accident classes presented in Table 5-4.

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

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

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

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

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

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

The frequencies for the severe accident classes defined in Table 5-4 were developed for JAF by first determining the frequencies for Classes 1, 2, 6, 7, and 8. Table 5-5 presents the grouping of each release category in EPRI Classes based on the associated description. Table 5-6 presents the release categories, frequency, and EPRI category for each sequence and the totals of each EPRI classification. Table 5-7 provides a summary of the accident sequence frequencies that can lead to radionuclide release to the public and have been derived consistent

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 17 of 69 with the definitions of accident classes defined in EPRI TR-104285 [Reference 2], the NEI Interim Guidance [Reference 3], and guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24]. Adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion of the steel liner per the methodology described in Section 5.2.6. Note: calculations were performed with more digits than shown in this section.

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

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

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

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

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

 =

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

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

 =
   =



 2.63E 2.69E-7 = 2.18E-8 In the database of 217 ILRTs, there are zero containment leakage events that could result in a large early release. Therefore, the Jeffreys non-informative prior is used to estimate a failure rate and is illustrated in the following equations:

Jeffreys Failure Probability = 4567 89 :;<5= + 1/2 4567 89 @=A= + 1

B = 0 + 1/2 217 + 1 = 0.0023 The frequency of a Class 3b failure is calculated by the following equation:

B =
B   =

.C

D 2.63E 2.69E-7 = 5.43E-9 For this analysis, the associated containment leakage for Class 3a is 10La and for Class 3b is 100La. These assignments are consistent with the guidance provided in Reference 24.

Class 1 Sequences. This group consists of all core damage accident progression bins for which the containment remains intact (modeled as Technical Specification Leakage). The frequency per year is initially determined from the EPRI Accident Class 1 frequency listed in Table 5-6 and then subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF), calculated below:

 =   + B

Class 2 Sequences. This group consists of core damage accident progression bins with large containment isolation failures. The frequency per year for these sequences is obtained from the

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 18 of 69 EPRI Accident Class 2 frequency listed in Table 5-6.

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

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

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

These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution. All other failure modes are bounded by the Class 2 assumptions.

This accident class is also not evaluated further.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure is induced by severe accident phenomena (e.g., overpressure). The Level 2 model defines early releases as happening before 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, intermediate releases as happening between 6 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and late releases as happening after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> [Reference 31]. The Level 3 model calculates population dose for early (less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) and late (greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) releases [Reference 32]. Therefore, all early and intermediate releases from the Level 2 quantification are grouped into the early group for multiplication with the early population dose in this calculation. This class is split into early (7a) and late (7b) failures. For this analysis, the frequency is determined from the EPRI Accident Class 7 frequencies listed in Table 5-6.

Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass occurs, which includes interfacing system LOCAs (ISLOCAs) and LOCAs outside containment (LOCAOC). For this analysis, the frequency is determined from the EPRI Accident Class 8 frequency listed in Table 5-6.

Level 2 quantification of the 2011 JAFSTG model [Reference 37] is distributed into EPRI classes based on release categories. Table 5-5 shows this distribution.

Table 5 Release Category Frequencies Release Category Description of Outcome EPRI Category Frequency (/yr)

NCF No Containment Failure 1

5.86E-07 CIS Containment Isolation Failure 2

3.31E-09 LERF (no bypass)

Large Early Release Frequency (excluding containment bypass) 7a 2.49E-07 MERF Medium Early Release Frequency 7a 3.35E-08 SERF Small Early Release Frequency 7a 3.42E-08 SSERF Small-Small Early Release Frequency 7a 0.00E+00 LIRF Large Intermediate Release Frequency 7a 0.00E+00 MIRF Medium Intermediate Release Frequency 7a 0.00E+00 SIRF Small Intermediate Release Frequency 7a 1.40E-07 SSIRF Small-Small Intermediate Release Frequency 7a 8.16E-08 LLRF Large Late Release Frequency 7b 1.20E-06 MLRF Medium Late Release Frequency 7b 6.34E-08 SLRF Small Late Release Frequency 7b 2.22E-07 SSLRF Small-Small Late Release Frequency 7b 0.00E+00 ISLOCA Interfacing Systems LOCA 8

9.64E-09 LOCAOC LOCA Outside Containment (that leads to LERF) 8 1.09E-08

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 19 of 69 Table 5 Release Category Frequencies Release Category Description of Outcome EPRI Category Frequency (/yr)

Contribution to EPRI Classification 2 3.31E-09 Contribution to EPRI Classification 7a 5.39E-07 Contribution to EPRI Classification 7b 1.49E-06 Contribution to EPRI Classification 8 2.05E-08 Table 5 Accident Class Frequencies Release Categories EPRI Category Frequency (/yr)

NCF Class 1 5.86E-07 CIS Class 2 3.31E-09 EARLY Class 7a 5.39E-07 LATE Class 7b 1.49E-06 ISLOCA, LOCAOC Class 8 2.05E-08 Total (CDF)

N/A 2.63E-06 Table 5 Baseline Risk Profile Class Description Frequency (/yr) 1 No containment failure 5.58E-072 2

Large containment isolation failures 3.31E-09 3a Small isolation failures (liner breach) 2.18E-08 3b Large isolation failures (liner breach) 5.43E-09 4

Small isolation failures - failure to seal (Type B) 1 5

Small isolation failures - failure to seal (Type C) 1 6

Containment isolation failures (dependent failure, personnel errors) 1 7a Severe accident phenomena induced failure (early) 5.39E-07 7b Severe accident phenomena induced failure (late) 1.49E-06 8

Containment bypass 2.05E-08 Total 2.63E-06

1.

represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.

2.

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

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

Plant-specific release analyses were performed to estimate the person-rem doses to the population within a 50-mile radius from the plant. Reference 19 provides the population doses for each class; however, the no containment failure dose did not consider the maximally allowable leakage of 1.0La and cannot be used in the EPRI methodology of scaling the intact dose to Classes 3a and 3b. Therefore, only the dose for Classes 7a and 7b were obtained from Reference 19, and the population dose for Classes 1, 2, and 8 are calculated using the methodology of scaling Peach Bottom population doses to JAF [Reference 7]. The adjustment factor for reactor power level (AFpower) is defined as the ratio of the power level at JAF (PLF)

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 20 of 69

[Reference 44] to that at Peach Bottom Unit 2 (PLP) [Reference 7]. This adjustment factor is calculated as follows:

AFpower = PLF / PLP = 2536 / 3293 = 0.770 The adjustment factor for technical specification (TS) allowed containment leakage is defined as the ratio of the containment leakage at FitzPatrick (LRF) to that at Peach Bottom Unit 2 (LRP).

This adjustment factor is calculated as follows:

AFleakage = LRF / LRP Since the leakage rates are in terms of the containment volume, the ratio of containment volumes is needed to relate the leakage rates. The TS maximum allowed containment leakage at JAF (TSJAF) is 1.5 volume%/day [Reference 43]; the containment free volume at JAF (VOLJAF) is 264,000 ft3 [Table 5.2-1 of Reference 44]. The TS maximum allowed containment leakage at Peach Bottom Unit 2 (TSPB) is 0.5 volume%/day [Reference 7]; the containment free volume at Peach Bottom Unit 2 (VOLPB) is 307,000 ft3 [Reference 7].Therefore, LRF = TSJAF

• VOLJAF LRP = TSPB

• VOLPB AFleakage = 1.5

• 264000 / 0.5

• 307000 = 2.58 The adjustment factor for population (AFPopulation) is defined as the ratio of the population within 50-mile radius of JAF (POPF) [Reference 19] to that of Peach Bottom Unit 2 (POPP) [Reference 7]. According to the 2000 census, the population within 50 miles of JAF was 914,668 and is projected to decline by approximately 6% over the remaining life of the power plant [Reference 19]. Therefore, the 2000 census population is used as a slightly conservative value. This adjustment factor is calculated as follows:

AFPopulation = POPF / POPP = 914668 / 3.02E+6 = 0.303 Consequences dependent on the INTACT TS Leakage (collapsed accident progression bins 8 and 10).

AFINTACT = AFpower

• AFLeakage

• AFPopulation = 0.770

• 2.58

• 0.303 = 0.602 Since the other categories are not dependent on the TS Leakage, the adjustment factor (AF) is calculated by combining the factors as follows:

AF = AFpower

• AFPopulation = 0.770

• 0.303 = 0.233 The population dose data in NUREG/CR-4551 for Peach Bottom Unit 2 [Reference 7] is reported in ten distinct collapsed accident progression bins (CAPBs). For this ILRT extension application, CAPB8 and CAPB10 are categorized in EPRI Accident Class 1; CAPB3 is categorized in EPRI Accident Class 2; and CAPB7 is categorized in EPRI Accident Class 8.

Based on the above adjustment factors and the 50-mile population dose (person-rem) for each CAPB considered in the NUREG/CR-4551 Peach Bottom Unit 2 study, the JAF population doses (JAFPD) for Classes 2 and 8 are calculated as follows:

JAFPDClass1 = AFINTACT

• PDCAPB8 + AFINTACT

• PDCAPB10 = 0.602

• 4.94E+3 + 0.602

• 0 = 2.97E+3 JAFPDClass2 = AF

• PDCAPB3 = 0.233

• 2.97E+6 = 6.93E+5 JAFPDClass8 = AF

• PDCAPB7 = 0.233

• 1.95E+6 = 4.55E+5 Table 5-8 provides a correlation of JAF population dose to EPRI Accident Class. Table 5-9 provides population dose for each EPRI accident class.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 21 of 69 The population dose for EPRI Accident Classes 3a and 3b were calculated based on the guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24] as follows:


Y <:== 3:
8Z5<:A;8[ 8= = 10 2.97E + 3 = 2.97+4


Y <:== 37
8Z5<:A;8[ 8= = 100 2.97E + 3 = 2.97+5 Table 5 Mapping of Population Dose to EPRI Accident Class Release Category EPRI Category Frequency (/yr)

Population Dose (person-rem)

INTACT Class 1 5.58E-07 2.97E+03 CIS Class 2 3.31E-09 6.93E+05 EARLY Class 7a 5.39E-07 1.28E+06 LATE Class 7b 1.49E-06 6.81E+05 ISLOCA, LOCAOC Class 8 2.05E-08 4.55E+05 Table 5 Baseline Population Doses Class Description Population Dose (person-rem) 1 No containment failure 2.97E+03 2

Large containment isolation failures 6.93E+05 3a Small isolation failures (liner breach) 2.97E+041 3b Large isolation failures (liner breach) 2.97E+052 4

Small isolation failures - failure to seal (type B)

N/A3 5

Small isolation failures - failure to seal (type C)

N/A3 6

Containment isolation failures (dependent failure, personnel errors)

N/A3 7a Severe accident phenomena induced failure (early) 1.28E+06 7b Severe accident phenomena induced failure (late) 6.81E+05 8

Containment bypass 4.55E+05

1.

10*La

2.

100*La

3.

The reason for the N/A is explained in Section 5.2.1.

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

Risk Impact Due to 10-Year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and Class 3b sequences is impacted. The risk contribution is

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 22 of 69 changed based on the NEI guidance as described in Section 5.1.3 by a factor of 10/3 compared to the base case values. The Class 3a and 3b frequencies are calculated as follows:

\\]^_ =

]





   =

]





 2.37E-6 = 7.27E-8

\\B]^_ =

]



.C

D   =

]



.C

D 2.37E-6 = 1.81E-8 The results of the calculation for a 10-year interval are presented in Table 5-10.

1.

represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.

2.

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

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

\\C^_ =

C





   = 5



 2.37E-6 = 1.09E-7

\\BC^_ =

C



.C

D   = 5

.C

D 2.37E-6 = 2.71E-8 The results of the calculation for a 15-year interval are presented in Table 5-11.

Table 5 Risk Profile for Once in 10 Year ILRT Class Description Frequency

(/yr)

Contribution

(%)

Population Dose (person-rem)

Population Dose Rate (person-rem/yr) 1 No containment failure2 4.95E-07 18.78%

2.97E+03 1.47E-03 2

Large containment isolation failures 3.31E-09 0.13%

6.93E+05 2.29E-03 3a Small isolation failures (liner breach) 7.27E-08 2.76%

2.97E+04 2.16E-03 3b Large isolation failures (liner breach) 1.81E-08 0.69%

2.97E+05 5.37E-03 4

Small isolation failures - failure to seal (type B) 1 1

1 1

5 Small isolation failures - failure to seal (type C) 1 1

1 1

6 Containment isolation failures (dependent failure, personnel errors) 1 1

1 1

7a Severe accident phenomena induced failure (early) 5.39E-07 20.44%

1.28E+06 6.87E-01 7b Severe accident phenomena induced failure (late) 1.49E-06 56.43%

6.81E+05 1.01E+00 8

Containment bypass 2.05E-08 0.78%

4.55E+05 9.33E-03 Total 2.63E-06 1.72E+00

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 23 of 69 Table 5 Risk Profile for Once in 15 Year ILRT Class Description Frequency

(/yr)

Contribution

(%)

Population Dose (person-rem)

Population Dose Rate (person-rem/yr) 1 No containment failure2 4.50E-07 17.06%

2.97E+03 1.34E-03 2

Large containment isolation failures 3.31E-09 0.13%

6.93E+05 2.29E-03 3a Small isolation failures (liner breach) 1.09E-07 4.14%

2.97E+04 3.24E-03 3b Large isolation failures (liner breach) 2.71E-08 1.03%

2.97E+05 8.06E-03 4

Small isolation failures - failure to seal (type B) 1 1

1 1

5 Small isolation failures - failure to seal (type C) 1 1

1 1

6 Containment isolation failures (dependent failure, personnel errors) 1 1

1 1

7a Severe accident phenomena induced failure (early) 5.39E-07 20.44%

1.28E+06 6.87E-01 7b Severe accident phenomena induced failure (late) 1.49E-06 56.43%

6.81E+05 1.01E+00 8

Containment bypass 2.05E-08 0.78%

4.55E+05 9.33E-03 Total 2.63E-06 1.72E+00

1.

represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.

2.

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

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

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

For JAF, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a 3-in-10-year test interval from Table 5-8, the Class 3b frequency is 5.43E-9; based on a 10-year test interval from Table 5-10, the Class 3b frequency is 1.81E-8/year; based on a 15-year test interval from Table 5-11, the Class 3b frequency is 2.71E-8/year. Thus, the increase in LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is 2.17E-8/year.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 24 of 69 Similarly, the increase due to increasing the interval from 10 to 15 years is 9.04E-9/year. As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology),

the estimated change in LERF is less than the threshold criteria for a very small change when comparing the 15-year results to the current 10-year requirement and the original 3-year requirement. Table 5-12 summarizes these results.

Table 5 Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Years 15 Years Class 3b (Type A LERF) 5.43E-09 1.81E-08 2.71E-08 LERF (3 year baseline) 1.27E-08 2.17E-08 LERF (10 year baseline) 9.04E-09 EPRI Report No. 1009325, Revision 2-A [Reference 24] states that a very small population dose is defined as an increase of 1.0 person-rem per year, or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. As shown in Table 5-13, the results of this calculation meet these criteria.

Table 5 Impact on Dose Rate due to Extended Type A Testing Intervals ILRT Inspection Interval 10 Years 15 Years Dose Rate (3 year baseline) 5.08E-03 8.71E-03 Dose Rate (10 year baseline) 3.63E-03

%Dose Rate (3 year baseline) 0.296%

0.508%

%Dose Rate (10 year baseline) 0.211%

5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability Another parameter that the NRC guidance in RG 1.174 [Reference 4] states can provide input into the decision-making process is the change in the conditional containment failure probability (CCFP). The CCFP is defined as the probability of containment failure given the occurrence of an accident. This probability can be expressed using the following equation:


= 1 9[`9



where f(ncf) is the frequency of those sequences that do not result in containment failure; this frequency is determined by summing the Class 1 and Class 3a results.

Table 5-14 shows the steps and results of this calculation. The difference in CCFP between the 3-year test interval and 15-year test interval is 0.824%.

Table 5 Impact on CCFP due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Years 15 Years f(ncf) (/yr) 5.80E-07 5.68E-07 5.59E-07 f(ncf)/CDF 0.220 0.215 0.212 CCFP 0.780 0.785 0.788 CCFP (3 year baseline) 0.480%

0.824%

CCFP (10 year baseline) 0.343%

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 25 of 69 As stated in Section 2.0, a change in the CCFP of up to 1.5% is assumed to be small. The increase in the CCFP from the 3 in 10 year interval to 1 in 15 year interval is 0.824%. Therefore, this increase is judged to be very small.

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

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

Differences between the containment basemat and the containment cylinder and dome

The historical steel liner flaw likelihood due to concealed corrosion

The impact of aging

The corrosion leakage dependency on containment pressure

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

Based on a review of industry events, an Oyster Creek incident is assumed to be applicable to JAF for a concealed shell failure in the floor. In the Calvert Cliffs analysis, this event was assumed not to be applicable and a half failure was assumed for basemat concealed liner corrosion due to the lack of identified failures (See Table 5-15, Step 1).

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

Consistent with the Calvert Cliffs analysis, the estimated historical flaw probability is also limited to 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date (and have been performed since the time frame of the Calvert Cliffs analysis)

(See Table 5-15, Step 1).

Consistent with the Calvert Cliffs analysis, the steel liner flaw likelihood is assumed to double every five years. This is based solely on judgment and is included in this analysis to address the increased likelihood of corrosion as the steel liner ages (See Table 5-15, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.

In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere, given that a liner flaw exists, was estimated as 1.1% for the cylinder and dome, and 0.11% (10% of the cylinder failure probability) for the basemat. These values were determined from an assessment of the probability versus containment pressure. For JAF, the ultimate pressure is 137 psig [Reference 31]. Probabilities of 1%

for the cylinder and dome, and 0.1% for the basemat are used in this analysis, and sensitivity studies are included in Section 5.3.2 (See Table 5-15, Step 4).

Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the basemat region is considered to be less likely than the containment cylinder and dome region (See Table 5-15, Step 4).

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 26 of 69

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

To date, all liner corrosion events have been detected through visual inspection (See Table 5-15, Step 5).

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

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

Containment Basemat (15%)

1 Historical liner flaw likelihood Failure data: containment location specific Success data: based on 70 steel-lined containments and 5.5 years since the 10CFR 50.55a requirements of periodic visual inspections of containment surfaces Events: 2 (Brunswick 2 and North Anna 2) 2 / (70 x 5.5) = 5.19E-03 Events: 1 (Oyster Creek) 1 / (70 x 5.5) = 2.60E-03 2

Aged adjusted liner flaw likelihood During the 15-year interval, assume failure rate doubles every five years (14.9% increase per year). The average for the 5th to 10th year set to the historical failure rate.

Year 1

average 5-10 15 Failure rate 2.05E-03 5.19E-03 1.43E-02 Year 1

average 5-10 15 Failure rate 1.03E-03 2.60E-03 7.14E-03 15 year average = 6.44E-03 15 year average = 3.22E-03 3

Increase in flaw likelihood between 3 and 15 years Uses aged adjusted liner flaw likelihood (Step 2),

assuming failure rate doubles every five years.

0.73% (1 to 3 years) 4.18% (1 to 10 years) 9.66% (1 to 15 years) 0.36% (1 to 3 years) 2.08% (1 to 10 years) 4.82% (1 to 15 years) 4 Likelihood of breach in containment given liner flaw 1%

0.1%

5 Visual inspection detection failure likelihood 10%

5% failure to identify visual flaws plus 5% likelihood that the flaw is not visible (not through-cylinder but could be detected by ILRT).

All events have been detected through visual inspection. 5%

visible failure detection is a conservative assumption.

100%

Cannot be visually inspected 6

Likelihood of non-detected containment leakage (Steps 3 x 4 x

5) 0.00073% (3 years) 0.73% x 1% x 10%

0.00418% (10 years) 4.18% x 1% x 10%

0.00966% (15 years) 9.66% x 1% x 10%

0.000360% (3 years) 0.36% x 0.1% x 100%

0.00208% (10 years) 2.08% x 0.1% x 100%

0.00482% (15 years) 4.82% x 0.1% x 100%

The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the containment cylinder and dome, and the containment basemat, as summarized below for JAF.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 27 of 69 Table 5 Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for JAF Description At 3 years: 0.00073% + 0.000360% = 0.00109%

At 10 years: 0.00418% + 0.00208% = 0.00626%

At 15 years: 0.00966% + 0.00482% = 0.01448%

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

The two corrosion events that were initiated from the non-visible (backside) portion of the containment liner used to estimate the liner flaw probability in the Calvert Cliffs analysis are assumed to be applicable to this containment analysis. These events, one at North Anna Unit 2 (September 1999) caused by timber embedded in the concrete immediately behind the containment liner, and one at Brunswick Unit 2 (April 1999) caused by a cloth work glove embedded in the concrete next to the liner, were initiated from the nonvisible (backside) portion of the containment liner. A search of the NRC website LER database identified two additional events have occurred since the Calvert Cliffs analysis was performed. In January 2000, a 3/16-inch circular through-liner hole was found at Cook Nuclear Plant Unit 2 caused by a wooden brush handle embedded immediately behind the containment liner. The other event occurred in April 2009, where a through-liner hole approximately 3/8-inch by 1-inch in size was identified in the Beaver Valley Power Station Unit 1 (BVPS-1) containment liner caused by pitting originating from the concrete side due to a piece of wood that was left behind during the original construction that came in contact with the steel liner. Two other containment liner through-wall hole events occurred at Turkey Point Units 3 and 4 in October 2010 and November 2006, respectively. However, these events originated from the visible side caused by the failure of the coating system, which was not designed for periodic immersion service, and are not considered to be applicable to this analysis. More recently, in October 2013, some through-wall containment liner holes were identified at BVPS-1, with a combined total area of approximately 0.395 square inches. The cause of these through-wall liner holes was attributed to corrosion originating from the outside concrete surface due to the presence of rayon fiber foreign material that was left behind during the original construction and was contacting the steel liner

[References 28 and 29]. For risk evaluation purposes, these five total corrosion events occurring in 66 operating plants with steel containment liners over a 17.1 year period from September 1996 to October 4, 2013 (i.e., 5/(66*17.1) = 4.43E-03) are bounded by the estimated historical flaw probability based on the two events in the 5.5 year period of the Calvert Cliffs analysis (i.e.,

2/(70*5.5) = 5.19E-03) incorporated in the EPRI guidance.

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

The IPEEE Fire PRA calculated a CDF of 2.12E-5 [Reference 35]. Per APLA RAI-1, this Fire PRA has been conservatively updated to incorporate containment bypass scenarios and update fire ignition frequencies and non-suppression probabilities based on updated industry data. See for more details on this calculation. The updated estimate of Fire CDF and LERF are 2.57E-5 and 6.98E-6, respectively.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 28 of 69 Note that because the re-analyzed fire LERF is likely conservative, LERF is not subtracted from the existing CDF. The following shows the calculation for Class 3b:

B =
B  = 0.5 218 2.57-5 = 5.89-8

B]^_ =

]


B  =

]



].C

D 2.57-5 = 1.96E-7

aBC^_ =

C


B  = 5

].C

D 2.57-5 = 2.95E-7 The IPEEE Seismic Margin Assessment (SMA) does not result in an estimate of CDF

[Reference 35]. The conclusions reached in 2010 by GI-199 [Reference 36] are used for estimating Seismic CDF at plants in the Central and Eastern United States, which includes JAF.

The most conservative Seismic CDF reported for JAF in Table D-1 of Reference 36 is 6.1E-6, which is the weakest link model [Reference 36]. Applying the internal event LERF/CDF ratio to the Seismic CDF yields an estimated seismic LERF of 6.24E-7, as shown by the equation below.

LERFSeismic CDFSeismic

• LERFIE / CDFIE = 6.1E-6

• 2.69E-7 / 2.63E-6 = 6.24E-7 Again subtracting LERF from CDF, the Class 3b frequency can be calculated by the following formulas:

B =
B   =

].C

D 6.1E 6.24E-7 = 1.26E-8

B]^_ =

]


B   =

]



].C

D 6.1E 6.24E-7 = 4.19E-8

BC^_ =

C


B   =

C



].C

D 6.1E 6.24E-7 = 6.28E-8 The IPEEE [Section 1.4.3 of Reference 35] states, 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, for this sensitivity a conservative CDF of 10-6 is used for all other external events. Applying the internal event LERF/CDF ratio to the other external events CDF yields an estimated high wind LERF of 1.02E-7. Again subtracting LERF from CDF, the Class 3b frequency can be calculated by the following formulas:

B =
B   =

].C

D 1.0E 1.02E-7 = 2.06E-9

B]^_ =

]


B   =

]



].C

D 1.0E 1.02E-7 = 6.86E-9

BC^_ =

C


B   =

C



].C

D 1.0E 1.02E-7 = 1.03E-8 The fire, seismic, and other external events contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The change in LERF is calculated for the 1 in 10 year and 1 in 15 year cases and the change defined for the external events in Table 5-17.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 29 of 69 Table 5 JAF External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 per 15 years) 3 per 10 year 1 per 10 year 1 per 15 years External Events 7.36E-08 2.45E-07 3.68E-07 2.94E-07 Internal Events 5.43E-09 1.81E-08 2.71E-08 2.17E-08 Combined 7.90E-08 2.63E-07 3.95E-07 3.16E-07 The internal event results are also provided to allow a composite value to be defined. When both the internal and external event contributions are combined, the total change in LERF of 3.16E-7 meets the guidance for small change in risk, as it exceeds 1.0E-7/yr and remains less than 1.0E-6 change in LERF. For this change in LERF to be acceptable, total LERF must be less than 1.0E-5. The total LERF value is calculated below:

LERF = LERFinternal + LERFfire + LERFseismic + LERFother + LERFclass3Bincrease

= 2.69E-7/yr + 6.98E-6/yr + 6.24E-7/yr + 1.02E-7/yr + 3.16E-7/yr = 8.29E-6/yr As specified in Regulatory Guide 1.174 [Reference 4], since the total LERF is less than 1.0E-5, it is acceptable for the LERF to be between 1.0E-7 and 1.0E-6.

5.3.1.1 Calculating Class 3b Contribution without Subtracting LERF The Finding against LE-E3 indicates the LERF definition may cause some sequences to be misclassified, and therefore, LERF to be miscalculated (see Table A-1 in Attachment A). A sensitivity is done where LERF is not subtracted from CDF before multiplying by the probability of a Class 3b leak. The results are shown in Table 5-18.

Table 5 JAF Revised Methodology for ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 per 15 years) 3 per 10 year 1 per 10 year 1 per 15 years External Events 7.52E-08 2.51E-07 3.76E-07 3.01E-07 Internal Events 6.04E-09 2.01E-08 3.02E-08 2.42E-08 Combined 8.13E-08 2.71E-07 4.06E-07 3.25E-07 As shown in Table 5-18, the estimated change in LERF of 2.42E-8 for internal events is less than the Regulatory Guide 1.174 [Reference 4] threshold criteria for a very small change when comparing the 15-year results to the current 10-year requirement and the original 3-year requirement.

When both the internal and external event contributions are combined, the total change in LERF of 3.25E-7 meets the guidance for small change in risk, as it exceeds the 1.0E-7/yr and remains less than 1.0E-6 change in LERF. For this change in LERF to be acceptable, total LERF must be less than 1.0E-5. The total LERF value is calculated below:

LERF = LERFinternal + LERFfire + LERFseismic + LERFother + LERFclass3Bincrease

= 2.69E-7/yr + 6.98E-6/yr + 6.24E-7/yr + 1.02E-7/yr + 3.25E-7/yr = 8.30E-6/yr As specified in Regulatory Guide 1.174 [Reference 4], since the total LERF is less than 1.0E-5, it is acceptable for the LERF to be between 1.0E-7 and 1.0E-6.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 30 of 69 5.3.2 Potential Impact from Steel Liner Corrosion Likelihood A quantitative assessment of the contribution of steel liner corrosion likelihood impact was performed for the risk impact assessment for extended ILRT intervals. As a sensitivity run, the internal event CDF was used to calculate the Class 3b frequency. The impact on the Class 3b frequency due to increases in the ILRT surveillance interval was calculated for steel liner corrosion likelihood using the relationships described in Section 5.2.6. The EPRI Category 3b frequencies for the 3 per 10-year, 10-year, and 15-year ILRT intervals were quantified using the internal events CDF. The change in the LERF, change in CCFP, and change in Annual Dose Rate due to extending the ILRT interval from 3 in 10 years to 1 in 10 years, or to 1 in 15 years are provided in Table 5 Table 5-21. The steel liner corrosion likelihood was increased by a factor of 1000, 10000, and 100000. Except for extreme factors of 10000 and 100000, the corrosion likelihood is relatively insensitive to the results.

Table 5 Steel Liner Corrosion Sensitivity Case: 3B Contribution 3b Frequency (3-per-10 year ILRT) 3b Frequency (1-per-10 year ILRT) 3b Frequency (1-per-15 year ILRT)

LERF Increase (3-per-10 to 1-per-10)

LERF Increase (3-per-10 to 1-per-15)

LERF Increase (1-per-10 to 1-per-15)

Internal Event 3B Contribution 5.43E-09 1.81E-08 2.71E-08 1.27E-08 2.17E-08 9.05E-09 Corrosion Likelihood X 1000 5.48E-09 1.92E-08 3.11E-08 1.37E-08 2.56E-08 1.18E-08 Corrosion Likelihood X 10000 6.02E-09 2.94E-08 6.64E-08 2.34E-08 6.04E-08 3.70E-08 Corrosion Likelihood X 100000 1.13E-08 1.31E-07 4.20E-07 1.20E-07 4.09E-07 2.89E-07 Table 5 Steel Liner Corrosion Sensitivity: CCFP CCFP (3-per-10 year ILRT)

CCFP (1-per-10 year ILRT)

CCFP (1-per-15 year ILRT)

CCFP Increase (3-per-10 to 1-per-10)

CCFP Increase (3-per-10 to 1-per-15)

CCFP Increase (1-per-10 to 1-per-15)

Baseline CCFP 7.80E-01 7.85E-01 7.88E-01 4.80E-03 8.24E-03 3.43E-03 Corrosion Likelihood X 1000 7.80E-01 7.85E-01 7.88E-01 4.86E-03 8.33E-03 3.47E-03 Corrosion Likelihood X 10000 7.80E-01 7.85E-01 7.89E-01 5.33E-03 9.13E-03 3.81E-03 Corrosion Likelihood X 100000 7.82E-01 7.92E-01 7.99E-01 1.00E-02 1.72E-02 7.17E-03

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 31 of 69 Table 5-21 -Steel Liner Corrosion Sensitivity: Dose Rate Dose Rate (3-per-10 year ILRT)

Dose Rate (1-per-10 year ILRT)

Dose Rate (1-per-15 year ILRT)

Dose Rate Increase (3-per-10 to 1-per-10)

Dose Rate Increase (3-per-10 to 1-per-15)

Dose Rate Increase (1-per-10 to 1-per-15)

Dose Rate 2.18E-03 7.26E-03 1.09E-02 5.08E-03 8.71E-03 3.63E-03 Corrosion Likelihood X 1000 2.20E-03 7.71E-03 1.25E-02 5.51E-03 1.03E-02 4.75E-03 Corrosion Likelihood X 10000 2.42E-03 1.18E-02 2.67E-02 9.39E-03 2.42E-02 1.49E-02 Corrosion Likelihood X 100000 4.55E-03 5.27E-02 1.69E-01 4.82E-02 1.64E-01 1.16E-01 5.3.3 Expert Elicitation Sensitivity Another sensitivity case on the impacts of assumptions regarding pre-existing containment defect or flaw probabilities of occurrence and magnitude, or size of the flaw, is performed as described in Reference 24. In this sensitivity case, an expert elicitation was conducted to develop probabilities for pre-existing containment defects that would be detected by the ILRT only based on the historical testing data.

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

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

The expert elicitation results are used to develop sensitivity cases for the risk impact assessment. Employing the results requires the application of the ILRT interval methodology using the expert elicitation to change the probability of pre-existing leakage in the containment.

The baseline assessment uses the Jeffreys non-informative prior and the expert elicitation sensitivity study uses the results of the expert elicitation. In addition, given the relationship between leakage magnitude and probability, larger leakage that is more representative of large early release frequency, can be reflected. For the purposes of this sensitivity, the same leakage magnitudes that are used in the basic methodology (i.e., 10 La for small and 100 La for large) are used here. Table 5-22 presents the magnitudes and probabilities associated with the Jeffreys non-informative prior and the expert elicitation used in the base methodology and this sensitivity case.

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

Leakage Size (La)

Expert Elicitation Mean Probability of Occurrence Percent Reduction 10 3.88E-03 86%

100 2.47E-04 91%

Taking the baseline analysis and using the values provided in Tables 5-10 and 5-11 for the expert elicitation sensitivity yields the results in Table 5-23.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 32 of 69 Table 5 JAF Summary of ILRT Extension Using Expert Elicitation Values Accident Class ILRT Interval 3 per 10 Years 1 per 10 Years 1 per 15 Years Base Frequency Adjusted Base Frequency Dose (person-rem)

Dose Rate (person-rem/yr)

Frequency Dose Rate (person-rem/yr)

Frequency Dose Rate (person-rem/yr) 1 5.86E-07 5.86E-07 2.97E+03 1.74E-03 5.53E-07 1.64E-03 5.37E-07 1.59E-03 2

3.31E-09 3.31E-09 6.93E+05 2.29E-03 3.31E-09 2.29E-03 3.31E-09 2.29E-03 3a N/A 9.18E-09 2.97E+04 2.73E-04 3.06E-08 9.09E-04 4.59E-08 1.36E-03 3b N/A 5.84E-10 2.97E+05 1.74E-04 1.95E-09 5.78E-04 2.92E-09 8.68E-04 6

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7a 5.39E-07 5.39E-07 1.28E+06 6.87E-01 5.39E-07 6.87E-01 5.39E-07 6.87E-01 7b 1.49E-06 1.49E-06 6.81E+05 1.01E+00 1.49E-06 1.01E+00 1.49E-06 1.01E+00 8

2.05E-08 2.05E-08 4.55E+05 9.33E-03 2.05E-08 9.33E-03 2.05E-08 9.33E-03 Totals 2.63E-06 2.63E-06 N/A 1.71E+00 2.63E-06 1.71E+00 2.63E-06 1.71E+00 LERF (3 per 10 yrs base)

N/A 1.36E-09 2.34E-09 LERF (1 per 10 yrs base)

N/A N/A 9.74E-10 CCFP 77.42%

77.84%

77.88%

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

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 33 of 69 5.3.4 Control Rod Drive Failure Sensitivity Per a Finding against Supporting Requirement AS-B7, a bounding sensitivity is performed to determine the effect of the CRD system failing. The use of CRD for makeup does not account for time dependency. The success criteria for CRD indicate that it is not a valid source of makeup immediately after a transient. As a bounding sensitivity, CRD is completely failed in the model, and the results are applied to this ILRT extension application. Failing CRD results in a 6.4% increase in CDF. Tables 5-24 and 5-25 present the results of this quantification.

Table 5 Release Category Frequencies Release Category Description of Outcome EPRI Category Frequency (/yr)

NCF No Containment Failure 1

5.86E-07 CIS Containment Isolation Failure 2

3.31E-09 LERF (no bypass)

Large Early Release Frequency (excluding containment bypass) 7a 2.49E-07 MERF Medium Early Release Frequency 7a 3.35E-08 SERF Small Early Release Frequency 7a 3.42E-08 SSERF Small-Small Early Release Frequency 7a 0.00E+00 LIRF Large Intermediate Release Frequency 7a 0.00E+00 MIRF Medium Intermediate Release Frequency 7a 0.00E+00 SIRF Small Intermediate Release Frequency 7a 1.40E-07 SSIRF Small-Small Intermediate Release Frequency 7a 8.16E-08 LLRF Large Late Release Frequency 7b 1.33E-06 MLRF Medium Late Release Frequency 7b 7.17E-08 SLRF Small Late Release Frequency 7b 2.57E-07 SSLRF Small-Small Late Release Frequency 7b 0.00E+00 ISLOCA Interfacing Systems LOCA 8

9.64E-09 LOCAOC LOCA Outside Containment (that leads to LERF) 8 1.09E-08 Contribution to EPRI Classification 2 3.31E-09 Contribution to EPRI Classification 7a 5.39E-07 Contribution to EPRI Classification 7b 1.66E-06 Contribution to EPRI Classification 8 2.05E-08 Table 5 CRD Sensitivity Baseline Risk Profile Class Description Frequency (/yr) 1 No containment failure 5.57E-072 2

Large containment isolation failures 3.31E-09 3a Small isolation failures (liner breach) 2.34E-08 3b Large isolation failures (liner breach) 5.81E-09 4

Small isolation failures - failure to seal (type B) 1 5

Small isolation failures - failure to seal (type C) 1 6

Containment isolation failures (dependent failure, personnel errors) 1 7a Severe accident phenomena induced failure (early) 5.39E-07 7b Severe accident phenomena induced failure (late) 1.66E-06 8

Containment bypass 2.05E-08 Total 2.80E-06

1.

represents a probabilistically insignificant value or a Class that is unaffected by the Type A ILRT.

2.

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

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 34 of 69 The same methodology as described in Section 5.2 is used to perform this sensitivity. As can be seen, even with the CRD system failed, the estimated change in LERF is much less than the threshold criteria for a small change when comparing the 15-year results to the current 10-year requirement and the original 3-year requirement. Table 5-26 summarizes the impact on LERF of this bounding sensitivity. Table 5-27 summarizes the impact on population dose of this bounding sensitivity.

Table 5 CRD Sensitivity: Impact on LERF due to Extended Type A Testing Intervals ILRT Inspection Interval 3 Years (baseline) 10 Years 15 Years Class 3b (Type A LERF) 5.81E-09 1.94E-08 2.91E-08 LERF (3 year baseline) 1.36E-08 2.33E-08 LERF (10 year baseline) 9.69E-09 Table 5 CRD Sensitivity: Impact on Dose Rate due to Extended Type A Testing Intervals ILRT Inspection Interval 10 Years 15 Years Dose Rate (3 year baseline) 5.44E-03 9.33E-03 Dose Rate (10 year baseline) 3.89E-03

%Dose Rate (3 year baseline) 0.298%

0.510%

%Dose Rate (10 year baseline) 0.212%

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 35 of 69 6.0 RESULTS The results from this ILRT extension risk assessment for JAF are summarized in Table 6-1.

Table 6 ILRT Extension Summary Class Dose (person-rem)

Base Case 3 in 10 Years Extend to 1 in 10 Years Extend to 1 in 15 Years Frequency Person-Rem/Year Frequency Person-Rem/Year Frequency Person-Rem/Year 1

2.97E+03 5.58E-07 1.66E-03 4.95E-07 1.47E-03 4.50E-07 1.34E-03 2

6.93E+05 3.31E-09 2.29E-03 3.31E-09 2.29E-03 3.31E-09 2.29E-03 3a 2.97E+04 2.18E-08 6.47E-04 7.27E-08 2.16E-03 1.09E-07 3.24E-03 3b 2.97E+05 5.43E-09 1.61E-03 1.81E-08 5.37E-03 2.71E-08 8.06E-03 6

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7a 1.28E+06 5.39E-07 6.87E-01 5.39E-07 6.87E-01 5.39E-07 6.87E-01 7b 6.81E+05 1.49E-06 1.01E+00 1.49E-06 1.01E+00 1.49E-06 1.01E+00 8

4.55E+05 2.05E-08 9.33E-03 2.05E-08 9.33E-03 2.05E-08 9.33E-03 Total 2.63E-06 1.71E+00 2.63E-06 1.72E+00 2.63E-06 1.72E+00 ILRT Dose Rate from 3a and 3b Total Dose Rate From 3 Years N/A 5.08E-03 8.71E-03 From 10 Years N/A N/A 3.63E-03

%Dose Rate From 3 Years N/A 0.296%

0.508%

From 10 Years N/A N/A 0.211%

3b Frequency (LERF)

LERF From 3 Years N/A 1.27E-08 2.17E-08 From 10 Years N/A N/A 9.04E-09 CCFP %

CCFP%

From 3 Years N/A 0.480%

0.824%

From 10 Years N/A N/A 0.343%

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 36 of 69

7.0 CONCLUSION

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

Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.0E-06/year and increases in LERF less than 1.0E-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in internal events LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 2.17E-8/year using the EPRI guidance; this value increases slightly to 2.20E-8/year if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].

When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 3.16E-7/year using the EPRI guidance, and baseline LERF is 8.29E-6/year. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4]. As discussed in Sections 5.1.3 and 5.3.1, the EPRI methodology used to estimate the increase in LERF is conservative. Therefore, even though the increase in LERF is near the Regulatory Guide 1.174 threshold, there is additional margin because conservative methodology was used.

The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.0087 person-rem/year. EPRI Report No. 1009325, Revision 2-A [Reference 24] states that a very small population dose is defined as an increase of 1.0 person-rem per year, or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. The results of this calculation meet these criteria. Moreover, the risk impact for the ILRT extension when compared to other severe accident risks is negligible.

The increase in the conditional containment failure probability from the 3 in 10 year interval to 1 in 15 year interval is 0.824%. EPRI Report No. 1009325, Revision 2-A

[Reference 24] states that increases in CCFP of 1.5% is very small. Therefore, this increase is judged to be very small.

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

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

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

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 37 of 69

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

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

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 38 of 69 A.

ATTACHMENT 1 - PRA TECHNICAL ADEQUACY A.1.

Internal Events PRA Quality Statement for Permanent 15-Year ILRT Extension The JAF PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause failure events. The PRA model quantification process used for the JAF PRA is based on the event tree and fault tree methodology, which is a well-known methodology in the industry.

Entergy employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating Entergy nuclear power plants.

This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the JAF PRA model.

A.1.1 PRA Maintenance and Update The Entergy risk management process ensures that the applicable PRA model is an accurate reflection of the as-built and as-operated plant. This process is defined in the procedure EN-DC-151, "PSA Maintenance and Update." This procedure delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating Entergy nuclear power plants. In addition, the procedure also defines the process for implementing regularly scheduled and interim PRA model updates, and for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, industry operating experience, etc.). To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plant, the following activities are routinely performed:

Design changes and procedure changes are reviewed for their impact on the PRA model. Potential PRA model changes resulting from these reviews are entered into the Model Change Request (MCR) database, and a determination is made regarding the significance of the change with respect to the current PRA model.

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

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

Industry standards, experience, and technologies are periodically reviewed to ensure that any changes are appropriately incorporated into the models.

In addition, following each periodic PRA model update, Entergy performs a self-assessment to assure that the PRA quality and expectations for all current applications are met. The Entergy PRA maintenance and update procedure requires updating of all risk informed applications that may have been impacted by the update including but not limited to:

System/component risk significance rankings

PRA training materials

AOV / MOV Risk Rankings

Online Risk Model (EOOS)

Mitigating System Performance Index input (MSPI)

The model results used in this calculation are quantified from the Living 2011 JAF PRA model, which is outside of the internally required four year periodic update cycle of the model of record, JAF-NE-09-00001 Fitzpatrick Probabilistic Safety Assessment (PSA), Rev 4, (Reference 45).

The results are considered reasonable for the ILRT extension risk analysis because potential

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 39 of 69 changes to the model are assessed using the process described in this section, and because plant specific initiating events, failure rates, and maintenance unavailability changes are not expected to result in an increase in the relatively low JAF PRA total CDF, LERF, and corresponding calculated delta LERF sufficient to challenge the conclusion of the analysis with respect to the risk acceptance criteria in Regulatory Guide 1.174 [Reference 4].

A.1.2 BWROG Regulatory Guide 1.200 Peer Review of the JAF PRA Model The JAF PRA internal events model went through BWROG Regulatory Guide 1.200 peer review in September 2009. The NEI 05-04 process [Reference 33], the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA Standard [Reference 27],

and Regulatory Guide 1.200, Rev. 2 [Reference 38]) were used for the peer review.

The 2009 JAF PRA Peer Review was a full-scope review of all the Technical Elements of the internal events, at-power PRA:

Initiating Events Analysis (IE)

Accident Sequence Analysis (AS)

Success Criteria (SC)

Systems Analysis (SY)

Human Reliability Analysis (HR)

Data Analysis (DA)

Internal Flooding (IF)

Quantification (QU)

LERF Analysis (LE)

Maintenance and Update Process (MU)

The JAF PRA Peer Review process uses capability categories to assess the relative technical merits and capabilities of each technical supporting requirement reviewed. Three capability category levels are used to indicate the relative quality level of each supporting requirement.

Capability category assignments are made based on the judgment of the Peer Review Team after reviewing: (1) the PRA model, (2) the documentation; and, (3) the prior PRA Peer Review results (for historical background).

During the JAF PRA model Peer Review, the technical elements identified above were assessed with respect to Capability Category II criteria to better focus the Supporting Requirement assessments. The ASME/ANS PRA Standard has 326 individual Supporting Requirements; 310 Supporting Requirements are applicable to the JAF PRA model. Sixteen (16) of the ASME/ANS PRA Standard Supporting Requirements are not applicable to JAF (e.g.,

PWR related, multi-site related). Of the 310 ASME/ANS PRA Standard Supporting Requirements applicable to the JAF PRA model, approximately 94% satisfied Capability Category II criteria or greater. Of the 53 Findings and Observations (F&Os) generated by the Peer Review Team, 24 were considered Findings, 27 were Suggestions, and 2 were Best Practices.

A.1.3 Consistency with Applicable PRA Standards As a result of the BWROG Regulatory Guide 1.200 peer review, 51 F&Os have been identified for potential improvement to the JAF PRA model. These F&Os are tracked in the Entergy Model Change Request (MCR) database. Of the identified 51 F&Os, 21 were considered not meeting at least the Capability Category II criteria. Table A-1 summarizes the open F&Os along with an initial assessment of the impact for this application. For each F&O in Table A-1, applicability for this ILRT extension application is evaluated. If an impact is not expected to be negligible, then

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 40 of 69 this assessment may include the performance of additional sensitivity studies or PRA model changes to confirm the impact on the risk analysis and justify why the change does not impact the PRA results used to support the application. Table A-2 summarizes the closed F&Os and the actions taken to close the F&Os.

A.2.

External Events PRA Quality Statement for Permanent 15-Year ILRT Extension External hazards were evaluated in 1996 in the JAF Individual Plant Examination of External Events (IPEEE) submittal in response to the NRC IPEEE Program (Generic Letter 88-20, Supplement 4) [Reference 39]. The IPEEE Program was a one-time review of external hazard risk and was limited in its purpose to the identification of potential plant vulnerabilities and the understanding of associated severe accident risks.

The results of the JAF IPEEE study are documented in the JAF IPEEE Report [Reference 35].

The primary areas of external event evaluation at JAF were internal fire, seismic, high winds, floods, and other external hazards.

A.2.1 Fire Analysis The JAF IPEEE fire analysis was performed using EPRIs Fire PRA Implementation Guide

[Reference 40]. The EPRI Fire-Induced Vulnerability Evaluation method was used for the initial screening, for treatment of transient combustibles, and as the source of fire frequency data

[Reference 41].

The fire analysis was revised after the original IPEEE submittal in response to NRC requests for additional information (RAIs) regarding fire-modeling progression, developed during their review of the IPEEE. The updated results are reflected in NUREG-1742, Perspectives Gained from the Individual Plant Examination of External Events (IPEEE) Program [Reference 42] for JAF. In addition, as noted in that NUREG, a number of plant and procedural changes (including strict limitations on storage and use of combustible and flammable material in plant areas) were made as a result of the IPEEE fire analysis. The impact of these enhancements is not reflected in the IPEEE fire results.

Other changes to the plant configuration, procedures and equipment performance have also taken place since completion of the IPEEE. These changes would tend to reduce the overall CDF as well as the fire risk contribution found in the IPEEE. The significant reduction in the internal events CDF since the original JAF IPEEE submittal bears this out. These changes include the following:

Service, instrument, and breathing air compressors were replaced.

Operators are directed to maximize CRD flow in certain accident sequences.

SRV Electric Lift mod to install an SRV alternate actuation system.

A new procedure (EP-10) directs operators to align the fire protection system to the tube side of the RHR heat exchanger in loss of containment heat removal accident sequences.

Revised station blackout procedures to explicitly address bus recovery.

Provision of a back-up battery charger that can be aligned to either station battery.

Proceduralized RCIC operation without DC power.

Proceduralized starting EDG without DC power, as well as field flashing without station batteries.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 41 of 69 The JAF IPEEE fire risk model has been updated since its original issuance to include new conditional core damage probabilities via a failure probability data update to the basic events in the original cutsets. Further update to the fire risk model, using updated ignition frequencies, circuit failure probabilities, and with re-analysis of containment bypass scenarios has also been performed to support the external events risk study in section 5.3.1.

A.2.2 Seismic Analysis In the IPEEE, FitzPatrick performed a SMA, which is a deterministic evaluation that does not calculate risk on a probabilistic basis. The conclusions reached in 2010 by GI-199 [Reference 36] are used for estimating Seismic CDF in this calculation.

A.2.3 High Winds, Floods, and Other External Hazards (HFO)

In addition to internal fires and seismic events, the JAF IPEEE analysis of HFO external hazards was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards. Since JAF was designed (with construction started) prior to the issuance of the 1975 Standard Review Plan (SRP) criteria, JAF performed a plant hazard and design information review for conformance with the SRP criteria. For HFO events that were not screened out by compliance with the 1975 SRP criteria, additional analyses were performed to determine whether or not the hazard frequency was acceptably low. Based on those analyses, these hazards were determined in the JAF IPEEE to be negligible contributors to overall plant risk.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 42 of 69 Table A-1 Internal Events PRA Peer Review - Open Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application Finding The use of CRD for makeup does not account for time dependency. Gate U3 is used after containment heat removal fails and venting is a success. The success criteria for CRD indicate that it is not a valid source of makeup immediately after a transient. It is implicit that HPCI or RCIC work until containment is vented.

AS-B7 MODEL time-phased dependencies (i.e., those that change as the accident progresses, due to such factors as depletion of resources, recovery of resources, and changes in loads) in the accident sequences.

Open. Although success of early injection (from HPCI, RCIC, LPCI, core spray) is not explicitly modeled in the event trees, CRD is only credited on the success branch, because early injection sequences (if initially successful) fail long-term due to the accident phenomena cause by the loss of containment heat removal. In addition, the random failure of early injection based on past event tree development is truncated out during accident sequence quantification.

Not significant. Changes in the ILRT frequency is not relevant to the modeling of the long-term use of the CRD system as a source of RPV injection during a loss of containment decay heat event.

A bounding sensitivity study performed in Section 5.3.5 shows that removing all credit for CRD resulted in a 6.4%

increase in CDF and no increase in LERF. As shown in Section 5.3.5, this results in an increase in LERF of 7.2%, to 2.33E-08, which is well within Region III of the R.G. 1.174 criteria. Therefore, this F&O has no impact on the ILRT extension application.

Finding Appendix H4 states that

'When applying the dependency model, the dependency was generally applied to the HEP with the higher HEP (versus assigning the dependency to the action which occurs later in time). This does not conform to the established calculation method. In general, it is more appropriate to choose the event that would occur first HR-G7 For multiple human actions in the same accident sequence or cut set, identified in accordance with supporting requirement QU-C1, ASSESS the degree of dependence, and calculate a joint human error probability that reflects the dependence.

ACCOUNT for the influence of success or failure in preceding human actions and system performance on the human event under consideration including Open. The HRA guidance on assessing dependency between post-initiator HFEs has been modified to incorporate this peer review comment. However, the new HRA guidance has not yet been incorporated into the JAF PRA model.

Not significant. This change does not have a significant impact on the results. There is an observation that the lower probability HFEs generally occur earlier in the sequences; this is an observation of the important HEPs to the JAF model (not an assertion about all PRAs).

The assigned dependencies are often conservative because they are based on the overall HEP value

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 43 of 69 Table A-1 Internal Events PRA Peer Review - Open Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application in the accident sequence.

The reason for this is that if the operators failed on the first action, then they are more likely to fail on subsequent actions (unless there is an intervening success). If the events occur relatively close in time, then it may be acceptable to choose the one with the lowest probability. The reason for this is that the lower probability is usually the easier action to perform or the one with more time available.

(a) time required to complete all actions in relation to the time available to perform the actions (b) factors that could lead to dependence (e.g., common instrumentation, common procedures, increased stress, etc.)

(c) availability of resources (e.g.,

personnel)

(whereas dependency between the execution portions of the HEPs can frequently be justified as being low or zero).

Re-analysis of the combined HEPs to address this F&O resulted in a CDF increase of 1.73% and a LERF increase of 0.02%. These increases are bounded by the sensitivity study described in the evaluation of the F&O regarding AS-B7 (the increase in total LERF does not impact the criteria in Region III and not subtracting it from the 3b frequency calculation is conservative since all additional CDF is assumed to be a class 3b LERF); therefore, there is no impact on the ILRT extension application.

Finding The JAF PRA determined a point estimate for core damage and documented an analysis for parametric uncertainty. However, the state of knowledge correlation was not fully accounted for due to the manner in which the JAF basic event data base is constructed.

QU-A3 ESTIMATE the mean CDF accounting for the state-of-knowledge correlation between event probabilities when significant.

Open. As future updates are performed, the state of knowledge correlation (SOKC) will be re-examined as part of the quantification process.

Not significant. A sensitivity performed at the time of the RG 1.200 peer review demonstrated that the state of knowledge correlation is not a significant issue for the current JAF PRA model. This is evidenced by the fact that the mean and point estimate CDF and LERF values are very close in value.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 44 of 69 Table A-1 Internal Events PRA Peer Review - Open Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application Finding Ranking of components by RAW and RRW is presented in tables I.3-1 and I.3-2. However, no discussion was found with respect to the reasonableness of the ranking.

QU-D7 REVIEW the importance of components and basic events to determine that they make logical sense.

Open. Reasonableness of the SSC rankings was verified through a review of the results, which was performed prior to issuing the PSA report.

None. This is a documentation issue only.

Finding The level 2 model uses point estimates for success branches while the failure logic is modeled for the failure branch LE-C4 INCLUDE model logic necessary to provide a realistic estimation of the significant accident progression sequences resulting in a large early release. INCLUDE mitigating actions by operating staff, effect of fission product scrubbing on radionuclide release, and expected beneficial failures in significant accident progression sequences.

PROVIDE technical justification (by plant-specific or application generic calculations demonstrating the feasibility of the actions, scrubbing mechanisms, or beneficial failures) supporting the inclusion of any of these features.

Open. For Level 2 top events that include severe accident phenomena, those phenomena dominate and use of point estimate values is typical and appropriate.

Where phenomena are not involved, system related success terms do not influence the outcome of the sequence quantification.

Not significant. Since use of point estimates is reasonable for top events that include phenomena, and system related success probabilities are approximately equal to 1.0, no impact on the application is expected.

Finding The definition of 'Early' is inconsistent within the document and may classify some LERF sequences as non-LERF.

LE-E3 INCLUDE as LERF contributors potential large early release (LER) sequences identified from the results of the accident progression analysis of LE-C except those LER sequences justified as non-LERF contributors in LE-C1.

Open. Resolution of this F&O entails defining an Early release based on the time to declare a General Emergency and subsequent linking of each accident progression sequence to a declaration time for Minimal. It is possible this issue causes LERF to be slightly miscalculated. A bounding sensitivity is performed in Section 5.3.1.1; it shows any change from resolving this F&O is insignificant to this application. Also, as shown in

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 45 of 69 Table A-1 Internal Events PRA Peer Review - Open Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application comparison to the 'Early' criterion.

Sections 5.2.5, 5.3.1, and 5.3.1.1, there is significant margin between the calculated results and the Regulatory Guide 1.174

[Reference 4] threshold criteria. Therefore, a small change in LERF will not affect the conclusions of this calculation.

Finding Spray-induced and submergence induced failures appear to have been addressed in the analysis. No documentation of a systematic assessment of the effects of jet impingement, pipe whip, humidity, temperature, etc.,

on SSCs could be identified. No evaluation of the specific equipment evaluated in the PRA compared to equipment considered in the design analyses, e.g., EQ lists, was documented. Since the PRA can credit non-safety-related equipment, relying on design basis evaluations to dismiss these dynamic effects may credit equipment that cannot withstand the effects considered in the design analysis. In IFSN-A6 For the SSCs identified in IFSN-A5, IDENTIFY the susceptibility of each SSC in a flood area to flood-induced failure mechanisms INCLUDE failure by submergence and spray in the identification process.

EITHER (a) ASSESS qualitatively the impact of flood-induced mechanisms that are not formally addressed (e.g., using the mechanisms listed under Capability Category III of this requirement), by using conservative assumptions; OR (b) NOTE that these mechanisms are not included in the scope of the evaluation Open. The additional impacts noted in the F&O are only required for Capability Category III. The only requirement to meet Capability Category II is to document that those mechanisms are not included in the scope. This will be documented in a future update.

Capability Category I is sufficient to provide an estimate of the total Internal Flooding CDF and LERF as used in the ILRT extension PRA analysis. Meeting Capability Category II could provide additional qualitative discussion regarding impact of flood-induced mechanisms that are not formally addressed using the mechanisms in Capability Category III, but would not impact the total CDF and LERF used in the ILRT extension analysis and therefore would not affect the final conclusions of this application.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 46 of 69 Table A-1 Internal Events PRA Peer Review - Open Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application addition, failure in a system containing high temperature fluid can actuate fire systems and impact additional equipment. Also, the PRA models may evaluate breaks beyond those of the design basis.

Finding Although the corresponding initiating event group is identified for several of the internal flooding scenarios, there is no such information provided in the internal flooding documentation for the vast majority of the scenarios.

IFEV-A1 For each flood scenario, IDENTIFY the corresponding plant initiating event group identified per 2-2.1 and the scenario-induced failures of SSCs required to respond to the plant initiating event. INCLUDE the potential for a flooding-induced transient or LOCA.

If an appropriate plant-initiating event group does not exist, CREATE a new plant-initiating event group in accordance with the applicable requirements of 2-2.1.

Open. Although not specifically provided in the internal flooding analysis documentation, the link to the initiating event utilized and the impacted equipment is provided in the flag file for each flooding initiator.

None. This is a documentation enhancement issue.

Finding Quantification of initiating event frequency is not documented in the Internal Flooding Analysis. A series of spreadsheets that were used to quantify internal flooding initiating event frequency values were provided. The documentation does not allow a reviewer to IFEV-A5 DETERMINE the flood initiating event frequency for each flood scenario group by using the applicable requirements in 2-2.1.

Open. As noted, the required information is provided in a series of spreadsheets.

These spreadsheets will be integrated in a single easily reviewed format in a future update.

None. This is a documentation enhancement issue.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 47 of 69 Table A-1 Internal Events PRA Peer Review - Open Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application correlate the piping and areas considered for each initiating event without recourse to the author and spreadsheets that are maintained outside of the approved documentation.

Finding Review and consideration of plant-specific information is required by the SR to meet Capability Category II.

IFEV-A6 GATHER plant-specific information on plant design, operating practices, and conditions that may impact flood likelihood (i.e., material condition of fluid systems, experience with water hammer, and maintenance-induced floods). In determining the flood-initiating event frequencies for flood scenario groups, USE a combination of the following:

(a) generic and plant-specific operating experience (b) pipe, component, and tank rupture failure rates from generic data sources and plant-specific experience (c) engineering judgment for consideration of the plant-specific information collected Open. A search of the JAF condition reporting system was performed for a period of 15 years for the Internal Flooding Analysis. No significant internal flooding events were identified which would significantly alter the generic data.

Not significant. Since plant specific data was reviewed for applicability and determined not to change the input, this is considered a documentation enhancement issue.

Finding Appendix C does not contain evidence of a review of initiating events for applicability during flood scenarios. The development of event trees and selection of initiating IFQU-A1 For each flood scenario, REVIEW the accident sequences for the associated plant initiating event group to confirm applicability of the accident sequence model.

If appropriate accident sequences do not exist, MODIFY sequences Open. Although not specifically provided in the internal flooding analysis documentation, evidence of the initiating event utilized and the impacted equipment None. This is a documentation enhancement issue.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 48 of 69 Table A-1 Internal Events PRA Peer Review - Open Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application events is only briefly discussed in Section C3.3.

as necessary to account for any unique flood-induced scenarios and/or phenomena in accordance with the applicable requirements described in 2-2.2.

is provided in the flag file for each flooding initiator.

Table A-2 Internal Events PRA Peer Review - Closed Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application Finding The time dependency of the DC to support HPCI (U1) and RCIC (U2) is inadequate. The battery and chargers are both in an AND gate.

This is incorrect for mission times greater than four hours. The battery, with load shed, will last four hours and needs the charger after that; therefore, the sole dependency should be on the battery charger.

AS-B7 MODEL time-phased dependencies (i.e., those that change as the accident progresses, due to such factors as depletion of resources, recovery of resources, and changes in loads) in the accident sequences.

Closed. Modified dependency logic for HPCI and RCIC fault trees (gates GHCI-FLC108 and GRCI-FLC-1391, respectively) to reflect that continued operation requires availability of both the charger and battery.

None. This F&O has been addressed and incorporated into the model.

Finding Appendices E1-E36 show many systems where support systems are explicitly modeled or assumed to be necessary in the absence of engineering analyses to determine whether they are needed.

Example: battery ventilation system.

SY-B6 PERFORM engineering analyses to determine the need for support systems that are plant-specific and reflect the variability in the conditions present during the postulated accidents for which the system is required to function.

Closed. Plant engineering analyses, where available, were used in modeling support systems. In the specific case of the battery ventilation system, the calculation did not provide sufficient basis for concluding that ventilation was unnecessary. Therefore, the Battery Room Ventilation System was included as a support system to the battery chargers.

None. Support system modeling was included where needed or best estimate analyses were not available or sufficient.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 49 of 69 Table A-2 Internal Events PRA Peer Review - Closed Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application Finding The majority of support systems are based on conservative success criteria and timing, but the criteria are not justified based on the impact on risk significant contributors.

SY-B7 BASE support system modeling on realistic success criteria and timing, unless a conservative approach can be justified (i.e., if their use does not impact risk significant contributors).

Closed. A further review of the existing support system success criteria documentation determined that use of conservative success criteria and timing was limited and where used, it had little impact on results.

None. This was a documentation issue only and has been addressed.

Finding RCIC/HPCI flow is delivered to the reactor vessel via feedwater line A/B. The HPCI fault tree includes the failure of the downstream feedwater check valve and manual valve but the RCIC fault tree does not.

SY-B13 Some systems use components and equipment that are required for operation of other systems. INCLUDE components that, using the criteria in SY-A15, may be screened from each system model individually, if their failure affects more than one system (e.g., a common suction pipe feeding two separate systems).

Closed. The RCIC system fault tree has been revised to include the A feedwater check valve and manual valve.

None. This F&O has been addressed and incorporated into the model.

Finding The documentation contained no evidence of checking the reasonableness of the posterior distribution.

DA-D4 When the Bayesian approach is used to derive a distribution and mean value of a parameter, CHECK that the posterior distribution is reasonable given the relative weight of evidence provided by the prior and the plant-specific data. Examples of tests to ensure that the updating is accomplished correctly and that the generic parameter estimates are consistent with the plant-specific application include the following:

(a) confirmation that the Bayesian updating does not produce a posterior distribution with a single bin histogram (b) examination of the cause of any unusual (e.g., multimodal) posterior distribution shapes Closed. Detailed discussions that describe the process of confirming the reasonableness of the posterior distribution mean value have been added to Appendix D.

None. This was a documentation issue only and has been addressed.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 50 of 69 Table A-2 Internal Events PRA Peer Review - Closed Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application (c) examination of inconsistencies between the prior distribution and the plant-specific evidence to confirm that they are appropriate (d) confirmation that the Bayesian updating algorithm provides meaningful results over the range of values being considered (e) confirmation of the reasonableness of the posterior distribution mean value Finding Significant contributors to CDF by initiating events, accident sequence, systems and operator errors were identified and discussed in the Summary Report, Sections 3.2 through 3.6. A listing of basic event importance is provided in Appendix I.3. It does not appear the SSCs and HFEs associated with initiating events modeled by fault tree were included.

QU-D6 IDENTIFY significant contributors to CDF, such as initiating events, accident sequences, equipment failures, common cause failures, and operator errors. INCLUDE SSCs and operator actions that contribute to initiating event frequencies and event mitigation.

Closed. Significant contributors to CDF, such as initiating events, accident sequences, equipment failures, common cause failures, and operator errors were identified.

The importance of SSCs and HFEs that contribute to initiating event frequencies and event mitigation are identified in the results to the extent possible. In general, it was found that basic events associated with initiating events which are important contributors to CDF will also be important for accident mitigation and their importance will be reasonably accounted for using the current method for importance ranking.

None. The importance of

systems, components and operator actions has been reasonably accounted for in the PRA model.

Finding While descriptions and some discussion of the top ten accident sequences are provided, additional information with regard QU-F3 DOCUMENT the significant contributors (such as initiating events, accident sequences, basic events) to CDF in the PRA results summary.

Closed. Each of the top 95%

accident sequences was provided with a description which summarizes the failures None. This was a documentation issue only and

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 51 of 69 Table A-2 Internal Events PRA Peer Review - Closed Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application to the contributors to the sequence frequency should be included.

PROVIDE a detailed description of significant accident sequences or functional failure groups.

and/or successes for that sequence. In addition, detailed descriptions were provided for all sequences in Appendix F.

has been addressed.

Finding Accident sequences that credited reactor building plate-out (reducing the release magnitude one classification) have not been justified.

LE-C1 DEVELOP accident sequences to a level of detail to account for the potential contributors identified in LE-B1 and analyzed in LE-B2.

Compare the containment challenges analyzed in LE-B with the containment structural capability analyzed in LE-D and identify accident progressions that have the potential for a large early release.

JUSTIFY any generic or plant-specific calculations or references used to categorize releases and non-LERF contributors based on release magnitude or timing. NUREG/CR-6595, App. A [2-16] provides a discussion and examples of LERF source terms.

Closed. The appendix addressing radionuclide release impacts was modified to document that for sequences where the Reactor Building (RB) node is effective, MAAP calculations indicate a substantial reduction in the release.

None. This was considered a documentation issue only and has been addressed.

Finding LERF sequences are identified by their containment failure (or bypass) mode and confirmed by accident progression calculations documented in Section J1.5.5 and Tables J1.5-6 and J1.5-8. Some MAAP runs in Table J1.5-6 labeled as Large-Early are not included in LERF because the MAAP run may only be implemented as a conservative run for a late scenario.

LE-E3 INCLUDE as LERF contributor potential large early release (LER) sequences identified from the results of the accident progression analysis of LE-C except those LER sequences justified as non-LERF contributors in LE-C1.

Closed. Added the following sentence to the last paragraph of Section J1.5.5:

Note that not all the MAAP analyses represented in Table J1.5-6 and Table J1.5-8 were used. The MAAP analyses selected reflect the most appropriate cases in terms of predicting the accident progression for each of the large early release endstates.

None. This was a documentation issue only and has been addressed.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 52 of 69 Table A-2 Internal Events PRA Peer Review - Closed Facts and Observations F&O Significance Finding/Observation Description Applicable SRs SR Description Current Status / Comment Importance to Application Finding Limitations discussed in Appendix J are focused on model conservatisms and other LERF issues, but these limitations are not addressed in terms of their impact on applications.

LE-G5 IDENTIFY limitations in the LERF analysis that would impact applications.

Closed. A discussion was added in Appendix J to document that no limitations of the LERF quantification process were identified that would impact applications.

None. This was a documentation issue only and has been addressed.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 53 of 69 B.

ATTACHMENT 2 - PRA RAI RESPONSES Probabilistic Risk Assessment Licensing Branch (APLA) RAI-1a Section 4.2.7 of Electric Power Research Institute (EPRI) Topical Report (TR)-1009325, Rev. 2-A, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325" states that "[w]here possible, the analysis should include a quantitative assessment of the contribution of external events (for example, fire and seismic) in the risk impact assessment for extended [Integrated Leak Rate Test] ILRT intervals. For example, where a licensee possesses a quantitative fire analysis and that analysis is of sufficient quality and detail to assess the impact, the methods used to obtain the impact from internal events should be applied for the external event."

The "Fire Analysis" section of Attachment 1 to the LAR states that "although the JAF Individual Plant Examination of External Events (IPEEE) fire risk model has not been updated since its original issuance, use of the IPEEE model would tend to give conservative results."

A fire PRA model with sufficient quality and detail that has undergone a peer review under American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) RA-SA-2009, "Addenda to ASME/ANS RA-S-2008 Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications", as clarified/qualified by Rev. 2 of Regulatory Guide 1.200 (RG 1.200), "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," is not utilized for this application. The IPEEE model was issued 20 years ago in June 1996, and there have been significant changes to Fire PRA methodology since then.

a. Identify any gaps between the FitzPatrick IPEEE fire risk model and the "Internal Fire Technical Elements" required by Rev. 2 of RG 1.200 that are relevant to this submittal and explain why these gaps do not have significant impact on the risk assessment results for this application.

OR Demonstrate that addressing the requirements would have no impact on the final results in the LAR by providing a sensitivity evaluation to show that the use of the IPEEE results remains conservative-replace the IPEEE results, including at least the following:

i.

Reanalysis of all containment bypass scenarios leading to LERF from the internal events PRA, replacing any random failure probabilities with appropriate fire-induced values (e.g., NUREG/CR-7150 values for MOVs for ISLOCAs) this should include any ISLOCA pathways previously screened out due to low random failure probability, as the probability could be significantly higher if fire induced.

ii. Reanalysis of the IPEEE fire LERF using updated fire frequencies and, if suppression was credited, updated non-suppression probabilities from NUREG-2169.

Response

All containment bypass scenarios leading to LERF were re-analyzed by replacing the applicable random failure probabilities with appropriate fire-induced values from NUREG/CR-7150. The bypass scenarios include both potential ISLOCA pathways and loss of containment isolation

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 54 of 69 pathways. The re-analysis for the potential ISLOCA pathways results in an increase in both fire CDF and fire LERF.

The fire LERF was re-analyzed using the updated fire frequencies and non-suppression probabilities from NUREG/CR-2169. The fire CDF was also re-analyzed based on the re-analysis of the ISLOCA pathways per part 1a.i of this Request for Additional Information.

All IPEEE fire scenarios were revised to include the updated ignition frequencies and non-suppression probabilities. An assessment of which scenarios would include risk contribution from the bypass scenarios not previously analyzed in the IPEEE was performed, which included the following methodology:

1. Conservatively consider all fire scenarios as impacting the valves whose failure would potentially lead to a containment bypass

2. Review the risk significant fire scenarios considering the cable routing information available in the JAF Safe Shutdown Analysis.

3. Remove the fire-induced failure of valves whose failure would potentially lead to a containment bypass scenario from the fire scenarios in Fire Areas where no cables associated with the valve are routed.

Note that not all containment isolation scenarios were considered in the Safe Shutdown Analysis and without cable routing information, the fire-induced failure of each valve was included in all IPEEE fire scenarios, with the exception of one containment isolation scenario involving two normally open valves. For the containment isolation scenarios that involve normally closed, fail-closed, air-operated valves, credit for hot short duration can be taken from NUREG/CR-7150 which limits the risk of conservatively including failure of the containment isolation system valves in all IPEEE fire scenarios. For the containment isolation scenario that involves two normally open valves, further assessment is needed.

The containment isolation scenario that involves two normally open valves is potentially risk significant to LERF when conservatively included in all IPEEE fire scenarios. The risk significant scenario would include a core damage event where the fire also failed the containment isolation signal, and power and air are still available to the fail-closed air-operated valve. The other isolation valve is a normally open motor-operated valve. The valves are associated with the drywell floor drain containment penetration. Each of these two valves receives multiple isolation signals, including a signal generated on the highest reactor vessel low water level, which is the level at which a SCRAM occurs. Fires that can cause both a core damage event and a failure of the containment isolation signal are most likely to occur in the Relay Room, Cable Spreading Room, and Main Control Room. For those Relay Room, Cable Spreading Room, and Control Room fires that would affect only one division of containment isolation and not affect air or power, the other division would be available to provide an isolation signal to the valves. Fires that fail air or power would result in closure of the fail closed, air-operated valve (unless a random failure occurs). Large fires that would affect both divisions, including those that would result in usage of the alternate shutdown strategy, could also impact air or power, which would result in closure of the fail closed, air-operated valve (unless a random failure occurs). Based on this, this re-analysis assumes that significant fires, which are defined as those in which suppression fails, a hot gas layer occurs, or the fire affects more than one division in the Relay Room, Cable Spreading Room, and Control Room will not occur in a way in which air or power to the fail-closed, air-operated valve is still available and the valve remains open.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 55 of 69 The results of this assessment, therefore, include the risk of the potential bypass scenario in all IPEEE fire scenarios in Fire Areas where the cables for the associated valves are routed, regardless of whether suppression success was modeled to limit the postulated fire damage for individual scenarios, with the exception of the containment isolation scenario described in the preceding paragraph. Conservatively, the bypass scenarios were also included in all IPEEE multi-compartment scenarios. The effect of the conservative inclusion of bypass scenarios in the multi-compartment analysis is relatively small given the relatively small overall contribution of the multi-compartment scenarios.

There is an increase in both fire CDF and fire LERF due to the evaluation performed above. The fire CDF increased from 2.12E-05 to 2.27E-05, a change of approximately 7%. The IPEEE did not calculate a fire LERF, so LERF is calculated as follows:

1. Obtain the ratio of LERF to CDF from the JAF Internal Events PRA.

2. Multiply the fire CDF by the ratio

3. Add the contribution from the re-analyzed containment bypass scenarios.

The re-analyzed fire LERF is 6.20E-06, an increase of approximately 186%.

An assessment of the LERF and total LERF due to the ILRT extension is performed using the revised fire CDF and LERF to demonstrate that the results of the risk analysis can be reasonably shown to be within Region II of the risk acceptance guidelines in RG 1.174. The assessment of the LERF and total LERF should be bounding given the re-analysis of the fire risk in response to this RAI and in response to RAI-6, which is related to fire risk. Thus, the response to RAI-6 is addressed here. RAI-6 states the following:

In Section 5.3.1.of Attachment 4 to the LAR, the fire LERF result was calculated based on the fire CDF (2.12E-5) taken from the original JAFs IPEEE, James A. FitzPatrick Nuclear Power Plant Individual Plant Examination of External Events, which was published in June, 1996.

However, an updated fire CDF (3.42E-5) that is higher than the one listed in the original IPEEE is available in Section 2.1.7 of James A. FitzPatrick Nuclear Power Plant - Review of FitzPatrick Individual Plant Examination of External Events (IPEEE) Submittal (TAC NO. M83622) dated September 21, 2000 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML003743771). Using a smaller fire CDF value from the original IPEEE to obtain the fire LERF produced a less conservative fire LERF value in the LAR.

Subsequent to resolving the question in RAI 1.a above regarding the conservatism when using the JAF Fire IPEEE risk, confirm that the total change in LERF is still within the very small and small risk increase regions of Regulatory Guide 1.174. That is, perform the calculations in Section 5.3 using the updated fire CDF value and correct all subsequent calculations in the LAR that are affected by using the non-conservative old fire CDF value (2.12E-5).

Section 2.1.7 of James A. FitzPatrick Nuclear Power Plant - Review of FitzPatrick Individual Plant Examination of External Events (IPEEE) Submittal (TAC NO. M83622) dated September 21, 2000 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML003743771), shows that credit for the alternate shutdown strategy was removed for fires that result in a hot gas layer in the Relay Room. This credit was still taken in the IPEEE fire PRA. No explanation is given as to why alternate shutdown would not be feasible given a hot gas layer fire in section 2.1.7.

Without a full re-evaluation of the hot gas layer fires in the Relay Room, credit for alternate shutdown has been conservatively removed from the applicable scenarios and the fire CDF and LERF and this LERF and total LERF of the ILRT extension analysis has been re-assessed. The resulting fire CDF

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 56 of 69 is 2.57E-05 and the fire LERF is 6.98E-06. The increase in CDF is not as significant as shown in Section 2.1.7 of ML003743771 because the update of ignition frequencies from NUREG/CR-2169 results in a lower fire scenario frequency for the hot gas layer fires in the Relay Room, largely due to the inclusion of plant-wide cabinets when estimating the ignition frequency for the scenario using the NUREG/CR-2169 generic ignition frequency for electrical cabinets as opposed to the EPRI FIVE method which apportioned generic frequency based on cabinets located in the control building.

Given the values of fire CDF and LERF above, the ILRT extension analysis risk metrics can be calculated. Note that because the re-analyzed fire LERF may be conservative, LERF is not subtracted from the existing CDF:

LERF = 2.57E-05 *.0023

• 5 - 2.57E-05 *.0023 = 2.96E 5.91E-08 = 2.36E-07.

The total LERF is the sum of LERF from the internal events and external event models discussed in Section 5.3.1. The revised total LERF is shown in the following table:

Table B JAF External Event Impact on ILRT LERF Calculation Hazard EPRI Accident Class 3b Frequency LERF Increase (from 3 per 10 years to 1 per 15 years) 3 per 10 year 1 per 10 year 1 per 15 years External Events 7.36E-08 2.45E-07 3.68E-07 2.94E-07 Internal Events 5.43E-09 1.81E-08 2.71E-08 2.17E-08 Combined 7.90E-08 2.63E-07 3.95E-07 3.16E-07 The LERF is within Region II of R.G. 1.174, so the total LERF is estimated by summing the LERF from all contributors, including internal events, fire, and the estimated seismic LERF, as shown below:

LERFIE = 2.69E-07 LERFFire = 6.98E-06 LERFSeismic = 6.24E-07 LERFother = 1.02E-07 Total = 8.29E-06.

The total LERF is the sum of the LERF from internal events, fire, and seismic, and total LERF from internal events, fire, and seismic, and is 8.29E-06, which is below the 1E-05 Region II boundary.

As discussed above, this assessment assumed that significant fires (fires that impact more than one division, hot gas layer fires, alternate shutdown fires) would not occur with the two normally open containment isolation valves associated with the drywell floor drain containment penetration failing to isolate. Given the uncertainty and that some portion of fires may cause a core damage event while failing the containment isolation signal but not failing air or power to the valve, examination of the penetration was performed to determine if additional margin existed that would affect modeling that the failure of the valves to isolate results in a large early release.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 57 of 69 According to JAF Containment Analysis [Reference 31], the drywell floor drain penetration is a 4 diameter pipe penetration and the selected penetration diameter criteria for a failure to isolate that would result in a large early release was chosen to be 2, thus, failure of the valves to isolate given core damage is considered a large early release in the internal events model. However, this may be conservative, so additional evaluation is performed to provide assurance that the failure of the penetration to isolate would not constitute a large early release considering an additional factor, which is evacuation timing. To this end, the containment leakage rate is assessed.

An explicit calculation of the wt%/day leakage versus penetration size was developed in Reference 31, which includes the following data:

Table J3.1-1 [Reference 31] - Containment Leak Size Vs. Containment Leak Rate Diameter (in)

Area (in^2)

Leak rate at design pressure (wt%/day) 0.25 0.05 0.5 0.34 0.09 1

0.5 0.2 2.4 1.25 1.2 14.4 2

3.1 31 3.4 9.1 100 The data in Table J3.1-1 was entered in an MS Excel workbook and a plot of diameter versus leak rate was created. The plot is shown below:

A trend line was added to fit a function to the data. The trend line has the following equation:

Containment Leakage Rate = 9.3447*(diameter)2 - 2.8206*(diameter) + 1.1876 y = 9.3447x2 - 2.8206x + 1.1876 R² = 0.9988 0

20 40 60 80 100 120 0

0.5 1

1.5 2

2.5 3

3.5 4

Containment Leak Rate Penetration Diameter Diameter vs. Leakage

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 58 of 69 The R2 parameter equals 0.9988, indicating a reasonable function fit. The equation was used to calculate the leak rate for a containment at design pressure in terms of the weight percent per day (wt%/day). Using this equation and diameter size of 4 for the drywell floor drain penetration, the equation yields a wt%/day value of ~140%/day. The equation, when fitted to the data in table J3.1-1 shows that some data points may be underestimated compared to the table values, however, the estimated leak rate value is below the wt%/day estimate in section 3.5 of Reference 24, which concludes that a wt%/day rate of 600%/day, when occurring over an assumed 4-hour period prior to population evacuation, would be needed for a large early release to occur. The JAF Level 2 analysis [Reference 31] places an upper limit of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for population evacuation.

Converting the 600%/day for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> criteria to a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> criteria yields a leakage rate of 400%/day. The estimated leak value is sufficiently below the 400%/day estimate to reasonably conclude that failure of the 4 penetration to isolate on a significant fire would not constitute a large early release, and excluding the failure of the containment isolation valves for the penetration from contributing to the updated fire LERF is a reasonable assumption given the occurrence of evacuation.

APLA RAI-1b The Seismic Analysis section of the LAR states that In the IPEEE, JAF performed a Seismic Margin Assessment (SMA), which is a deterministic and conservative evaluation that does not calculate risk on a probabilistic basis. This section does not cite the use of the results from Gl-199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment. It implies that there was no seismic quantification or sensitivity analysis performed for the application. However, Section 4.0 of to the LAR indicates that a sensitivity analysis was performed based on GI-199.

b. Provide additional details in the Seismic Analysis section of the LAR to clarify that an estimate from GI-199 based on the USGS seismic hazard curves was used for quantification (bounding CDF = 6.1E-6/y from Weakest Link Model stated in GI-199) performed in Attachment 4 of the LAR.

Response

Section A.2.2 has been revised to state the following: In the IPEEE, FitzPatrick performed a SMA, which is a deterministic evaluation that does not calculate risk on a probabilistic basis.

The conclusions reached in 2010 by GI-199 [Reference 36] are used for estimating Seismic CDF in this calculation.

APLA RAI-2 Section 1.0 of Attachment 4 to the LAR states that the purpose of this analysis is to provide a risk assessment of permanently extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) to fifteen years. The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for the James A.

FitzPatrick Nuclear Power Plant (JAF). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1].

Section 2.0 of Attachment 4 to the LAR states that the basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, September 1995, provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 59 of 69 assessment of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. This section does not contain the basis for extending the Type A Primary Containment ILRT interval to 15 years.

Provide the basis for extending the ILRT to a 15-year test interval per NEI 94-01, Rev. 3-A.

Response

NEI 94-01 Rev. 3-A supports using EPRI Report No. 1009325 (EPRI 1018243) Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, for performing risk impact assessments in support of ILRT extensions per Section 11.2. The basis for the 10-year test interval is provided in Section 11.2 of NEI 94-01, Rev. 0. Section 2.0 of this calculation has been revised to include the basis from NEI 94-01 Rev. 3A.

APLA RAI-3 Section 5.1.2, Table 5-1 and Table 5-2 of Attachment 4 to the LAR contain the summary of the Internal Events core damage frequency (CDF) and large early release frequency (LERF) results, respectively, for the FitzPatrick PRA Model when quantified using the Level 2 model top gates. For the loss-of-coolant accident "LOCA Outside Containment" internal event, the LERF (1.09E08) is lower than the CDF (1.78E-08). If there is no release mitigation credited for that internal event, the CDF and LERF values would be expected to be the same.

What was credited in the Probabilistic Risk Assessment (PRA) model that contributed to a lower LERF value for the "LOCA Outside Containment" internal event? Verify the credits taken reflect the "as-designed, as-built, as-operated" plant configurations per RG 1200, Rev. 2.

Response

The value for the LOCA Outside Containment event given in Tables 5-1 was incorrect. Table 5-1 of Attachment 4 to the LAR has been updated with the correct value of 1.09E-08 for both CDF and LERF. This error has no impact on the results of this analysis.

APLA RAI-4 Section 4.2.2 of EPRI TR-1009325, Rev. 2-A states that "The most relevant plant-specific information should be used to develop population dose information. The order of preference shall be plant-specific best estimate, Severe Accident Mitigation Alternative (SAMA) for license renewal, and scaling of a reference plant population dose."

Section 5.2.2 of Attachment 4 to the LAR states that the population doses were calculated using the methodology of scaling Peach Bottom population doses to JAF and the scaling produced more conservative population dose values. JAF's plant-specific data is not used in calculating the population doses, while more recent plant-specific data exists.

Provide justification of (1) why plant-specific data is not used and (2) why using Peach Bottom's data would produce a more conservative value. Is the data of the two sites sufficiently comparable to render this scaling approach based on power level, leakage, and population conservative?

Response

The plant-specific dose value for no containment failure provided in Reference 19 (the SAMA) was not used because it was not calculated using the maximally allowable leakage of 1.0La.

Therefore, the methodology of scaling the Peach Bottom population doses was used to provide a more conservative value. Section 5.2.2 of Attachment 4 to the LAR has been revised to include this discussion.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 60 of 69 The scaling of Peach Bottom dose is applicable as both plants are BWR-4s with Mark I containments and have been operating for a similar amount of time, and the power level, leakage, and population for the Peach Bottom assessment are similar enough to render the scaling approach reasonable.

APLA RAI-5 In Section 5.2.2 of Attachment 4 to the LAR, Table 5-9 lists the population dose for Class 7a, "Severe accident phenomena induced failure (early)" is "1.28E+06 person-rem," while the population dose for Class 7b, "Severe accident phenomena induced failure (late)" is "6.81E+05 person rem." When used in conjunction with the release class frequencies for Class 7a and Class 7b in Table 5-10, the annual population dose resulting from late containment failure (1.01 person-rem per year) is higher than that resulting from early containment failure (0.687 person-rem per year).

In NUREG-1437, "Generic Environment Impact Statement for License Renewal of Nuclear Plants", Table 5-4 and Table G-2 show that the annual population doses due to early containment failure and late containment failure are 0.87 person-rem per year and 0.76 person rem per year, respectively. That is, the value for early containment failure exceeds that for late containment failure-a trend opposite to the one in the LAR.

Explain the differences between the NUREG-1437 and the LAR results; provide references for supporting assumptions and calculations.

Response

According to Table 5-4 of NUREG-1437, Supplement 31, the Late Containment Failure population dose exceeds that of Early Containment Failure, 0.87 and 0.76, respectively. Shown below is Table 5-4 of NUREG-1437, Supplement 31. This agrees with Table 5-9 of Section 5.2.2 of Attachment 4 to the LAR, which lists a population dose for Class 7a, Severe accident phenomena induced failure (early) as 1.28E+06 person rem, and a population dose for Class 7b, Severe accident phenomena induced failure (late) as 6.81E+05 person rem.

APLA RAI-6 In Section 5.3.1.of Attachment 4 to the LAR, the fire LERF result was calculated based on the fire CDF (2.12E-5) taken from the original FitzPatrick IPEEE, "James A FitzPatrick Nuclear Power Plant Individual Plant Examination of External Events", which was published in June,

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 61 of 69 1996. However, an updated fire CDF (3.42E-5) that is higher than the one listed in the original IPEEE is available in Section 2.1. 7 of "James A FitzPatrick Nuclear Power Plant Review of FitzPatrick Individual Plant Examination of External Events (IPEEE) Submittal (TAC NO.

M83622)" dated September 21, 2000 ADAMS Accession No. ML003743771). Using a smaller fire CDF value from the original IPEEE to obtain the fire LERF produced a less conservative fire LERF value in the LAR.

Subsequent to resolving the question in RAI 1.a above regarding the conservatism when using the FitzPatrick Fire IPEEE risk, confirm that the total change in LERF is still within the "very small" and "small" risk increase regions of Regulatory Guide 1.174. That is, perform the calculations in Section 5.3 using the updated fire CDF value and correct all subsequent calculations in the LAR that are affected by using the non-conservative old fire CDF value (2.12E-5).

Response

This question is addressed in the response to RAI 1a.i and 1a.ii. The response considered the reason for the higher CDF value reported in Section 2.1.7 of James A. FitzPatrick Nuclear Power Plant - Review of FitzPatrick Individual Plant Examination of External Events (IPEEE)

Submittal (TAC NO. M83622) dated September 21, 2000 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML003743771) and adjusted the fire PRA results to account for the difference. The calculation results in Attachment 4 to the LAR have been updated; the results show that the total change in LERF is still within the small risk increase region of Regulatory Guide 1.174.

APLA RAI-7 In Table 5-21, Section 5.3.2 of Attachment 4 to the LAR, the dose rate for dose rate increase from 3-per-10 years to 1-per-10 years is 3.76E-03; the dose rate for dose rate increase from 3-per-10 years to 1-per-15 years is 6.45E-03; the dose rate for dose rate increase from 1-per-10 years to 1-per-15 years is 2.69E-03. However, based on Table 5-13, these three increases should be 5.08E-3, 8.71 E-3 and 3.63E-3, respectively, which are higher than those listed in Table 5-21.

Correct the dose rates in Table 5-21 and all affected subsequent calculations in the LAR based on the correct dose rates listed in Table 5-13 to ensure the affected calculations do not change the final results of this LAR.

Response

Table 5-13 provided the total dose increase, which included the increases in Classes 3a and 3b as and the decrease in Class 1. Table 5-21 only provided the dose increase of Class 3b. Table 5-21 has been revised to provide the total dose increase for consistency. The three dose rate increases now match the values given in the RAI (5.08E-03, 8.71E-03 and 3.63E-03).

APLA RAI-8 Section 5.3.3 of Attachment 4 to the LAR states that "Taking the baseline analysis and using the values provided in Tables 5-10 and 5-10 for the expert elicitation sensitivity yields the results in Table 5-23."

Confirm that the 2nd "5-10" in the above sentence is a typo and it should be replaced by "5-22".

Response

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 62 of 69 The second 5-10 is a typo, and should actually be 5-11. Table 5-10 and 5-11 are inputs into the Expert Elicitation sensitivity, which uses the Expert Elicitation Mean Probability of Occurrence as opposed to the Jeffreys Non-Informative Prior used by the baseline analysis.

APLA RAI-9 In Table 6-1, Section 6.0 of Attachment 4 to the LAR lists the dose values for Class 1, 3a and 3b are 9.15E+02, 9.15E+03 and 9.15E+04, respectively. In Table 5-10, Section 5.2.3 of to the LAR, the dose values for Class 1, 3a and 3b are 2.97E+03, 2.97E+04 and 2.97E+05, respectively.

The values in Table 6-1 and Table 5-10 are inconsistent. Justify the lower dose values used in Table 6-1, Section 6.0 of Attachment 4 to the LAR; or if the dose values in Table 5-10 are correct, replace Table 6-10 dose values with the values from Table 5-10 and recalculate all subsequent results as necessary.

Response

The dose values given in Table 6-1 for Classes 1, 3a, and 3b were incorrect, though the correct values were used in the analysis. Table 6-1 has been updated with the proper values of 2.97E+03, 2.97E+04, 2.97E+05 for Classes 1, 3a, and 3b, respectively. The dose values given in Tables 5-10 and 6-1 are now consistent.

APLA RAI-10a Address the following in the F&Os section in Table A-1 of Attachment 1, "PRA Technical Adequacy" to the LAR:

a. Finding (Page 43 of 53 in the LAR, Attachment 4): "Appendix H4 states that when applying the dependency model, the dependency was generally applied to the Human Error Probability (HEP) with the higher HEP (versus assigning the dependency to the action which occurs later in time). This does not conform to the established calculation method. In general, it is more appropriate to choose the event that would occur first in the accident sequence... "

To ensure the final CDF and LERF results remain conservative and are not affected by some inadequately low joint HEP values, JAF is expected to follow the requirement of HR-G7 listed in Table 2-2.5-8(g) in ASME/ANS RA-Sa-2009 to assess the HEP dependencies in the cutsets that contain multiple HEPs. Although NRC has not endorsed any standards regarding the floor value of a joint HEP, Section 2.4.3.5 of NUREG-1792, "Good Practices for Implementing Human Reliability Analysis (HRA)" states that "The total combined probability of all the HFEs in the same accident sequence/cut set should not be less than a justified value. It is suggested that the value not be below -0.00001 since it is typically hard to defend that other dependent failure modes that are not usually treated (e.g., random events such as even a heart attack) cannot occur. Depending on the independent HFE values, the combined probability may need to be higher." EPRI TR-1021081, "Establishing Minimum Acceptable Values for Probabilities of Human Failure Events," provides a basis for extending this lowest limit by an order of magnitude to 1 E-6, which may be characteristic only for internal events.

The status of this finding is still open because "the new HRA guidance has not yet been incorporated into the JAF PRA model." The LAR states that this finding is "not significant", since "the assigned dependencies are often conservative because they are based on the overall HEP value (whereas dependency between the execution

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 63 of 69 portions of the HEPs can frequently be justified as being low or zero). Re-analysis of the combined HEPs to address this Fact and Observation (F&O) resulted in a CDF increase of 1.73% and a LERF increase of 0.02%... "

Regarding the re-analysis of the combined HEPs, was there any joint HEP floor value applied to the cutsets that contain multiple HEPs? If so, and if the floor value is significantly less than the value stated in NUREG-1792 or EPRI TR-1021081, provide justification.

If there is no floor value applied to those cutsets that contain multiple HEPs, provide justification and a sensitivity evaluation using a joint HEP floor value of 1E-6 to show that these low joint HEP values have no significant impact on the overall results stated in the LAR.

Response

The re-analysis of the combined HEPs did not apply a joint HEP floor value to the cutsets.

Although specific justification could be provided, sensitivity studies using a joint HEP floor value of 1E-05 and 1E-06 have been performed to demonstrate that using a minimum joint HEP floor does not affect the conclusions of the ILRT extension risk analysis. The following tables show the results of the risk application when using the minimum joint HEP floors of 1E-05 and 1E-06.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 64 of 69 Sensitivity study using minimum joint HEP floor of 1.0E-05:

RAI 10b - ILRT Extension Summary Class Dose (person-rem)

Base Case 3 in 10 Years Extend to 1 in 10 Years Extend to 1 in 15 Years Frequency Person-Rem/Year Frequency Person-Rem/Year Frequency Person-Rem/Year 1

2.97E+03 5.07E-07 1.51E-03 3.24E-07 9.63E-04 1.94E-07 5.75E-04 2

6.93E+05 9.37E-09 6.49E-03 9.37E-09 6.49E-03 9.37E-09 6.49E-03 3a 2.97E+04 6.28E-08 1.86E-03 2.09E-07 6.22E-03 3.14E-07 9.32E-03 3b 2.97E+05 1.56E-08 4.64E-03 5.21E-08 1.55E-02 7.81E-08 2.32E-02 6

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7a 1.28E+06 5.39E-07 6.87E-01 5.39E-07 6.87E-01 5.39E-07 6.87E-01 7b 6.81E+05 5.93E-06 4.04E+00 5.93E-06 4.04E+00 5.93E-06 4.04E+00 8

4.55E+05 2.05E-08 9.33E-03 2.05E-08 9.33E-03 2.05E-08 9.33E-03 Total 7.08E-06 4.75E+00 7.08E-06 4.76E+00 7.08E-06 4.77E+00 ILRT Dose Rate from 3a and 3b Total Dose Rate From 3 Years N/A 1.46E-02 2.51E-02 From 10 Years N/A N/A 1.05E-03

%Dose Rate From 3 Years N/A 0.308%

0.529%

From 10 Years N/A N/A 0.220%

3b Frequency (LERF)

LERF From 3 Years N/A 3.65E-08 6.25E-08 From 10 Years N/A N/A 2.60E-08 CCFP %

CCFP%

From 3 Years N/A 0.515%

0.883%

From 10 Years N/A N/A 0.368%

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 65 of 69 Sensitivity study using minimum joint HEP floor of 1.0E-06:

RAI 10b - ILRT Extension Summary Class Dose (person-rem)

Base Case 3 in 10 Years Extend to 1 in 10 Years Extend to 1 in 15 Years Frequency Person-Rem/Year Frequency Person-Rem/Year Frequency Person-Rem/Year 1

2.97E+03 5.56E-07 1.65E-03 4.87E-07 1.45E-03 4.37E-07 1.30E-03 2

6.93E+05 3.61E-09 2.50E-03 3.61E-09 2.50E-03 3.61E-09 2.50E-03 3a 2.97E+04 2.38E-08 7.07E-04 7.93E-08 2.36E-03 1.19E-07 3.53E-03 3b 2.97E+05 5.92E-09 1.76E-03 1.97E-08 5.86E-03 2.96E-08 8.79E-03 6

0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7a 1.28E+06 5.39E-07 6.87E-01 5.39E-07 6.87E-01 5.39E-07 6.87E-01 7b 6.81E+05 1.70E-06 1.16E+00 1.70E-06 1.16E+00 1.70E-06 1.16E+00 8

4.55E+05 2.05E-08 9.33E-03 2.05E-08 9.33E-03 2.05E-08 9.33E-03 Total 2.85E-06 1.86E+00 2.85E-06 1.87E+00 2.85E-06 1.87E+00 ILRT Dose Rate from 3a and 3b Total Dose Rate From 3 Years N/A 5.55E-03 9.51E-03 From 10 Years N/A N/A 3.96E-03

%Dose Rate From 3 Years N/A 0.298%

0.511%

From 10 Years N/A N/A 0.212%

3b Frequency (LERF)

LERF From 3 Years N/A 1.38E-08 2.37E-08 From 10 Years N/A N/A 9.87E-09 CCFP %

CCFP%

From 3 Years N/A 0.485%

0.831%

From 10 Years N/A N/A 0.346%

APLA RAI-10b

b. Finding (Page 44 of 53 in the LAR, Attachment 4): The JAF PRA determined a point estimate for damage and documented an analysis for parametric uncertainty.

However, the state of knowledge correlation was not fully accounted for due to the manner in which the JAF basic event data base is constructed."

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 66 of 69 The SR description for this finding did not include LERF. However, the QU-A3 requirement listed in Table A-2 of RG 1.200 states that "The state-of-knowledge correlation should be accounted for all event probabilities. Left to the analyst to determine the extent of the events to be correlated. Need to also acknowledge LERF quantification."

Provide the quantitative results from the sensitivity evaluation performed at the time of the RG 1.200, Rev. 2 peer review-the results should include the effect on both CDF and LERF due to parametric/state of knowledge uncertainty.

Response

Since the sensitivity evaluation performed at the time of the RG 1.200, Rev. 2 peer review was not available, a new sensitivity was performed.

The type codes in the JAF PRA are based on the system, equipment type, and failure mode; Bayesian updating of failure data leads to different failure rates for a given equipment type and failure mode based on its system. A sensitivity is performed by re-defining the type codes based solely on equipment type and failure mode to allow a state-of-knowledge correlation. When multiple failure rates exist for different systems within a type code, a bounding type code failure rate is assigned for the sensitivity. Even with this conservative failure rate assignment, the CDF only increases from 2.40E-6 to 2.67E-6; the LERF only increases from 2.69E-6 to 2.75E-5.

The parametric uncertainty was run on UNCERT using a Monte Carlo method with 10,000 samples. This number of samples was selected to ensure stable results with robust estimates of key percentiles. The figures below show the cumulative probability, probability density function, and key statistics of the CDF and LERF, respectively. The results of the parametric uncertainty show that the correlated mean is reasonably close (~1.9% for CDF and ~1.1% for LERF) to the point estimates, so using the point estimates in the ILRT extension PRA analysis is reasonable.

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 67 of 69 UNCERT Results for Fire-Induced CDF UNCERT Results for Fire-Induced LERF

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 68 of 69 APLA RAI-10c

c. Finding (Page 45 of 53 in the LAR, Attachment 4): "The level 2 model uses point estimates for success branches while the failure logic is modeled for the failure branch."

The LAR states that "no impact on the application is expected" because the "use of point estimates is reasonable for top events that include phenomena, and system related success probabilities are approximately equal to 1.0. Provide the values of all success probabilities to show that they are sufficiently close to 1.0 and have minimal impact on the quantification.

Provide the values of all success probabilities to show that they are sufficiently close to 1.0 and have minimal impact on the quantification.

Response

The success probabilities in the Level 2 fault tree model are modeled for both phenomenological failures and system failures. The modeling of the success probabilities was performed by taking the complement of the failure probability for the modeled function. That is, the failure probability was quantified from the appropriate model gate and the complement was calculated and assigned to a single basic event in the Level 2 model fault tree. The complement of the failure probabilities are approximately 1.0, with the definition of approximately as any probability with value that is greater than 0.9. There are three exceptions to this definition. The first two relate to the probability that the Reactor Pressure Vessel is at low pressure at the time of vessel breach for accident Classes IBL and IVA. These two probabilities are the complement of a gate that contains both phenomenological and hardware failures and the values of the two success probabilities are 0.830 and 0.889. The third success probability that is not greater than 0.9 relates to the probability of successful in-vessel recovery with no vessel breach for a Class IVA accident. The probability is the complement of a gate that contains both phenomenological and hardware failures, and the value of the probability is 0.499. The 0.499 success probability is dominated by a phenomenological failure event that has probability of 0.5.

Although these three probabilities are not greater than 0.9, the failure probability is driven by phenomenological basic events, and thus the use of the complement value is reasonable. For the success probabilities that are less than 0.9, quantifying only the system portion of the failure logic gate and taking the complement of the system failure probability shows system success to be ~.999, which is greater than 0.9 and considered approximately 1.0. Therefore, the total LERF used in the ILRT Extension Analysis is not affected by the modeling of success probabilities that are not approximately 1.0, and there is no impact on the change in risk estimate of the ILRT extension analysis and no impact on the conclusion of the ILRT extension risk analysis.

APLA RAI-10d

d. Finding (Page 46 of 53 in the LAR, Attachment 4): "Spray-induced and submergence induced failures appear to have been addressed in the analysis. No documentation of a systematic assessment of the effects of jet impingement, pipe whip, humidity, temperature, etc., on structure, system or components (SSCs) could be identified. No evaluation of the specific equipment evaluated in the PRA compared to equipment considered in the design analyses, e.g., EQ lists was documented. Since PRA can credit non-safety-related equipment, relying on design basis evaluations to dismiss these dynamic effects may credit equipment that cannot withstand the effects considered in the design analysis. In addition, failure in a system containing high temperature fluid can

33010-CALC-01 Evaluation of Risk Significance of Permanent ILRT Extension Revision 2 Page 69 of 69 actuate fire systems and impact additional equipment. Also, the PRA models may evaluate breaks beyond those of the design basis."

The status of this finding is open and the licensee states that "The only requirement to meet Capability Category II (CC-II) is to document that those mechanisms are not included in the scope."

However, Table A-3 in Rev. 2 of RG 1.200, IFSN-A6 requires that "For Cat II, it is not acceptable to just note that a flood-induced failure mechanism is not included in the scope of the internal flooding analysis. Some level of assessment is required." Also, "For the SSCs identified in IFSN-A5, IDENTIFY the susceptibility of each SSC in a flood area to flood-induced failure mechanisms. INCLUDE failure by submergence and spray in the identification process. ASSESS qualitatively the impact of flood-induced mechanisms that are not formally addressed (e.g., using the mechanisms listed under Capability Category Ill of this requirement), by using conservative assumptions."

Provide the results from the analysis related to IFSN-A6 to show that CC-II requirements are met; or explain how meeting only CC-I requirements does not have impact on the final conclusions of this application.

Response

Capability Category I is sufficient to provide an estimate of the total Internal Flooding CDF and LERF as used in the ILRT extension PRA analysis. Meeting Capability Category II could provide additional qualitative discussion regarding impact of flood-induced mechanisms that are not formally addressed using the mechanisms in Capability Category III, but would not impact the total CDF and LERF used in the ILRT extension analysis and therefore would not affect the final conclusions of this application. The disposition of F&O IFSN-A6 has been revised to include the clarification from R.G. 1.200 Rev 2 and to state that Capability Category I is sufficient for the ILRT extension analysis.

JAFP-16-0170 Revised Pages Page 17 Page 22 Page 23 New Page 23a Page 76

ATTACHMENT 1

EVALUATION OF THE PROPOSED CHANGE

17

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

NEI 94-01 Revision 3-A supports using EPRI Report No. 1009325 Revision 2-A (EPRI 1018243), Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, for performing risk impact assessments in support of ILRT extensions (Reference 20). The guidance provided in Appendix H of EPRI Report No. 1009325 Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI TR-104285.

This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

In the SER issued by NRC letter dated June 25, 2008 (Reference 9), the NRC concluded that the methodology in EPRI TR-1009325, Revision 2, was acceptable for referencing by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.0 of the Safety Evaluation (SE). Table 3.3.1-1 addresses each of the four limitations and conditions for the use of EPRI 1009325, Revision 2.

Table 3.3.1-1, EPRI Report No. 1009325 Revision 2 Limitations and Conditions Limitation/Condition (From Section 4.2 of SE)

JAF Response

1. The licensee submits documentation indicating that the technical adequacy of their PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension.

JAF PRA technical adequacy is addressed in Section 3.3.2 of this submittal and, Evaluation of Risk Significance of Permanent ILRT Extension, Appendix A, Attachment 1, which addresses the Technical Adequacy of the PRA modeling.

ATTACHMENT 1

EVALUATION OF THE PROPOSED CHANGE

22

Fire Analysis The JAF IPEEE fire analysis was performed using EPRIs Fire PRA Implementation Guide (Reference 28). The EPRI Fire-Induced Vulnerability Evaluation method was used for the initial screening, for treatment of transient combustibles, and as the source of fire frequency data (Reference 22).

The fire analysis was revised after the original IPEEE submittal in response to NRC requests for additional information (RAIs) regarding fire-modeling progression, developed during their review of the IPEEE. The updated results are reflected in NUREG-1742, Perspectives Gained from the Individual Plant Examination of External Events (IPEEE) Program (Reference 31) for JAF. In addition, as noted in that NUREG, a number of plant and procedural changes (including strict limitations on storage and use of combustible and flammable material in plant areas) were made as a result of the IPEEE fire analysis. The impact of these enhancements is not reflected in the IPEEE fire results.

Other changes to the plant configuration, procedures and equipment performance have also taken place since completion of the IPEEE. These changes would tend to reduce the overall CDF as well as the fire risk contribution found in the IPEEE. The significant reduction in the internal events CDF since the original JAF IPEEE submittal bears this out. These changes include the following:

Service, instrument, and breathing air compressors were replaced.

Operators are directed to maximize CRD flow in certain accident sequences.

SRV Electric Lift mod to install an SRV alternate actuation system.

A new procedure (EP-10) directs operators to align the fire protection system to the tube side of the RHR heat exchanger in loss of containment heat removal accident sequences.

Revised station blackout procedures to explicitly address bus recovery.

Provision of a back-up battery charger that can be aligned to either station battery.

Proceduralized RCIC operation without DC power.

Proceduralized starting EDG without DC power, as well as field flashing without station batteries.

The JAF IPEEE fire risk model has been updated since its original issuance to include new conditional core damage probabilities via a failure probability data update to the basic events in the original cutsets. Further update to the fire risk model, using updated ignition frequencies, circuit failure probabilities, and with re-analysis of containment bypass scenarios has also been performed to support the external events risk study in Attachment 4, Section 5.3.1 of this submittal.

Seismic Analysis In the IPEEE, JAF performed a Seismic Margin Assessment (SMA), which is a deterministic evaluation that does not calculate risk on a probabilistic basis. The conclusions reached in 2010 by Generic Issue (GI) 199 (Reference 34) are used for estimating Seismic CDF in this calculation.

ATTACHMENT 1

EVALUATION OF THE PROPOSED CHANGE

23

High Winds, Floods, and Other External Hazards (HFO)

In addition to internal fires and seismic events, the JAF IPEEE analysis of HFO external hazards was accomplished by reviewing the plant environs against regulatory requirements regarding these hazards. Since JAF was designed (with construction started) prior to the issuance of the 1975 Standard Review Plan (SRP) criteria, JAF performed a plant hazard and design information review for conformance with the SRP criteria. For HFO events that were not screened out by compliance with the 1975 SRP criteria, additional analyses were performed to determine whether or not the hazard frequency was acceptably low. Based on those analyses, these hazards were determined in the JAF IPEEE to be negligible contributors to overall plant risk.

Technical Adequacy Summary The technical adequacy of the JAF PRA is consistent with the requirements of Regulatory Guide 1.200 (Reference 4) as is relevant to this ILRT interval extension, as detailed in Attachment 4 of this submittal, Appendix A Attachment 1.

3.3.3 Summary of Plant-Specific Risk Assessment Results The findings of the JAF Risk Assessment contained in Attachment 4 confirm the general findings of previous studies that the risk impact associated with extending the ILRT interval from three in ten years to one in 15 years is small. The JAF plant-specific results for extending ILRT interval from the current 10 years to 15 years are summarized below:

Based on the results from Attachment 4, Section 5.2, Analysis, and the sensitivity calculations presented in Attachment 4, Section 5.3 Sensitivities, the following conclusions regarding the assessment of the plant risk associated with permanently extending the Type A ILRT test frequency to fifteen years are:

Regulatory Guide 1.174 (Reference 3) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.0E-06/year and increases in LERF less than 1.0E-07/year. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 2.17E-8/year using the EPRI guidance; this value increases slightly to 2.20E-8/year if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174.

When external event risk is included, the increase in LERF resulting from a change in the Type A ILRT test interval from 3 in 10 years to 1 in 15 years is estimated as 3.16E-7/year using the EPRI guidance, and baseline LERF is 8.29E-6/year. As such, the estimated change in LERF is determined to be small using the acceptance guidelines of Regulatory Guide 1.174. As

ATTACHMENT 1 EVALUATION OF THE PROPOSED CHANGE 23a

§ The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.0087 person-rem/year. EPRI Report No. 1009325, Revision 2-A (Reference 20) states that a very small discussed in Sections 5.1.3 and 5.3.1, the EPRI methodology used to estimate the increase in LERF is conservative. Therefore, even though the increase in LERF is near the Regulatory Guide 1.174 threshold, there is additional margin because conservative methodology was used.

ATTACHMENT 1 EVALUATION OF THE PROPOSED CHANGE 76 30.

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

31.

U.S. Nuclear Regulatory Commission, NUREG-1742 Perspectives Gained From the Individual Plant Examination of External Events (IPEEE) Program, Volume 1

& 2, Final Report, April 2002.

32.

Letter from Mr. C. H. Cruse (Constellation Nuclear, Calvert Cliffs Nuclear Power Plant) to U.S. Nuclear Regulatory Commission, Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Accession Number ML020920100, March 27, 2002.

33.

Engineering Report No. JAF-NE-09-00001, Revision 0, Fitzpatrick Probabilistic Safety Assessment (PSA), Rev. 4, August 2009.

Generic Issue 199 (GI-199), ML100270582, September 2010, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants: Safety/Risk Assessment.

34.