View source for RNP-RA/15-0090, Proposed Amendment to Technical Specification 5.5.16 for the Adoption of Option B of 10 CFR 50, Appendix J for Type B and Type C Testing and the Permanent Change in 10 CFR 50, Appendix J, Integrated Leak Rate Test Interval And..

{{Adams
| number = ML15323A085
| issue date = 11/19/2015
| title = Proposed Amendment to Technical Specification 5.5.16 for the Adoption of Option B of 10 CFR 50, Appendix J for Type B and Type C Testing and the Permanent Change in 10 CFR 50, Appendix J, Integrated Leak Rate Test Interval And..
| author name = Glover R
| author affiliation = Duke Energy Progress, Inc
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000261
| license number = DPR-023
| contact person = 
| case reference number = RNP-RA/15-0090
| document type = Letter type:RNP, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 172
| project = 
| stage = Other
}}

=Text=
{{#Wiki_filter:R. Michael Glover
  /_~ DUKE                                                                          H. B. Robinson Steam Electric Plant Unit 2 ENERGY'&deg;                                                                       Site Vice President Duke Energy Progress 3581 West Entrance Road Hartsville, SC 29550 0 . 843 857 1704 F: 843 857 1319 Mike.Glawr@*.lke-ener!1f.can 10 CFR 50.90 Serial: R.Ne-RA/15-0090 NUV I 9 Z015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 I RENEWED LICENSE NO. DPR-23 Proposed Amendment to Technical Specification 5.5.16 for the Adoption of Option B of 1 O CFR 50, Appendix J for Type B and Type C Testing and the Permanent Change in 1 OCFR 50, Appendix J, Integrated Leak Rate Test Interval and Type C Leak Rate Testing Frequency

==Dear Sir/Madam:==

In accordance with the provisions of 10 CFR 50.90 Duke Energy Progress, Inc. is submitting a request for an amendment to the technical specifications (TS) for H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2). The proposed change is a request to revise TS 5.5.16, Containment Leakage Rate Testing Program, TS 3.6.1, Containment, and TS 3.6.2, Containment Air Lock, for HBRSEP2, to allow the following:
Increase in the existing integrated leak rate test (ILRT) program test interval from 10 years to 15 years, Adopt 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, for the performance-based testing of Types B and C tested components in accordance with the guidance of Technical Specification Task Force (TSTF)-52, Implement 10 CFR 50, Appendix J, Option B (Reference 11 of the Enclosure),
Allow an extension to the 120-month frequency currently permitted by Option B for Type B leakage rate testing, Allow an extension from the 60-month frequency currently permitted by Option B to a 75-month frequency for Type C leakage rate testing,

U. S. Nuclear Regulatory Commission Serial: RNP-RA/15-0090 Page 2 The proposed change would also adopt a more conservative grace interval of 9 months, for Type Band Type C tests in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, revision 3-A.
The Enclosure provides a description and basis of the proposed change, a No Significant Hazards Consideration Determination, and an Environmental Analysis. Attachment 1 provides the existing TS pages marked up to show the proposed changes. Attachment 2 provides revised (clean) TS pages that reflect the proposed change. Attachment 3 provides the existing TS bases pages marked up to show the proposed changes, and are provided for information only. Attachment 4 provides an evaluation of risk significance of permanent ILRT extension. provides E-C AUGMENT plan tables for the second ten-year interval. Attachment 6 provides the containment liner detail at moisture barrier.
HBRSEP2 requests approval of the proposed license amendment by November 30, 2016, with the amendment being implemented within 120 days of issuance.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated South Carolina Official.
Please address any comments or questions regarding this matter to Mr. Richard Hightower, Manager - Nuclear Regulatory Affairs at (843) 857-1329.
I declare under penalty of perjury that the foregoing is true and correct. Executed on N0\1 ,~..:.-   ,,  I 2015.
Sincerely, R. Michael Glover Site Vice President RMG/jmw Enclosure Attachments
: 1. Proposed Technical Specification Changes
: 2. Revised Technical Specification Pages
: 3. Proposed Technical Specifications Bases Changes
: 4. Evaluation of Risk Significance of Permanent ILRT Extension
: 5. E-C AUGMENT Plan Tables for the Fifth Ten-Year Interval
: 6. Containment Liner Detail at Moisture Barrier cc: Region Administrator, NRC, Region II Ms. Martha C. Barillas, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP2 Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)

Enclosure Page 1 of 60 Evaluation of the Proposed Change

==Subject:==
Proposed Amendment to Technical Specification 5.5.16 for the Adoption of Option B of 10 CFR 50, Appendix J for Type B and Type C Testing and the Permanent Change in 10 CFR 50, Appendix J, Integrated Leakage Rate Test Interval and Type C Leak Rate Testing Frequency 1.0   

==SUMMARY==
DESCRIPTION 2.0   DETAILED DESCRIPTION

==3.0   TECHNICAL EVALUATION==

==4.0   REGULATORY EVALUATION==

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

==5.0   ENVIRONMENTAL CONSIDERATION==

==6.0   REFERENCES==

Enclosure Page 2 of 60 1.0      

==SUMMARY==
DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy Progress requests an amendment to the H. B.
Robinson Steam Electric Plant Unit No. 2 (HBRSEP2) Renewed Facility Operating License (DPR-23) by incorporating the attached proposed change into the HBRSEP2 Technical Specifications (TS). Specifically, the proposed change is a request to revise TS 5.5.16, "Containment Leakage Rate Testing Program," TS 3.6.1, Containment, and TS 3.6.2, Containment Air Lock, for HBRSEP2, to allow the following:
* Increase in the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A and the conditions and limitations specified in NEI 94-01, Revision 2-A.
* Adopt 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, for the performance-based testing of Types B and C tested components in accordance with the guidance of Technical Specification Task Force (TSTF)-52, Implement 10 CFR 50, Appendix J, Option B (Reference 11).
* Adopt an extension of the containment isolation valve leakage testing (Type C) frequency from the 60 months currently permitted by 10 CFR 50, Appendix J, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.
* Adopt the use of ANSI/ANS 56.8-2002, Containment System Leakage Testing Requirements.
* Adopt a more conservative grace interval of 9 months, for Type A, Type B and Type C tests in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A.
The proposed change to the TS contained herein would revise HBRSEP2 TS 5.5.16, by replacing the reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, (Reference 1) and 10 CFR 50, Appendix J, Option A with a reference to NEI topical report NEI 94-01, Revision 3-A (Reference 2), dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A (Reference 3), dated October 2008, as the implementation documents used by HBRSEP2 to implement the performance-based leakage testing program in accordance with 10 CFR 50, Appendix J, Option B. This license amendment request (LAR) also proposes an administrative change to TS 5.5.16 by deleting the information regarding the performance of the next Type A test no later than April 9, 2007, as this has already occurred.
The proposed change to the TS contained herein would revise HBRSEP2 TS SRs 3.6.1.1 and 3.6.2.1, by replacing the reference to 10 CFR 50, Appendix J, Option A with a reference to the Containment Leakage Rate Testing (CLRT) program and incorporate the changes recommended by TSTF-52, Revision 3, as applicable to HBRSEP2.
The associated TS Bases for SR 3.0.2, SR 3.6.1.1 and SR 3.6.1.2 are also being revised to reflect the proposed change removing references to 10 CFR 50, Appendix J, Option A and

Enclosure Page 3 of 60 incorporate the bases changes recommended by TSTF-52 as applicable to HBRSEP2.
2.0     DETAILED DESCRIPTION 2.1     Current Containment Leakage Rate Testing Program This program provides controls for implementation of the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guidelines contained in RG 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception:
: a.      NEI 94 1995, Section 9.3.2: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than April 9, 2007.
Types B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix J, Option A.
2.2     TS Change Description The following changes are requested as part of the implementation of 10 CFR 50, Appendix J, Option B, as described in TSTF 52, Revision 3, Implement 10 CFR 50, Appendix J, Option B, dated March 2000.
Surveillance Requirement (SR) 3.6.1.1, Containment, currently states:
SURVEILLANCE                                 FREQUENCY SR 3.6.1.1     Perform required Type B and C leakage rate           --------NOTE-------
testing except for containment air lock testing, in  SR 3.0.2 is not accordance with 10 CFR 50, Appendix J, Option        applicable A, as modified by approved exemptions.               ------------------------
The leakage rate acceptance criterion is  1.0 La. In accordance However, during the first unit startup following     with 10 CFR 50, testing performed in accordance with 10 CFR 50,      Appendix J, Appendix J, Option A, as modified by approved        Option A, as exemptions, the leakage rate acceptance criterion    modified by is < 0.6 La for the Type B and Type C tests.         approved exemptions TS SR 3.6.1.1 is being revised for consistency with TSTF-52 and the adoption of the guidelines contained in NEI Topical Report, NEI 94-01, Revision 3-A for Type B and Type C local leakage rate testing (LLRT). The proposed change will revise TS SR 3.6.1.1 to state:
SURVEILLANCE                                 FREQUENCY SR 3.6.1.1     Perform required visual examinations and Type B      --------NOTE-------
and C leakage rate testing except for containment    SR 3.0.2 is not air lock testing, in accordance with 10 CFR 50,      applicable

Enclosure Page 4 of 60 Appendix J, Option A, as modified by approved                 ------------------------
exemptions.
The leakage rate acceptance criterion is  1.0 La.            In accordance However, during the first unit startup following              with 10 CFR 50, testing performed in accordance with 10 CFR 50,               Appendix J, Appendix J, Option A, as modified by approved                 Option A, as exemptions, the leakage rate acceptance criterion             modified by is < 0.6 La for the Type B and Type C tests. the              approved Containment Leakage Rate Testing Program.                     exemptions the Containment Leakage Rate Testing Program SR 3.6.1.3, currently states:
SURVEILLANCE                                         FREQUENCY SR 3.6.1.3     Perform required visual examinations and Type A               In accordance leakage rate testing, in accordance with the                  with the Containment Leakage Rate testing Program                      Containment Leakage Rate Testing Program The proposed change will delete TS SR 3.6.1.3 in its entirety as it is now incorporated into SR 3.6.1.1:
SURVEILLANCE                                         FREQUENCY SR 3.6.1.3     Perform required visual examinations and Type A               In accordance leakage rate testing, in accordance with the                  with the Containment Leakage Rate testing Program                      Containment Leakage Rate Testing Program TS 3.6.2, Containment Air Lock, Required Action C.1 is being revised to remove an extra revision bar character. This is an administrative change. Please refer to the markup in .
SR 3.6.2.1, Containment Air Lock, currently states:
SURVEILLANCE                                         FREQUENCY SR 3.6.2.1     ---------------------------NOTES-----------------------------
: 1.         An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage

Enclosure Page 5 of 60 test.
: 2.         Results shall be evaluated against acceptance criteria of SR 3.6.1.1, in accordance with 10 CFR 50, Appendix J,                    --------NOTE-------
Option A, as modified by approved                         SR 3.0.2 is not exemptions.                                               applicable Perform required air lock leakage rate testing in                   In accordance accordance with 10 CFR 50, Appendix J, Option                       with 10 CFR 50, A, as modified by approved exemptions.                              Appendix J, Option A, as modified by approved exemptions The TS SR 3.6.2.1 is being revised for consistency with TSTF-52 and the adoption of the guidelines contained in NEI Topical Report, NEI 94-01, Revision 3-A for Containment Air Lock LLRT. The proposed change will revise TS SR 3.6.2.1 to state, as follows:
SURVEILLANCE                                               FREQUENCY SR 3.6.2.1     ---------------------------NOTES-----------------------------
: 1.         An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
: 2.         Results shall be evaluated against acceptance criteria of applicable to SR 3.6.1.1, in accordance with 10 CFR 50,                    --------NOTE-------
Appendix J, Option A, as modified by                      SR 3.0.2 is not approved exemptions.                                      applicable Perform required air lock leakage rate testing in                   In accordance accordance with 10 CFR 50, Appendix J, Option                       with 10 CFR 50, A,                                                                  Appendix J, as modified by approved exemptions. the                             Option A, as Containment Leakage Rate Testing Program.                           modified by approved exemptions the Containment Leakage Rate Testing Program.
HBRSEP2 TS 5.5.16, "Containment Leakage Rate Testing Program," currently states:

Enclosure Page 6 of 60 This program provides controls for implementation of the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception:
: a.      NEI 94 1995, Section 9.3.2: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than April 9, 2007.
Type B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix J, Option A.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42 psig, which exceeds the peak calculated containment internal pressure for the design basis loss of coolant accident.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of the containment air weight per day.
Leakage rate acceptance criteria are:
: a.      Containment leakage rate acceptance criteria is < 1.0 La. During the first unit 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 Type A tests.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
The proposed changes to HBRSEP2 TS 5.5.16, "Containment Leakage Rate Testing Program" will be (1) the administrative change to delete the performance of the next Type A test no later than April 9, 2007, that was previously approved by the NRC via Amendment No. 193, (2) to include the addition of the references to the guidelines contained in NEI Topical Report, NEI 94-01, Revision 3-A and the conditions and limitations contained in NEI Topical Report, NEI 94-01 Revision 2-A for Type A, Type B and Type C LLRT, and (3) incorporate the updates recommended by TSTF-52. The proposed change will revise TS 5.5.16, as follows, to state:
A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42 psig. The containment design pressure is 42 psig.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of the containment air weight per day.

Enclosure Page 7 of 60 Leakage rate acceptance criteria are:
: a.      Containment leakage rate acceptance criteria is < 1.0 La. During the first unit 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 Type A tests.
: b.      Air lock testing acceptance criteria are:
: 1. Overall air lock leakage rate is < 0.05 La when tested at > Pa.
: 2. For each door, leakage rate is < 0.01 La when pressurized to > 42 psig.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.
Mark-ups of TS 5.5.16; SR 3.6.1.1; SR 3.6.1.3; TS 3.6.2, Required Action C.1; and, SR 3.6.2.1 are provided in Attachment 1. The retyped TS pages are provided in Attachment 2.
A markup of TS Bases for SR 3.0.2, TS 3.6.1, SR 3.6.1.1, SR 3.6.1.3, B 3.6.2, and SR 3.6.2.1 are provided in Attachment 3 for informational purposes only.
This proposed change is requested to extend the performance of the next HBRSEP2 ILRT from the Spring 2017 refueling outage to a subsequent refueling outage no later than Spring 2021. contains the plant specific risk assessment conducted to support this proposed change. This risk assessment followed the guidelines of Nuclear Regulatory Commission (NRC) RG 1.174 (Reference 4) and NRC RG 1.200, Revision 2 (Reference 5). The risk assessment concluded that the increase in risk as a result of this proposed change is considered to be insignificant since it represents a very small change to the HBRSEP2 risk profile.
2.3      Deviations From TSTF-52, Revision 3 TS 1.1, Definitions TSTF-52 deletes La from the definitions in TS 1.1.
Deviation - There is no definition for La in TS 1.1 of the HBRSEP2 TS, so this change is unnecessary.

==3.0      TECHNICAL EVALUATION==

3.1      Justification for the Technical Specification Change

Enclosure Page 8 of 60 The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the TS. 10 CFR 50, Appendix J, also ensures that periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment and the systems and components penetrating primary containment. The limitation on containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant design basis accident. Appendix J identifies three types of required tests: (1) Type A tests, intended to measure the primary containment overall integrated leakage rate; (2) Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage limiting boundaries (other than valves) for primary containment penetrations; and (3) Type C tests, intended to measure containment isolation valve leakage rates. Types B and C tests identify the vast majority of potential containment leakage paths.
Type A tests identify the overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Types B and C testing.
In 1995, 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach.
Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50, Appendix J, refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.
Also, in 1995, RG 1.163 was issued. The RG endorsed NEI 94-01, Revision 0, (Reference 6) with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) test from three tests in 10 years to one test in 10 years. This relaxation was based on an NRC risk assessment contained in NUREG-1493 (Reference 7), and Electric Power Research Institute (EPRI) TR-104285 (Reference 8), both of which showed that the risk increase associated with extending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval, provisions for extending the test interval an additional 15 months were considered in the establishment of the intervals allowed by RG 1.163 and NEI 94-01, but that this "should be used only in cases where refueling schedules have been changed to accommodate other factors."
In 2008, NEI 94-01, Revision 2-A, (Reference 3) was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC Safety Evaluation Report (SER) on NEI 94-01. The NRC SER was included in the front matter of this NEI report. NEI 94-01, Revision 2-A, includes provisions for extending Type A ILRT intervals to up to fifteen years and incorporates the regulatory positions stated in Regulatory Guide 1.163 (September 1995). It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights.

Enclosure Page 9 of 60 Acceptability for referencing by licensees proposing to amend their TS is provided in Section 5.0 of the SER and states the following:
The NRC staff, therefore, finds that this guidance is acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.0 of this SE. In addition, in accordance with the NRC staff's resolution of the comments provided by NEI on the draft SE, the following changes will be made by NEI to the "-A" version of the TR. Therefore, consistent with the language in this final SE:
A. NEI TR 94-01, Revision 2, will be revised in the "-A" version of the report, as discussed in the last paragraph of Section 3.1.2.2, "Extending Type B & C Test Intervals," to the final SE.
B. EPRI Report No. 1009325, Revision 2, will be revised in the "-A" version of the report, to change the population dose acceptance guidelines and the CCFP guidelines. (As stated in Section 4.2 of the final SE Limitation and Condition #2).
In 2012, NEI 94-01, Revision 3-A, was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, and includes provisions for extending Type A ILRT intervals to up to fifteen years. NEI 94-01 has been endorsed by RG 1.163 and NRC SERs dated June 25, 2008 (Reference 9) and June 8, 2012 (Reference 10) as an acceptable methodology for complying with the provisions of Option B to 10 CFR Part 50. The regulatory positions stated in RG 1.163 as modified by NRC SERs dated June 25, 2008, and June 8, 2012, are incorporated in this document. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights. Extensions of Type B and Type C test intervals are allowed based upon completion of two consecutive periodic as-found tests where the results of each test are within a licensees allowable administrative limits.
Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment airlocks) and up to a maximum of 75 months for Type C tests. If a licensee considers extended test intervals of greater than 60 months for Type B or Type C tested components, the review should include the additional considerations of as-found tests, schedule and review as described in NEI 94-01, Revision 3-A, Section 11.3.2.
Acceptability for referencing by licensees proposing to amend their TS is provided in Section 5.0 of the SER and states the following:
The NRC staff, therefore, finds that this guidance, as modified to include two limitations and conditions, is acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing. Any applicant may reference NEI TR 94-01, Revision 3, as modified by this SE and approved by the NRC, in a licensing action to satisfy the requirements of Option B to 10 CFR Part 50, Appendix J. The NRC staff is not required to repeat its review of the matters described in the TR conditioned upon the changes described in this SE (Sections 3 and 4) to be incorporated when the report appears as a reference which was complied with a request for relief, or other related licensing actions.

Enclosure Page 10 of 60 NEI 94-01, Revision 3-A, Section 10.1 concerning the use of grace in the deferral of Type B and Type C LLRTs past intervals of up to 120 months for the recommended surveillance frequency for Type B testing and up to 75 months for Type C testing, states:
Consistent with standard scheduling practices for Technical Specifications Required Surveillances, intervals of up to 120 months for the recommended surveillance frequency for Type B testing and up to 75 months for Type C testing given in this section may be extended by up to 25 percent of the test interval, not to exceed nine months.
Notes: For routine scheduling of tests at intervals over 60 months, refer to the additional requirements of Section 11.3.2.
Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. This provision (nine-month extension) does not apply to valves that are restricted and/or limited to 30-month intervals in Section 10.2 (such as BWR MSIVs) or to valves held to the base interval (30 months) due to unsatisfactory LLRT performance.
3.1.1   Current HBRSEP2 10 CFR 50, Appendix J Requirements Title 10 CFR Part 50, Appendix J, was revised, effective October 26, 1995, to allow licensees to choose containment leakage testing under either Option A, "Prescriptive Requirements," or Option B, "Performance Based Requirements." HBRSEP2 has implemented the requirements of 10 CFR Part 50, Appendix J, Option A for Types B and C testing and Option B for Type A testing. Current TS 5.5.16 requires the following:
This program provides controls for implementation of the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception:
: a.      NEI 94 1995, Section 9.3.2: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than April 9, 2007.
Type B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix J, Option A.
RG 1.163, Section C.1, states that licensees intending to comply with 10 CFR Part 50, Appendix J, Option B, should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01 (Reference 6) rather than using test intervals specified in American National Standards Institute (ANSI)/American Nuclear Society (ANS) 56.8-1994. NEI 94-01, Section 11.0, refers to Section 9, which states that Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per ten years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage was less than 1.0 La (where La is the maximum allowable leakage rate at design pressure). Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine performance shall be at least 24 months.

Enclosure Page 11 of 60 Adoption of the Option B performance based containment leakage rate testing program altered the frequency of measuring primary containment leakage in Type A tests but did not alter the basic method by which Appendix J leakage testing is performed. The test frequency is based on an evaluation of the "as found" leakage history to determine a frequency for leakage testing which provides assurance that leakage limits will not be exceeded. The allowed frequency for Type A testing as documented in NEI 94-01, is based, in part, upon a generic evaluation documented in NUREG-1493 (Reference 7). The evaluation documented in NUREG-1493 included a study of the dependence or reactor accident risks on containment leak tightness for differing types of containment types, including a reinforced, shallow domed concrete containment similar to HBRSEP2 containment structures. NUREG-1493 concluded in Section 10.1.2 that reducing the frequency of Type A tests (ILRT) from the original three tests per ten years to one test per twenty years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Types B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements. Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, NUREG-1493 concluded that increasing the interval between ILRTs is possible with minimal impact on public risk.
3.1.2   HBRSEP2 10 CFR 50, Appendix J, Option B Licensing History Amendment No. 169 - SER dated May 28, 1996 - ML020530639 (Reference 11)
The Commission issued Amendment No.169 to Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2). This amendment changed the HBRSEP2 TS in response to the request dated January 31, 1996.
The amendment revised TS section 4.4 to allow the use of 10 CFR Part 50, Appendix J, Option B, Performance-Based Containment Leakage Rate Testing. A new TS section 6.12 described the containment leakage rate testing program that adopted 10 CFR Part 50, Appendix J, Option B for Type A tests; and 10 CFR Part 50, Appendix J, Option A, for Types B and C tests.
Amendment No. 193 - SER dated September 16, 2002 - ML022690765 (Reference 12)
The Commission issued Amendment No. 193 to Facility Operating License No. DPR-23 for HBRSEP2. This amendment consisted of changes to the TS in response to the application dated March 26, 2002, as supplemented by letters dated June 19, and August 8, 2002.
The amendment to the TS for HBRSEP2 modified TS SR 5.5.16, Containment Leakage Rate Testing Program, to require the performance of a Type A test within 12.1 years from the last Type A test, which was performed on April 9, 1992. This was a one-time extension to the 10-year performance-based Type A test interval based on an acceptably low level of risk as supported by a plant-specific risk assessment.
Amendment No. 199 - SER dated February 11, 2004 - ML040430023 (Reference 13)
The Commission issued Amendment No. 199 to Facility Operating License No. DPR-23 for HBRSEP2. This amendment changed the HBRSEP2 TS in response to the request dated June 11, 2003, as supplemented by letters dated August 20, and October 13, 2003.

Enclosure Page 12 of 60 The amendment allows the extension of the Appendix J, Type A, Containment Integrated Leak Rate Test, Option B, for HBRSEP2 from the scheduled May 2004 timeframe to no later than April 9, 2007.
Amendment No. 215 - SER dated June 15, 2007 - ML071070170 (Reference 14)
The Commission issued Amendment No. 215 to Renewed Facility Operating License No. DPR-23 for HBRSEP2. This amendment changed the TS in response to the application dated July 17, 2006. Specifically, the amendment revised the containment design pressure in SRs 3.6.8.1 and 3.6.8.5 concerning the Isolation Valve Seal Water System, and TS Section 5.5.16, Containment Leakage Rate Testing Program.
Amendment No. 220 - SER dated October 3, 2008 - ML0082210549 (Reference 15)
The Commission issued Amendment No. 220 to Renewed Facility Operating License No. DPR-23 for HBRSEP2 in response to the application dated November 29, 2007. The amendment consisted of changes to TS Section 3.6.8, "Isolation Valve Seal Water (IVSW) System."
The amendment revised SRs 3.6.8.2 and 3.6.8.6 related to IVSW tank volume and header flow rates. Specifically, the change clarified the wording of SR 3.6.8.2, and revised SR 3.6.8.6 to provide a total flow rate limit from all four headers in place of the individual header limits.
Relief Request IST-RR SER dated July 13, 2012 - ML12174A010 (Reference 16)
By letter to the NRC dated March 16, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12086A067), as supplemented by letter dated May 10, 2012, (ML12138A041), Carolina Power & Light Company doing business as Progress Energy Carolinas, Inc., submitted Relief Request-3 for the Inservice Testing (1ST) Program Plan for the fifth 10-year Interval for the H. B. Robinson Steam Electric Plant, Unit No.2 (HBRSEP).
HBRSEP requested approval to use an alternative test plan in lieu of certain IST requirements of the 2004 Edition through 2006 Addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM) Code for the Category C check valves exercise tests or exams at HBRSEP. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(a)(3)(i), HBRSEP requested to use proposed alternatives on the basis that the alternatives provided an acceptable level of quality and safety.
The NRC staff determined that the proposed alternative, described in IST-RR-3, provided reasonable assurance that valves IVSW-71, IVSW-72, IVSW-74 through IVSW-97, IVSW-100A, IVSW-100B, and IVSW-100C were operationally ready and provided an acceptable level of quality and safety. Accordingly, the NRC staff concluded that HBRSEP adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55(a)(3)(ii) and is in compliance with the ASME OM Code requirements. Therefore, the NRC staff authorized the proposed alternative in IST-RR-3 for the fifth IST interval at HBRSEP Unit No.2, which began on July 21, 2012, and is scheduled to end on February 18, 2022.
3.1.3    Containment Building Description General Description of the Containment (Reference UFSAR Section 3.8.1.1.1)

Enclosure Page 13 of 60 The reactor containment structure is a steel lined concrete shell in the form of a vertical right cylinder with a hemispherical dome and a flat base supported by means of piles. The containment structure is designed for an accident pressure based upon the pressure transients as shown in UFSAR Section 15.6. The containment structure is designed to contain radioactive material, which might be released from the core following a loss-of-coolant accident as described in UFSAR Section 6.2.1.
The structure consists of sidewalls measuring 126 ft. from the liner on the base to the springline of the dome and an inside diameter of 130 ft. The containment free volume is 1,950,000 ft3. The sidewalls of the cylinder and dome are 3 ft. 6 in. and 2 ft. 6 in. thick, respectively. The inside radius of the dome is equal to the inside radius of the cylinder (i.e., the discontinuity at the springline due to the change in thickness is on the outer surface). The base consists of a 10-foot thick structural concrete slab. The base liner is installed on top of the structural slab and covered with two feet of concrete.
The basic structural elements considered in the design of the containment structure are the piles, base slab, sidewalls, and dome acting as essentially one structure under all loading conditions. The bottom plates of the liner are laid loose on the foundation slab and are anchored only at the hang ways for the crane wall and primary shield. In the vertical walls and dome, the liner is anchored to the concrete shell by means of "KSM" shaped anchor studs fusion welded to the liner plate so that it forms an integral part of the entire composite structure under all loadings. The cylindrical portion of the liner is insulated. The dome of the containment is reinforced concrete. The cylinder walls are concrete-reinforced circumferentially and prestressed vertically. The base slab is reinforced concrete.
Containment Liner (Reference UFSAR Section 3.8.1.1.5)
The containment liner is designed to serve as a leakproof membrane and is not relied upon for the structural integrity of the containment except for resisting tangential shears in the dome. It is anchored to the concrete by means of "KSM" shaped steel studs. The liner is not anchored to the concrete base slab hence does not act compositely with it. It was laid loose on the base slab and the butt weld backing strips were set in grooves in the base slab. After welding, the distortions in the liner were considered too great and a neat cement grout was flowed beneath it to fill the voids. A bond breaker, form oil, was flowed first on the base slab to prevent the liner from acting compositely with the slab.
Stress conditions in the liner under all conditions of design have been analyzed to assure that the principle stresses do not exceed the yield or buckling stresses as provided in design stress criteria. Fatigue, accident, and operational loads are discussed in UFSAR Section 3.8.1.4.
Sample buckling calculations are shown indicating maximum loading conditions that prevail. A discussion on the liner anchors and transfer on their loads into the concrete has been submitted in the Containment Design Report.
The loading condition which produces maximum biaxial compression in the liner is that of winter operation combined with 1.0 times the hypothetical earthquake. Under this condition, the allowable buckling stress is not exceeded.
The steel liner and its welded seam joints are covered by carbon steel channels with pressurizing connections. These seam weld channels can be used to determine the leak-tightness of the liner seam welds.

Enclosure Page 14 of 60 The liner plate was protected against corrosion as follows: The entire inside surfaces of the dome and walls were sandblasted. The plate from elevation (el.) 228' to approximately el. 352' was coated with a zinc rich primer and modified alkyd topcoat (Keeler and long 6820/7230 system). The plate above el. 352' was coated with a zinc filed inorganic primer such as Carbozinc 11 and Phenoline 305 phenolic epoxy topcoat. The coated surfaces of the liner plate are protected by insulation and sheathing up to el. 367'10". All painted surfaces were inspected after erection and any damaged areas reprimed before finishing. The face of the liner plate in contact with the concrete has no primer or paint applied; the intimate contact with the concrete provides corrosion protection. Maintenance coatings are specified and applied equal to or better than the original coatings. (Reference UFSAR 3.8.1.6.1.5)
The only exception to the above described Liner Protection coating and inspection is in the areas of the field welds at the new Conax Cartridge Type Electrical Penetrations. The inside surface of the field weld is inaccessible after installation, and the coating and inspection described above cannot take place. These welds are radiographed at installation.
The containment liner is 1/2 in. thick from el. 226'-0" to el. 253'-0". The liner is 3/8 in. thick from el. 253'-0" to 352'-0" and 1/2 in. thick above 352'-0". The liner on top of the base mat at el. 226-0 is 1/4 in. thick. A partial diagram of the containment liner is provided in Attachment 6, Containment Liner Detail At Moisture Barrier.
Containment Penetrations Penetrations through the containment reinforced concrete pressure barrier for pipe, electrical conductors, ducts, and access hatches are typically of the double barrier type. In general, a penetration consists of a sleeve embedded in the concrete wall and welded to the containment liner. The weld to the liner is shrouded by a channel, which was used to demonstrate the integrity of the penetration-to-liner weld joint. The pipe, electrical conductor cartridge, or duct passes through the embedded sleeve and the ends of the resulting annulus are closed off, either by welded end plates, bolted flanges, or a combination of these. Provisions are made for differential expansion and misalignment between pipe or cartridge, and sleeve. Pressurizing connections are provided to demonstrate the integrity of the penetration assemblies.
(Reference UFSAR3.8.1.1.6)
An exception to this are electrical penetrations C1, C2, C3, C5, C9, C10, D9, E1, E5, E10 and F1. These penetrations have double pressure barrier protection in their header plate and therefore an endplate is required at one end only.
Electrical Penetrations (Reference UFSAR 3.8.1.1.6.1)
Cartridge type penetrations are used for all electrical conductors passing through the containment, with the exception of the penetration at sleeve numbers C9, C10, D9, E5 and E10 in the north cable vault, and sleeve numbers C1, C2, C3, C5, E1 and F1 in the south cable vault which are of the capsule type design. The penetration cartridge is a hollow cylinder closed on both ends, through which the conductors pass. This cartridge is provided with a pressure connection to allow continuous or intermittent pressurization of the penetration. The method used to seal the joint between the cartridge end plate and the conductor depends upon the type of cable involved.
In general, there are four types used:
: 1. Type 1 - High voltage power, 4160 volts

Enclosure Page 15 of 60
: 2. Type 2 - Power, control, and instrumentation, 600 volts and below
: 3. Type 3 - Thermocouple leads
: 4. Type 4 - Coaxial and triaxial cables Type 1 penetrations are rubber insulated copper rods. These insulated rods will pass through a leak tight gland fitting threaded into each end plate of the cartridge. Either alumina insulating bushings or fused glass seals may be used to provide the double barrier.
Type 2 penetrations are single or multiconductor mineral insulated cable with a metallic sheath.
This cable will pass through a leak tight gland threaded into each end of the penetration cartridge. The ends of the mineral insulated cable are potted with epoxy resin. Copper rod conductors with fused glass seals in the cartridge and plates is an alternate which may be used.
Type 3 penetrations are the same as Type 2 except the conductors will be thermocouple material. The sealing methods are the same as for the Type 2 penetrations.
Type 4 penetrations are used for coaxial and triaxial cables. In addition to the leak tight gland fittings in the cartridge end plate, a plug and receptacle connection provides a double barrier to leakage through the cable itself. An alternate method uses fused glass seals in the cartridge end plates and fused glass seals between the conductors of the coaxial or triaxial cable.
In the capsule penetration design, a single stainless plate is machined with the required quantity of feed-through ports which are interconnected by peripherally machined gun drills which creates a manifold system for pressure monitoring. Feed-throughs are assembled through the plate and sealed in place with a patented metal compression fitting assembly which creates seal zones at the front and backside of the plate, while allowing for a chamber to form between the seal zones to accommodate leakage monitoring.
The capsule series penetration is designed for a weldment interface to the containment nozzle.
The weldment interface is by a transition ring, factory welded to the penetration header plate, and field welded to the containment nozzle.
The penetration sleeves to accommodate the electrical penetration assembly cartridges are 10 in., Schedule 80 carbon steel pipe, except where otherwise noted. For the electrical penetrations C1, C2, C3, C5, C9, C10, D9, E1, E5, E10, and F1, the header plate and conductors are pressurized. There are 51 electrical penetrations.
Piping Penetrations (Reference UFSAR 3.8.1.1.6.2)
Double barrier piping penetrations are typically provided for piping passing through the containment. The pipe is centered in the embedded sleeve which is welded to the liner, except for small pipes where several pipes may pass through the same penetration sleeve.
The penetrations for the main steam, feedwater, blowdown, and sample lines are designed so that the penetration is stronger than the piping system and that the vapor barrier will not be breeched due to a hypothesized pipe rupture.
Typically, end plates are welded to the pipe at both ends of the sleeve. Several pipes may pass through the same embedded sleeve to minimize the number of penetrations required. In this case, each pipe is welded to both end plates. A connection to the penetration sleeve is provided to allow continuous or intermittent pressurization of the compartment formed between the piping

Enclosure Page 16 of 60 and the embedded sleeve. In the case of piping carrying hot fluid, the pipe is insulated. The RHR supply pipe is insulated to keep the concrete surrounding the embedded sleeve below 200&#xba;F. Typical hot and cold pipe penetrations are shown in UFSAR Figure 3.8.1-15. There are 46 containment penetrations sleeves for pipes. Pipes are anchored to the structural steel girders as close as possible to the inside of the wall or to the crane wall. Loads due to pipe ruptures within the containment or due to thermal stresses are not transferred to the liner.
An exception to this is the steam generator blowdown penetrations and two safety injection penetrations (RHR penetrations S-14 and S-15). The end plate is welded directly to the sleeve.
The sleeve is welded to the liner reinforcement plate. Piping loads are transmitted to the concrete wall, except for torsion loads, which are carried by the liner plate. However, the torsion loads are below the liner allowable stress.
Two piping penetrations are provided in the containment sump area.
Equipment and Personnel Access Hatches (Reference UFSAR 3.8.1.1.6.3)
An equipment hatch is provided which is fabricated from welded steel and furnished with a double-gasketed flange and bolted dished door. Equipment up to a diameter of approximately 18 ft. can be transferred into and out of containment via this hatch. The hatch barrel is embedded in the containment wall and welded to the liner and is a portion of the structural frame embedded in the wall. Provision is made to pressurize the space between the double gaskets of the door flanges and the weld seam channels at the liner joint, hatch flanges, and dished door. Pressure is relieved from the double gasket spaces prior to opening the door.
The personnel hatch is a double door, hydraulically-latched, welded steel assembly. It is attached to the structural frame embedded in the wall of which the frame barrel forms the central portion of the lock. A quick-acting type, equalizing valve connects the personnel hatch with the interior of the containment vessel for the purposes of equalizing pressure in the two systems when entering or leaving the containment.
The personnel hatch doors are interlocked to prevent both being opened simultaneously and to ensure that one door is completely closed before the opposite door can be opened. Indicating lights and annunciators situated in the control room indicate the door operational status.
Provision is made to permit bypassing the door interlocking system to allow doors to be left open during plant cold shutdown. Each door lock hinge is designed to be capable of independent three-dimensional adjustment to assist proper seating. An Emergency Lighting and Communication System operating from an external emergency supply is provided in the lock interior. Emergency access to either the inner door, from the containment interior; or to the outer door, from outside, is possible by the use of special door unlatching tools.
Fuel Transfer Penetration A fuel transfer penetration is provided for fuel movement between the refueling transfer canal in the reactor containment and the spent fuel pit. The penetration consists of a 20 in. stainless steel pipe installed inside a 24 in. pipe (Reference UFSAR Figure 3.8.1-16). The inner pipe acts as the transfer tube and is fitted with a double-gasketed blind flange in the refueling canal and a standard gate valve in the spent fuel pit. This arrangement prevents leakage through the transfer tube in the event of an accident. The outer pipe is welded to the containment liner and provision is made by use of a special seal ring for testing all welds essential to the integrity of

Enclosure Page 17 of 60 the penetration. Bellows expansion joints are provided on the pipes to compensate for any differential movement between the two pipes or other structures. (Reference UFSAR 3.8.1.1.6.4)
Containment Supply and Exhaust Purge Ducts The ventilation system purge ducts are each equipped with two quick-acting tight-sealing butterfly valves for isolation purposes. The valves are manually opened for containment purging, but are automatically actuated to the closed position upon a safety injection signal or high containment radiation level signal. (Reference UFSAR 3.8.1.1.6.5)
Containment Dome The dome is a hemispherical dome 65 ft. inside diameter and 2 ft. 6 in. thick reinforced concrete. The difference in cylinder and dome thickness is effected on the outside surface, the transition between thicknesses being accomplished 13 ft. above the springline of the dome at the anchor surface of the cylinder prestressing steel tendons. The inside of the dome is insulated from the springline to a point above the anchor surface of the cylinder prestressing steel tendons. The outer surface of the dome is covered with a membrane roof to provide weather protection. (Reference UFSAR 3.8.1.1.7)
Insulation Containment liner insulation consists of 44 in. x 84 in. x 1 1/4 in. thick, 4 lb/ft3 density crosslinked PVC foam and/or 2 lb/ft3 density Polyimide foam with an outer covering of 0.019 in.
thick stainless steel. Panels are erected with the 44 in. dimension vertical and the 84 in.
dimension horizontal. (Reference UFSAR 3.8.1.6.1.7)
Prestressing Steel Tendon Design The prestressing system chosen for post-tensioning the Robinson containment structure in the vertical direction, consists of 1 3/8 in. diameter high strength steel bars closely grouped into tendons consisting of six bars per tendon.
These tendons are placed within heavy wall 6 in. galvanized steel pipe sheaths.
Tendons are on the centerline of the wall and are spaced approximately every 3 ft around the periphery of the containment. This concept of grouping a number of high strength bars is not new to the industry, a seven bar group having been used in a dam in South America. It utilizes standard concepts and components proven adequate through experience in the United States, and more experience using the similar system in Europe.
The bottom anchorage is a steel plate with six threaded holes into which are screwed the steel bars. The thread used is a tapered, cut thread designed and proven to develop the minimum guaranteed ultimate tensile strength of the bar. Complete seating of the thread is necessary to develop the full tensile strength of the bar and this is easily accomplished and inspected since the thread is fully engaged when no threads are showing at the inner face of the plate.
Threaded holes in the bottom bearing plate have been sealed from the bottom by a steel plug with a binding compound between the plug and the bar. The entire bottom anchorage was designed and tested to show no permanent physical distortion at the minimum ultimate tensile strength of the tendon, which is a 25 percent greater load than the maximum load to which it will be subjected in the life of the structure.

Enclosure Page 18 of 60 Couplings consist of internally threaded sleeves into which the high strength steel bars are securely screwed. The same screw thread details are used as described for the bottom anchorage. The void space between the bars within the coupler body is filled with the same binding compound used to bind the threads to eliminate any possibility of corrosion.
The tendon coupling consists of a set of six individual bar couplers staggered in elevation. Each tendon has two couplers in its length, one at the construction joint in the cylinder wall at El. 250 ft. which is a field assembled coupling and a second half way up the wall between the construction joint and the top anchorage which is a shop assembled coupling. To assure the integrity of the coupler, the threads are coated with an epoxy compound, which binds the coupler sleeve to the bar and prevents the possibility of unthreading due to vibration during shipping or erection. Once tensioned, the friction within the coupler threads eliminates any tendency for unthreading.
The top anchorage consists of a steel plate bearing on the concrete with three of the bars anchored to this plate by means of Howlett Grip Nuts. A second and smaller plate bears on the top of these grip nuts and the remaining three bars are anchored to the top plate by means of Howlett Grip Nuts. The Howlett Grip Nut is a modified positive action wedge anchor, which has the advantages of a wedge anchor and the positive adjustment capability of a threaded anchor.
It does not require an exact predetermination of bar length along with all of the fine shimming required at the top anchorage with such a predetermination. This concept of stacking the top anchorage details allows a closer grouping of bars than is ordinarily required with a bar system.
The steel sheath surrounding each tendon is made of 6 in. Schedule 40 galvanized steel pipe with threaded and flanged connections. The sheath is connected to the bottom anchorage plate by means of a threaded coupling and provides protection of the tendon both during and after construction.
The corrosion protection scheme used during construction was a nitrogen atmosphere, a system widely used for corrosion protection. As an additional check, six extra removable bars were placed in six of the lower tendon assemblies. These bars were removed to monitor the effectiveness of the nitrogen atmosphere corrosion protection. All tendons were inspected visually from the top before coupling or grouting to check that corrosion had not occurred.
The tendons were initially tensioned in late March/early April 1970, and then re-tensioned in May 1970. The last exterior wall concrete lifts were placed in December 1969 and the first lifts were placed in November 1968. (Reference UFSAR 3.8.1.4.7)
Tendon Surveillance Duke Energy Progress (DEP) believes that there is sufficient evidence in the history of the prestressed concrete industry to justify the specifying of an uniaxially prestressed concrete containment vessel such as the HBRSEP2 containment with full confidence that it will perform within the criteria set in its design. Conservative values have been used in estimating qualities of materials, which affect the net prestressing force. (UFSAR 3.8.1.7.2)
As an example of the conservatism used, consider creep and shrinkage. Design values of 0.0003 in. shrinkage and 2.25 for coefficient of creep (creep strain/elastic strain) were specified.
It is expected that values of 0.0001 in. shrinkage and 1.7 for coefficient of creep are realistic values based on preliminary estimates using as a guide Hanson, T. C. and Mattock A. H.,

Enclosure Page 19 of 60 "Influence of Size and Shape of Member on the Shrinkage and Creep of Concrete," ACI Jour.,
Proceedings, Vol. 63, p. 267 (1966).
Such a conservative design can only result in higher precompression stresses in the concrete and higher tensile stresses in the tendons. This is of little interest, since even with these higher tensile stresses, the tendons will never reach the tensile stress imposed upon them with the initial prestressing operation.
There is no practical method of surveying the tendon stress and corrosion, creep and shrinkage of the concrete for a grouted tendon. Known conservative analytical procedures, in addition to successful experience application for grouted tendons, do not warrant a surveillance program.
However, two surveillance tendons similar to the service tendons and in a similar environment are provided. These may be uncovered at any time for surveillance of any corrosion.
The surveillance tendons consist of two short tendons similar to the service tendons. Each tendon consists of six 3/8 in. &#xf8; bars in 6 in. pipe sheath with anchor plates, prestressing hardware, and grout pipe identical except for length to the working tendons. They are embedded in a section of concrete approximating the same environment as that of the service tendons.
The program for inspection consisted of removing one tendon after 5 years and the other after 25 years.
The removed tendons were sent to a commercial laboratory qualified to perform material tests and analysis. The tendon bars were removed from the sheath and the grout removed. The visual inspection was performed to detect and record evidence of corrosion. Tensile tests were then performed on selected bars to develop stress-strain diagrams and determine the bars' ultimate tensile strengths. The results of these tests were compared with the original properties to determine any significant changes. DEP retains a qualified engineering firm to assess the results of these tests and make recommendations.
The first containment surveillance tendon sample was removed in March 1976. The second containment surveillance tendon sample was removed in April 1997. Based on the information presented in the reports, and on other available data on the Robinson containment system, the tendon surveillance program is judged to be satisfactory. The tests showed that surveillance tendon specimens tested exceeded the minimum-breaking load of 238,000 pounds given in the FSAR. It can be reasonably concluded that similar results would be obtained if bars from the actual containment tendons inservice were tested.
3.1.4    Isolation Valve Seal Water System (IVSW)
The IVSW System assures the effectiveness of certain containment isolation valves during any condition, which requires containment isolation, by providing a water seal at the valves. These valves are located in lines that are connected to the Reactor Coolant System (RCS), or that could be exposed to the containment atmosphere in the event of a loss of coolant accident (LOCA). The system provides a reliable means for injecting seal water between the seats and stem packing of the globe and double disc types of isolation valves, and into the piping between other closed isolation valves. The system provides assurance that, should an accident occur, the containment leak rate is no greater than that assumed in the accident analysis by providing seal water at a pressure  1.1 times Pa (46.2 psig). The system is designed to maintain this seal

Enclosure Page 20 of 60 for at least 30 days. The possibility of leakage from the containment or RCS past the first isolation point is thereby prevented by assuring that if leakage does exist, it will be from the IVSW System into containment.
The system includes one 175-gallon seal water tank capable of supplying the total requirements of the system. The IVSW tank's required volume is maintained and the tank is pressurized with nitrogen. The normal supply of makeup water to the IVSW tank is the Primary Water System. In the event Primary Water is not available, emergency makeup can be supplied from the Service Water System. The Plant Nitrogen System provides the normal supply of nitrogen to the IVSW tank. An automatic backup supply is provided from two dedicated high pressure nitrogen bottles.
The system is normally in a static condition with the seal water injection tank filled and pressurized. Indication of IVSW tank level and pressure along with corresponding low level and low pressure alarms are provided in the Control Room. The tank supplies pressurized water to four distribution headers. Header "A" requires manual operation and serves lines that are normally filled with fluid following a LOCA, and lines that must remain in service for a period of time following the accident. Headers "B", "C", and "D" are automatic headers that are pressurized through one or both of two redundant, fail open, air operated valves arranged in parallel. A loss of power will cause these valves to fail open. System operation is initiated by a Phase A containment isolation signal which accompanies any Safety Injection (SI) signal.
Safety Analyses The Design Basis Accident (DBA) that results in a release of radioactive material within containment is a LOCA. The analyses for the LOCA assumes the isolation of containment is completed and leakage from containment is at a rate equivalent to the design leakage rate. As part of the containment boundary, containment isolation valves function to support the leak tightness of containment. By maintaining this barrier, offsite dose calculations will be less than the limits of 10 CFR 100 or 10 CFR 50.67, as applicable, during a DBA.
The IVSW System actuates on a containment isolation signal and functions to assure the actual leakage is no greater than the design value. IVSW assures the effectiveness of certain isolation valves to limit containment leakage by pressurizing the affected containment penetration flow paths at a pressure  1.1 times Pa. IVSW is designed to maintain this seal for at least 30 days.
A single failure analysis shows the failure of any active component will not prevent fulfilling the design function of the system. By meeting these requirements, IVSW is considered a qualified seal system in accordance with 10 CFR 50, Appendix J.
The IVSW System satisfies Criterion 3 of the NRC Policy Statement.
Acceptance Review / Evaluation (Reference 28)
By letter dated April 23, 1979, the NRC completed the review and evaluation regarding the acceptability of the IVSW system. The results of this evaluation was as follows:
Paragraph III.C.3.b of Appendix J to 10 CFR 50 requires that the installed isolation valve seal water system fluid inventory be sufficient to assure the sealing function for at least 30 days at a pressure of 1.1 Pa. The IVSW system proposed by the licensee can provide seal water at a pressure equal or greater than 1.1 Pa and for a period greater than 30 days. We, therefore, conclude that the proposed IVSW system meets the requirements of Appendix J.

Enclosure Page 21 of 60 We have also reviewed the system design with respect to the requirements of an engineered safety feature, because the system will be used and relied upon during and following an accident. Based on the information in the licensee's submittal and the FSAR, we find that the proposed system including associated components, piping, and structures are designed to Class I seismic criteria. In addition, two separate, independent, seismically qualified sources of makeup water are provided for long term operation at a pressure greater than 1.1 Pa. A single failure analysis shows that the failure of any single active component will not prevent fulfilling the design function of the system. We, therefore, conclude that the proposed IVSW system meets the requirements of an engineered safety feature, and can be relied upon to fulfill its design function during and following an accident.
3.1.5   Containment Penetration Pressurization System (PPS)
The Containment PPS provides a means of testing pressure zones incorporated into the containment penetrations. It was originally designed to provide a means of continuously pressurizing the positive pressure zones in order to maintain these zones above the maximum containment post-accident pressure and to provide a means for continuous or intermittent monitoring of the leakage status of the containment penetrations.
Modification ESR/MOD 95-00888 removed the automatic continuous pressurization and monitoring features of this system. It is now only used during power operation to test the personnel hatch and during outages to test containment penetrations (LLRTs). The system is capable of providing continuous pressurization should the need arise.
3.2     Inspections 3.2.1   Primary Containment Coatings Condition Assessment Assessment of the protective coatings inside Primary Containment is conducted during each refueling outage. Protective coatings inside Primary Containment are assessed to identify and quantify coatings degradation and unqualified coatings. The condition assessment is consistent with ASTM D5163 and NUREG 1801, Chapter XI.S8, "Protective Coating Monitoring and Maintenance Program", and meets the requirements for protective coating monitoring and maintenance.
Coatings are considered to be acceptable, provided none of the following conditions are observed:
* Blistering is not greater than size No. 6 (Medium) as specified in ASTM D714
* Cracking greater than standard No. 6 as specified in ASTM D661 (Checking of any grade specified in ASTM D660 is acceptable and need not be recorded)
* Flaking greater than standard No. 6 as specified in ASTM D772
* Rusting equal to or greater than Grade 7 as specified in ASTM D610
* Rust staining, accumulated dirt or dirt containing iron compounds should not be confused with actual rusting of the steel substrate. Only rusting of the substrate under a coating need be considered.
* Insufficient adhesion, as determined by the Coating Program Manager (e.g., visual observation of loose coatings, qualitative or quantitative assessment, etc.). Adhesion

Enclosure Page 22 of 60 tests may be conducted to quantify the condition of coating adhesion per ASTM D4541 or ASTM D7234.
* Unqualified Coatings.
Mechanical damage is considered to be acceptable, provided none of the above noted conditions are observed in the surrounding coatings, and provided that any exposed substrate or underlying coating has been evaluated for acceptability.
Areas within the scope of this assessment, but not inspected, should be identified with justification provided. Justification should state reason the areas is considered inaccessible and assess potential for adverse impact on emergency core cooling system (ECCS) Sump performance from potential coating debris from the un-inspected areas.
Areas inspected under the ASME Section XI, Subsection IWE/IWL program do not require examination in accordance with the coatings assessment, although the results of the Service Level 1 coating inspections shall be included in the evaluation.
Recordable conditions shall be addressed in accordance with the Corrective Action Program.
Multiple recordable conditions found during any one outage assessment may be documented on the same NCR. The NCR should include consideration of the following:
* Past and projected impact on ECCS performance
* Extent of condition
* Maintenance Rule impact
* Corrective actions taken and/or planned If corrective action is taken to remediate a recordable condition during the refueling outage in which it was identified, a follow-up assessment of the area shall be performed.
A Condition Assessment Summary Report shall be prepared by the Program Manager, which includes the following:
* A brief description of assessment results and the general condition of Primary Containment coatings
* Apparent cause(s) of degradation and trend evaluation
* Recommendations for future repairs and surveillance
* Condition Assessment forms which include:
o Location, characterization, and disposition of recordable conditions o Corrective actions taken and recommended o Photographic documentation
* Total quantity of Unqualified Coatings (UC) and comparison to previous inspection results and the applicable Maintenance Rule and/or design limits for ECCS functionality
* List of NCRs initiated as a result of the condition assessment
* Identity and functional role (i.e.; QC, Coatings Program Manager, Coatings Service Level I Contractor, etc.) of those individuals performing the inspections.
Include the Service Level III - Safety Related coatings in the Condition Assessment Report as they relate to Primary Containment conditions.
3.2.2    Inservice Inspection Program for Containment - IWE/IWL On August 8, 1996, an amendment to 10 CFR 50.55a was published in the Federal Register (61 FR 41303) to incorporate by reference, the requirements of the ASME Code, Section XI,

Enclosure Page 23 of 60 Subsections IWE and IWL. The objective of this amendment was to specify requirements to ensure that the critical areas of a containment structure were routinely inspected to detect and correct defects that could compromise its structural integrity. The rationale of the NRC to issue this amendment was based on (1) the rate of occurrence of degradation of containments and (2) their determination that current licensee containment inspection programs were not adequate to detect degradation. The effective date of this amendment to 10 CFR 50.55a was September 9, 1996.
Until September 9, 1996, requirements for the inservice inspection of components classified as Class MC and CC were not mandated by Federal Regulation. For this reason, preservice and/or inservice inspection of the components classified as Class MC and CC at HBRSEP2 had not been performed in previous Inspection Intervals per ASME Section XI.
This program implements required License Renewal commitments.
IWE/IWL Program Description The program was developed to implement the requirements of ASME Section XI as modified by 10 CFR 50.55a. The examinations required by this program may be utilized to satisfy 10 CFR 50, Appendix J inspections and10 CFR 50.65 Maintenance Rule inspections, as deemed appropriate by plant organizations responsible for the implementation of those programs.
The components subject to ASME Section XI, Subsection IWE and IWL requirements are those that make up the containment structure, its leak tight barrier (including integral attachments) and those that contribute to its structural integrity. These components are listed in the Second Ten-Year Inservice IWE/IWL Inspection Plan.
Specifically included are Class MC pressure retaining components and their integral attachments, (including metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments), per IWE-1100; and Class CC reinforced concrete containments and unbonded posttensioning systems, per IWL-1100. Because Subsection IWL requires examination and testing of unbonded post-tensioning systems only, the grouted bonded post-tensioning system at HBRSEP2 is not subject to ASME Section XI rules.
The First Containment Inspection Interval at HBRSEP2 was effective from September 9, 1998 to September 8, 2008, for Subsection IWE and Subsection IWL activities. This time frame included the first and second five-year examinations required by Subsection IWL. As required by the NRC final rulemaking, this plan was developed to support the completion of expedited examinations by September 9, 2001. The applicable ASME Section XI Code utilized for the First Ten Year Interval was the 1992 Edition, 1992 Addenda.
The Second Containment Inspection Interval at HBRSEP2 is effective from September 9, 2008, to September 8, 2018, for Subsection IWE and Subsection IWL activities.
The applicable ASME Section XI Code Year and Addenda for the Second Ten-Year IWE/IWL Program is the 2001 Edition with the 2003 Addenda of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Subsections IWA, IWE and IWL and Inspection Program B of IWE-2412 and IWL-2420.
The Class MC components subject to examination include the following:
* Pressure Retaining Metal Containment Liner

Enclosure Page 24 of 60
* Penetration Assemblies
* Airlock
* Equipment Hatch
* Moisture Barriers
* Pressure Retaining Bolting (Shown on the applicable Examination Boundary Drawings)
* Integral Attachments (Shown on the applicable Examination Boundary Drawings)
Per IWE-1220, the following components (or parts of components) are exempted from the examination requirements of IWE-2000. This section describes some of the design and access limitations. In addition, specific access component limitations are typically identified on the visual examination data sheets.
* Vessels, parts, and appurtenances that are outside the boundaries of the containment as defined in the Design Specification.
* Embedded or inaccessible portions of containment vessels, parts, and appurtenances that met the requirements of the Original Construction Code; components that meet this criteria include:
o Containment Vessel (CV) liner made inaccessible by the concrete floor slab at elevation 228 (see detail G-G on drawing G-190367) o Nelson studs welded to the CV liner embedded in the containment concrete o CV liner at the containment sumps embedded in concrete o Penetration assembly for the 14 ECCS containment sump piping (Penetrations S-X1 and S-X2) embedded in concrete (see drawings G-190267 and G-190359) o CV liner made inaccessible by weld test channels.
* The CV Liner considered inaccessible by the insulation and sheathing between elevations 228 and 367-10. Although not exempted from examination per IWE-1220, the CV Liner at these locations are considered inaccessible, unless a maintenance activity requires removal of the insulation and sheathing.
* Portions of containment vessels, parts, and appurtenances that becomes embedded or inaccessible as a result of vessel repair or replacement if the conditions of IWE-1232(a) and (b) and IWE-5220 are met.
As allowed by 10 CFR 50.55a(b)(2)(ix)(C), HBRSEP2 has elected to exclude Category E-B and Category E-F weld examinations from the Subsection IWE portion of the Inservice IWE/IWL Program.
IWE Examination Schedule The current period/interval schedule for IWE examinations is summarized in the Table below and scheduling of individual components per the requirements of Table IWE-2412-1 is contained in the Second Ten Year Interval Inservice IWE/IWL Inspection Plan.
Table 3.2.2-1, IWE PERIOD/INTERVAL SCHEDULE Interval               Period 1                Period 2              Period 3 1                 9/98 - 9/01            9/01 - 9/05            9/05 - 9/08 2                 9/08 - 9/11            9/11 - 9/15            9/15 - 9/18 3                 9/18 - 9/21            9/21 - 9/25            9/25 - 9/28 Inaccessible Class MC Areas

Enclosure Page 25 of 60 Inaccessible Class MC components subject to 10 CFR 50.55a requirements are those that meet the criteria of 10 CFR 50.55a(b)(2)(ix)(A), which requires the identification and evaluation of inaccessible Class MC areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
The following areas that meet this criterion have been identified and evaluated:
* Moisture barrier degradation (Behind six insulation panels)
* CV Liner at Elevation 228. (Behind six insulation panels)
* CV Liner from Elevation 232 to Elevation 367 (behind insulation)
* CV Liner behind weld channels (does not meet criteria of 10 CFR 50.55a(b)(2)(ix)(A).
* CV Liner below the floor slab.
The requirements of Subsection IWL for Class CC components are identified on the Examination Boundary Drawings and are listed in the Second Ten Year Interval Inservice IWE/IWL Inspection Plan.
The Class CC components subject to examination include the following:
* Dome roof
* Cylinder walls Per IWL-1220, the following components (or parts of components) are exempted from the examination requirements of IWL-2000. This section describes some of the design and access limitations. In addition, specific component access limitations are typically identified on the visual examination data sheets.
* Tendon end anchorages that are inaccessible, subject to the requirements of IWL-2521.1. This exemption is not applicable to HBRSEP2. The plant design includes a bonded post tensioning system. Per IWL-2220.2, examination requirements are applicable to unbonded Post-Tensioning systems.
* Portions of the concrete surface that are covered by liner, foundation material, or backfill, or are otherwise obstructed by adjacent structures, components, parts, or appurtenances. Components that meet this criteria include:
o Inside face of containment concrete made inaccessible by the liner o All surfaces of the vertical containment concrete slab made inaccessible by the liner and floor slab (inside containment) and by backfill (outside containment)
(see drawing G-190358).
o Outside face of containment concrete made inaccessible by the adjacent structures such as the auxiliary and fuel buildings.
Although normally exempt from the examination requirements of IWL-2000, when below-grade concrete is exposed by excavation for any reason, these IWL components are subject to examination consistent with the requirements for nonexempt IWL components in accordance with the respective License Renewal Commitments.
Degradation of below-grade IWL concrete attributed to aggressive groundwater will be used as a leading indicator for potential degradation to other below grade concrete structures in the scope of license renewal. Trending requirements for structures based on monitoring aggressive ground water are addressed in the respective License Renewal Commitments.
The piles supporting the containment concrete slab are outside the scope of Subsection IWL.

Enclosure Page 26 of 60 Implementation of the Subsection IWL portion of the Inservice IWE/IWL Program from a schedule standpoint will be in accordance with the Second Ten-Year Inservice IWE/IWL Inspection Plan IWL-2400, Inspection Schedule, which requires concrete to be examined at 1, 3-, and 5-year frequencies following the containment Structural Integrity Test (SIT) and every 5 years thereafter per IWL-2410. The 10-year and subsequent examinations shall commence not more than 1 year prior to the specified dates and shall be completed not more than 1 year after such dates. If plant-operating conditions are such that examination of portions of the concrete cannot be completed within this stated time interval, examination of those portions may be deferred until the next regularly scheduled plant outage.
IWL Examination Schedule The current schedule for IWL examinations is summarized in the Table below.
Table 3.2.2-2, IWL Schedule Inspection                               Year of Inspection 1st 5-Year Inspection                                  2000 nd 2 5-Year Inspection                                    2005 3rd 5-Year Inspection                                  2010 4th 5-Year Inspection                                  2015 5th 5-Year Inspection                                  2020 th 6 5-Year Inspection                                    2025 7th 5-Year Inspection                                  2030 The evaluation of nondestructive examinations of Subsection IWL components will be performed in accordance with Article IWL-3000.
The repair of Subsection IWL components will be performed in accordance with Article IWL-4000 (per IWA-4000).
Pressure testing of IWL components will be in accordance with Subsection IWL-5000.
Inaccessible Class CC Areas The components subject to examination per 10 CFR 50.55a are those that meet the criteria of 10 CFR 50.55a(b)(2)(viii)(E), which requires licensees to identify and evaluate the inaccessible Class CC areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
Currently, no areas have been identified that meet this criterion. In the event that such areas are identified, these areas will be added to the IWE/IWL Program scope of components and will be evaluated.
Relief Requests (RRs)
This Program contains requests for relief from certain Code requirements. The bases for these RRs and proposed alternative examinations or requirements are summarized within Appendix B, Relief Requests. More detailed descriptions, bases, and proposed alternative examinations

Enclosure Page 27 of 60 or requirements associated with each request for relief were submitted as a separate enclosure to the NRC.
The portions of the concrete surface that are covered by the liner, foundation material, or backfill or are otherwise obstructed by adjacent structures, components or appurtenances are exempt from the requirements of ASME Section XI, IWL-1220(b).
Examinations of components are scheduled in accordance with Inspection Program B (IWA, IWE, IWL and Tables IWE-2500-1 and IWL-2500-1). Examinations are scheduled, to the extent practical, based upon the preceding sequence of the First Ten-Year Interval IWE/IWL Inspection Plan.
IWE/IWL Augmented Examinations In accordance with the Second Ten-Year Interval IWE/IWL Program and Plan, IF areas are identified requiring Augmented Examinations (E-C) during the interval, THEN they will be listed in the E-C AUGMENT section of the Second Ten-Year Interval Plan.
Examinations identified in the Plan shall show the period in which the deficiency was identified as well as the successive examination period.
If an augmented examination is identified during the third period of an interval, the scheduling and re-examination shall occur during the first period of the successive interval.
The listing of augmented examinations is provided in Attachment 5, E-C AUGMENT Plan Tables for the Second Ten-Year Interval.
New Technologies to Perform Inspections of Inaccessible Class MC/CC Areas HBRSEP2 has not needed to implement any new technologies to perform inspections of any inaccessible areas at this time. However, Duke Energy Progress actively participates in various nuclear utility owners groups and ASME Code committees to maintain cognizance of ongoing developments within the nuclear industry. Industry operating experience is also continuously reviewed to determine its applicability to HBRSEP2. Adjustments to inspection plans and availability of new, commercially available technologies for the examination of the inaccessible areas of the containment would be explored and considered as part of these activities.
ASME Code Cases There are NO ASME Code Cases applicable to HBRSEP2.

Enclosure Page 28 of 60 Relief Requests Relief       Affected      Examination   Item   Examination         Alternative         Status Request     Component(s)       Category     No.        Area         Examinations Number IWE/IWL     Containment      E-A           E1.11  Insulated      General/Detailed      Granted RR-01       Liner                                 portions of    visual exam when      TAC No.
the            panels are removed    (MD8509)
Containment    for maintenance       Dated Liner          purposes and 1/3 of   4/30/2009 the lowest elevation panels per Period, 100% for the Interval IWE/IWL     Containment      E-A           E1.30  Containment    General/Detailed      Granted RR-02       Moisture Barrier                      moisture       visual exam when      TAC No.
barrier        panels are removed    (MD8509) for maintenance       Dated purposes and 1/3 of   4/30/2009 the lowest elevation panels per Period, 100% for the Interval A review of the data associated with the moisture barrier inspections indicated degradation of the moisture barrier in some locations, which required removal and reapplication of the moisture barrier in the First Ten-Year Interval. Liner inspections had revealed different grades of degradation of coatings, which required removal and reapplication of coatings in the First Ten-Year Interval. Because of the degraded conditions identified during the First Ten-Year Interval, the above listed relief requests will be discussed in detail in support of the continued acceptability of CV liner and moisture barriers.
The reliefs will be implemented during the HBRSEP2, Second Ten-Year IWE/IWL Inspection Interval (September 9, 2008, through September 8, 2018) for containment inspections required by ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Subsections IWE and IWL.
Relief Request No. IWE/IWL-RR-01, Visual Examination Of Insulated Containment Liner Code Requirements for Which Relief is Requested The ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Table IWE-2500-1, "Examination Categories," Examination Category E-A, Item Number E1.11, requires a general visual examination of 100% of the accessible surface areas of containment in accordance with acceptance standard IWE-3510, "Standards for Examination Category E-A, Containment Surfaces," paragraph IWE-3510.2, "Visual Examination of Coated and Noncoated Areas," for Class MC and metallic liners of Class CC components.
This relief was previously approved for the First Ten-Year IWE/IWL Inspection Interval under a Safety Evaluation Report (TAC No. MA4637) dated July 26, 1999.
Specific Relief Requested Relief is requested from performing general visual examinations in accordance with ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Table IWE-2500-1, Examination Category E-A, Item Number E 1.11 on the accessible surface areas of the containment liner,

Enclosure Page 29 of 60 which are insulated. The code requires 100% examination of the liner each examination period.
Proposed alternative examinations are provided below.
This request for relief is applicable to the insulated portion of the containment liner classified as Class MC and subject to the requirements of Table IWE-2500-1, Examination Category E-A.
Alternative Examination(s)
Currently, the IWE/IWL Program/Plan identifies 62 insulation panels at the interface between the concrete and the containment base mat (228-foot Elevation). Approximately one-third of the panels at the base mat interface will be removed and a general visual examination of the containment liner performed each examination period during the Second Ten-Year Interval. This will ensure that over the Second Ten-Year Interval, a 100% general visual examination of the liner at the base mat elevation will be performed. Attachment 6 provides details of that interface.
In addition, during the Second Ten-Year Interval, when an insulation panel at any elevation is removed for maintenance activities, a general visual examination of the liner beneath that panel will be performed.
Basis for Requesting Relief In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested for HBRSEP2, on the basis that the proposed alternative examinations/in conjunction with the examinations that have occurred during the First Ten-Year Interval, provide an acceptable level of quality and safety.
Table IWE-2500-1, Examination Category E-A does not address an insulated containment liner.
The containment liner at HBRSEP2, is partially insulated and covered by a stainless steel sheathing to provide for thermal protection of the liner during a design basis accident.
Justification for Granting Relief Relief is requested from the Code requirements for general visual examinations of the containment liner in areas that are insulated. Proposed alternative examinations provide an acceptable level of quality and safety.
The containment liner at HBRSEP2, is partially covered by insulation and stainless steel sheathing. The insulation and stainless steel sheathing form part of the defense-in-depth philosophy of the containment liner at HBRSEP2. The removal and reinstallation of the insulation sheathing panels has been determined to be time consuming and results in hardship and unusual difficulty.
During the First Ten-Year Interval, inspections were performed under 108 insulation panels that were removed. This included the planned removal of the entire inventory of 62 panels at the base mat elevation, for which both liner and moisture barrier inspections were performed. A total of 46 panels at higher elevations were removed for various reasons. Once removed, a liner inspection was performed.
A review of the data associated with the First Ten-Year Interval for the liner inspections indicated a degradation of the coating, which required coating removal and reapplication.
Subsequent ultrasonic and visual examination after coating removal revealed the minimum liner wall thickness was not violated and was acceptable to the procedural criteria. Liner coatings were reapplied and as-left examinations were performed prior to panel insulation and sheathing replacement.

Enclosure Page 30 of 60 The proposed alternative examinations provide an acceptable level of quality and safety.
Relief Request No. IWE/IWL-RR-02, Visual Examination of Moisture Barriers Code Requirements for Which Relief is Requested The ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Table IWE-2500-1, "Examination Categories," Examination Category E-A, Item Number E1.30, requires a visual examination of the containment moisture barrier, in accordance with Figure IWE-2500-1, "Examination Areas for Moisture Barriers," for Class MC, and metallic liners of Class CC, components.
The required method is a general visual examination of 100% of the moisture barrier materials intended to prevent intrusion of moisture against inaccessible areas of the pressure-retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces, which are not seal-welded. Containment moisture barrier materials include caulking, flashing, and other sealants used for this application. Deferral of the test to the end of the interval is not applicable due to the 100% per period requirement.
This relief was previously approved for the First Ten-Year IWE/IWL Inspection Interval under a Safety Evaluation Report (TAC No. MA4637) dated July 26, 1999.
Specific Relief Requested Relief is requested from performing general visual examinations, in accordance with ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Table IWE-2500-1, Examination Category E-A, on the containment moisture barriers. Proposed alternative examinations are provided below.
This request for relief is applicable to components classified as Class MC and subject to the requirements of Table IWE-2500-1, Examination Category E-A, at HBRSEP2.
Alternative Examination(s)
The IWE/IWL Program/Plan identifies 62 insulation panels at the interface between the concrete and the containment base mat (228-foot Elevation). Approximately one-third of the panels at the base mat interface will be removed and a general visual examination of the moisture barrier performed each examination period during the Second Ten-Year Interval. This will ensure that over the Second Ten-Year Interval, a 100% general visual examination of the moisture barrier will be performed.
Additionally, during the Second Ten-Year Interval, when an insulation panel on the 228-foot El.
is removed for maintenance activities, a general visual examination of the moisture barrier will be performed.
Basis for Requesting Relief In accordance with 10 CFR 50.5 5a(a)(3)(i), relief is requested for HBRSEP2, on the basis that the proposed alternative examinations, in conjunction with the examinations that have occurred during the First Ten-Year Interval, provide an acceptable level of quality and safety.
Table IWE-2500-1 Examination Category E-A does not address an insulated containment moisture barrier. The containment moisture barrier at HBRSEP2 is covered by stainless steel

Enclosure Page 31 of 60 sheathing and insulation to provide for thermal protection of the carbon steel liner during a design basis accident.
Justification for Granting Relief As shown in Figure IWE-2500-1, and noted in Table IWE-2500-1, moisture barrier materials are intended to prevent intrusion of moisture against inaccessible areas of the pressure-retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces which are not seal-welded. For HBRSEP2, the moisture barrier that meets this definition is the epoxy joint filler that interfaces with the concrete-to-containment liner interface at the 228-foot Elevation. Attachment 6 provides details of that interface.
The containment internal moisture barrier is covered with a layer of insulation and stainless steel sheathing. The removal and reinstallation of the insulation sheathing panels has been determined to result in hardship and unusual difficulty.
During the First Ten-Year Interval, 100% of the moisture barrier was inspected. A review of the data associated with the First Ten-Year Interval for the moisture barrier inspections indicated a degradation of the moisture barrier in some locations, which required removal and reapplication.
Visual examination of the liner after moisture barrier removal revealed that the minimum wall thickness of the liner behind the moisture barrier was not violated and was acceptable to the procedural criteria. Liner coatings and the moisture barrier were reapplied and as-left examinations were performed prior to panel insulation and sheathing replacement.
The proposed examination, which will ensure 100% moisture barrier inspection over the Second Ten-Year Interval, provides an acceptable level of quality and safety while not presenting an undue challenge to the moisture barrier insulation panels.
3.2.3   Results of recent IWE and IWL Examinations Recent IWE Examinations The results of recent IWE examinations is discussed in Section 3.4.1, IN 2010-12, Containment Liner Corrosion.
Recent IWL Examinations Examinations required by ASME B&PV Code, Section XI, Division 1, Subsection IWL (Inspection of concrete for the containment building) were performed the fall of 2013. The inspections were documented on "Visual examination of IWL (General)" Report Forms and had several newly reported adverse conditions. Per the requirements of ASME Section XI, a Registered Professional Engineer (RPE) with the ESG-0090N qualifications evaluated these conditions. The Items Identified have been screened by the RPE and are acceptable in that there are no operability concerns and the conditions do not affect the structural integrity or leak tightness of the HBRSEP2 containment.
VT-13-101, CONCRETE 0&deg; - 90&deg; DEGREES SURFACE - Reconfirmed previously reported condition from Report No. VT-10-001. No new conditions other than those previously identified and accepted.

Enclosure Page 32 of 60 VT-13-102, CONCRETE 90&deg; - 180&deg; DEGREES SURFACE - Between horizontal joints approximately 180 degrees, a horizontal rust material approximately 12" long, observed what appears to be a 1" board between forms 3rd and 4th from top 110 degrees, approximately 3" long partially covered by grout, and possibly an unpatched support hole at approximately 125 degrees four forms down from the top.
VT-13-103, CONCRETE 180&deg; - 270&deg; DEGREES SURFACE - Area at 230 degrees showed signs of further deterioration exposing underlying rebar.
VT-13-104, CONCRETE 270&deg; - 0&deg; DEGREES SURFACE - Concrete pop out with metal between 2nd and 3rd form from ground.
VT-13-107, CONCRETE DOME SURFACE - Reconfirmed previously reported condition from Report No. VT-10-007. No new conditions other than those previously identified and accepted.
R022 Disposition:
The recordable data from R020 has not changed. The separation at hooks #20 and 121 was minor with no active corrosion observed. The second hollow spot at hooks #29 and
       #30 was also at the dome/gutter interface. The concrete was not exposed and the depth of the hollow area appeared to be limited. Therefore, this 2nd hollow spot is also acceptable. The coatings have continued to degrade in limited areas on the dome. This potentially allows moisture to contact the concrete or exposes the concrete to the environment. In addition, another area of coating was missing adjacent to the dome ladder. Although the coating continues to degrade, corrosion or degradation of the concrete was not observed.

== Conclusion:==
No reinforcing steel was exposed and there was no evidence of active corrosion staining. These conditions do not affect the structural integrity of the containment structure and are therefore acceptable.
R024 Disposition The recordable data disposition from R022 has not changed except as noted below.
Coating at additional locations had deteriorated but the concrete at those locations did not identify any areas of concrete degradation. Although the amount of chipping on the concrete curb had increased, corrosion staining or rebar was not evident at these locations. This is acceptable. The chipping of the tie off hooks has no affect on the structural integrity of the containment structure. This is acceptable. It should be noted that several small areas of coating degradation on the normally inaccessible areas of the dome were evident. When viewed at a distance with binoculars, the exposed concrete in those areas did not exhibit any degradation.

== Conclusion:==
No reinforcing steel was exposed and there was no evidence of active corrosion staining. These conditions do not affect the structural Integrity of the containment structure and are therefore acceptable.
VT-13-109, CONCRETE CABLE VAULT ROOM NORTH SURFACE - Inspection revealed a pop out area approximately 3" x 7" with a depth of 5/8" and a crack-like Indication adjacent to both ends of the pop out extending 4" to 5" from pop out area.

Enclosure Page 33 of 60 VT-13-111, CONCRETE PURGE INLET ROOM SURFACE - Pop outs behind electrical equipment. Approximate size of pop out 5" x 3 1/4" x 1/2" deep and 2 1/2" x 2 1/4 x 14" deep.
VT-13-113, CONCRETE ROD CONTROL ROOM SURFACE - Reconfirmed previously reported condition from evaluation performed in 2007. No new conditions other than those previously identified and accepted.
R022 Disposition:
The recordable data from R020 has not changed. The 1 1/2 diameter hole did not appear to change from previous examination. The scaling documented in this examination was less than the recordable depth criteria with no active corrosion staining or rebar exposed.

== Conclusion:==
No reinforcing steel was exposed and there was no evidence of active corrosion staining. These conditions do not affect the structural integrity of the containment structure and are therefore acceptable.
R024 Disposition:
The recordable data disposition from R022 has not changed except as noted below. The horizontal portion of the foundation was included in this examination. The small voids meet the acceptance criteria.

== Conclusion:==
No reinforcing steel was exposed and there was no evidence of active corrosion staining. These conditions do not affect the structural integrity of the containment structure and are therefore acceptable.
The report forms above indicate either previously accepted items, original construction materials left from the initial concrete placement, unpatched support holes, minor rusting, and some pop outs. All are cosmetic in nature and will not affect the structural integrity or leak tightness of the containment. No action is required.
3.2.4    Integrated Leakage Rate Testing (ILRT) History Previous Type A tests confirmed that the HBRSEP2 reactor containment structure has leakage well under acceptance limits and represents minimal risk to increased leakage. Continued Type B and Type C testing for direct communication with containment atmosphere minimize this risk.
Also, the Inservice Inspection (IWE/IWL) program and maintenance rule monitoring provide confidence in containment integrity.
To date, seven operational Type A tests have been performed on HBRSEP2. There is considerable margin between these Type A test results and the TS 5.5.16 limit of 0.75 La, where La is equal to 0.1% by weight of the containment air per day at the peak accident pressure. These test results demonstrate that HBRSEP2 has a low leakage containment.
Table 3.2.4-1, ILRT Test Results Test Date              As Found Test Results             Results Adjusted to Pa
(% Weight per Day)                (% Weight per Day)
May 1974                       0.013 1                            0.013 2
Feb. 1978                      0.035                              0.049

Enclosure Page 34 of 60 Mar. 1982                      0.026 2                            0.037 Nov. 1984                      0.011 2                            0.016 April 1987                     0.041 2                            0.058 April 1992                   0.0602 1, 3                         0.0602 May 2007                     0.0244 1, 4                         0.0244 Note 1:         ILRT performed at Peak Containment Post LOCA pressure as identified in the plants TS in effect at the time of the test.
Note 2:         ILRT performed at 1/2 Pa and results are calculated at 1/2 Pa.
Note 3:         The test method used was the Absolute Method and the leakage rates were calculated using the Mass Point Analysis equations as described in ANSI/ANS 56.8-1987, "Containment System Leakage Testing Requirements.
Note 4:         The test method used was the Absolute Method and the leakage rates were calculated using the Mass Point Analysis equations as described in ANSI/ANS 56.8-1994, "Containment System Leakage Testing Requirements.
3.2.5   ILRT License Renewal Commitment To provide additional assurance of the tendons design capacity, testing at Integrated Leak Rate Test pressure, similar to the Structural Integrity Test performed in 1992, will be scheduled to coincide with Appendix J containment Integrated Leak Rate Testing conducted during the period of extended operation (required frequency in accordance with 10 CFR 50 Appendix J). The monitoring criteria for these tests will be limited to deformations and cracking associated with the vertical prestressed tendons, and will not include radial monitoring. Guidelines for performing the IWL examinations for these tests will include additional emphasis on looking for a pattern of horizontal cracks, and additional cracking in the discontinuity areas.
Procedures should include requirements for documentation and evaluation of test results showing structural integrity of the vertical prestressed tendons. Coordination with Civil Design group is required in order to ensure adequate controls are in place to complete the required test/inspections.
3.3     Containment Leakage Rate Testing Program, Type B and Type C Testing HBRSEP2 Types B and C testing program currently requires testing of electrical penetrations, airlocks, hatches, flanges, and containment isolation valves in accordance with 10 CFR Part 50, Appendix J, Option A. The results of the test program are used to demonstrate that proper maintenance and repairs are made on these components throughout their service life. The Types B and C testing program provides a means to protect the health and safety of plant personnel and the public by maintaining leakage from these components below appropriate limits. Per TS 5.5.16, the allowable maximum pathway total Types B and C leakage is 0.6 La where 0.6 La equals 91,490.78 sccm.
As discussed in NUREG-1493, Type B and Type C tests can identify the vast majority of all potential Containment leakage paths. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

Enclosure Page 35 of 60 As-Found Testing 10 CFR 50, Appendix J, Option A does not require As-found testing for Type B and Type C Penetrations.
Upon the implementation of the proposed amendments to HBRSEP2 Technical Specifications, As-Found LLRT testing will be required in accordance with the requirements of NEI 94-01 Revision 3-A, Section 10.2.1 for Type B Test Intervals, and Section 10.2.3 for Type C Test Interval.
Type B and C Acceptance Criteria 10 CFR 50, Appendix J, Option A, Acceptance Criterion.
The combined leakage rate for all penetrations and valves subject to Type B and C tests shall be less than 0.60 La. Leakage from containment isolation valves that are sealed with fluid from a seal system may be excluded when determining the combined leakage rate: Provided, That; (a) Such valves have been demonstrated to have fluid leakage rates that do not exceed those specified in the technical specifications or associated bases, and (b) The installed isolation valve seal-water system fluid inventory is sufficient to assure the sealing function for at least 30 days at a pressure of 1.10 Pa.
The IVSW system has been reviewed by the NRC and accepted as meeting the requirements of a seal system as defined in Appendix J. This review concluded that the IVSW system could be used in performing Type C tests. A description of the IVSW system is provided in Section 3.1.4.
Test results for valves that receive IVSW are not compared to the 0.6 La acceptance criteria of Appendix J Option A, Section III.C.3. Leakage from valves served by the IVSW system is excluded from this comparison. These test results are compared to the acceptance criteria of OST-933 and Technical Specification 3.6.8.6, which is based on a total allowable leakage value for the IVSW Tank. The allowable leakage is based on the UFSAR 6.8 criteria of 50-cc/hr./valve in.
Upon the implementation of the proposed amendments to HBRSEP2 Technical Specifications, the test results for valves that receive IVSW will continue to be exempt from the comparison to 0.6 La in accordance with NEI 94-01, Revision 3-A, Section 6.0, General Requirements, which provides the following relevant exemption:
An LLRT is not required for the following cases:
* Boundaries sealed with a qualified seal system Type B and Type C Test Results A review of the Type B and Type C test results from 2007 through 2015 for HBRSEP2 has shown an exceptional amount of margin between the actual As-Found (AF) and As-left (AL) outage summations and the regulatory requirements as described below:
* The As-Left leak rate average for HBRSEP2 shows an average of 15.3% of 0.6 La with a

Enclosure Page 36 of 60 high of 25.1% of 0.6 La.
Table 3.3.1-1 provides LLRT data summaries for HBRSEP2 since 2005 that encompasses the previous ILRT. This summary shows that there has been no failures that resulted in exceeding the TS 5.5.16 limit of 0.6 La (91,490.78 sccm) and demonstrates a history of successful tests.
The summations represent the high quality of maintenance of Type B and Type C tested components and the effective management of the Containment Leakage Rate Testing Program by the program owner.
Table 3.3-1, HBRSEP2 Types B and C LLRT As-Left Trend Summary RFO        RO-23       RO-24       RO-25     RO-26        RO-27        RO-28       RO-29 Fall       Spring       Fall   Summer        Spring         Fall       Spring 2005         2007       2008      2010         2012         2013         2015 As-Left      11985        18392      22972     16983        10436         9844         7122 sccm        sccm       sccm       sccm         sccm         sccm        sccm Fraction 0.131        0.201      0.251      0.186        0.114       0.108        0.079 of 0.6 La As shown in Table 3.3.1-1 above, the record keeping requirements for HBRSEP2 are different from other LARs requesting a permanent 15-year ILRT Interval with Containment Leakage Rate Testing Programs already on 10 CFR 50, Appendix J, Option B. 10 CFR 50, Appendix J, Option A and ANSI/ANS 56.8-1987 are not performance-based regulations and standards.
The recordkeeping requirements found in 10 CFR 50, Appendix J, Option A, Section V.B.2 are associated with the Type A test only and are as stated below:
For each periodic test, leakage test results from Type A, B, and C tests shall be included in the summary report. The summary report shall contain an analysis and interpretation of the Type A test results and a summary analysis of periodic Type B and Type C tests that were performed since the last type A test. Leakage test results from type A, B, and C tests that failed to meet the acceptance criteria of III.A.5(b), III.B.3, and III.C.3, respectively, shall be included in a separate accompanying summary report that includes an analysis and interpretation of the test data, the least squares fit analysis of the test data, the instrumentation error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria.
Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements shall also be included.
The requirements regarding as-found and as-left, and minimum and maximum pathway leakage rates as normally reported in other LARs was not contained in ANSI/ANS 56.8-1987, hence they were also not reported in this LAR as which would have been reported in other LARs requesting a permanent 15-year ILRT Interval with Containment Leakage Rate Testing Programs already on 10 CFR 50 Appendix J, Option B. With the adoption of 10 CFR 50 Appendix J Option B, and NEI 94-01 Revision 3-A and the conditions and limitations of NEI 94-01 Revision 2-A as proposed in this LAR and, the recording/reporting of As-Found, As-Left, Minimum and Maximum pathway leakage rates as stated in ANSI/ANS 56.8-2002 will become a requirement of the HBRSEP2 Containment Leakage Rate Testing Program.
With the adoption of the proposed TS Amendment, the recordkeeping requirements will be changed to meet the following requirements:

Enclosure Page 37 of 60 10 CFR 50 Appendix J Option B, Section IV:
The results of the preoperational and periodic Type A, B, and C tests must be documented to show that performance criteria for leakage have been met. The comparison to previous results of the performance of the overall containment system and of individual components within it must be documented to show that the test intervals established for the containment system and components within it are adequate.
These records must be available for inspection at plant sites.
NEI 94-01 Revision 3-A, Section 12.1, 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 the report and a corrective action plan developed to restore the margin to an acceptable level.
Table 3.3-2 identifies the components that have not demonstrated acceptable performance during the previous two outages for HBRSEP2:
Table 3.3-2, HBRSEP2 Types B and C LLRT Program Implementation Review Penetration      Initial  Admin Limit   As-left   Cause    Corrective     Scheduled Number /        LLRT        SCCM       SCCM        of        Action       Interval Component        SCCM                             Failure 2013 RO-28 4            2800        2000        736     Note 1     Note 1      Appendix J Option A WD-1713 WD-1793 Primary System Vent Header 42           3000        3000         53     Note 2     Note 2      Appendix J Option A V12-12 V12-13 Containment Vacuum Relief 2015 RO-29 None

Enclosure Page 38 of 60 Note 1: While performing the initial LLRT on Penetration 4 (Primary System Vent Header) during RO-28, the leakage rate exceeded the Administrative Limit of 2000 sccm. The actual leak rate was 2800 sccm.
Carbon corrosion residue was discovered in check valve WD-1713 interior components after it had been dissembled. A light film of the carbon residue was present on the valves seat as well. The residue was likely to be from the carbon steel piping that runs downstream to the Nitrogen Supply Drain Tank Isolation Valve. The carbon corrosion residue was cleaned off the seat of the valve and the valves spring and disc were replaced.
Post maintenance LLRT Surveillance was performed with satisfactory results of 736 sccm.
The valve failed the same test in Refueling Outage-26. Based on its history of failures the past three refueling cycles, check valve WD-1713 will now be inspected every two refueling cycles in accordance with the corrective action program.
Note 2: While performing the initial LLRT on Penetration 42 (Containment Vacuum Relief) during RO-28, the leakage rate reached the Administrative Limit of 3000 sccm.
Containment vacuum relief isolation valves V12-12 and V12-13 were removed from the containment penetration piping to allow for routing communication cables into containment during the outage. The valves were removed and re-installed under work orders 2063378 (V12-12) and 2063365 (V12-13). The valves were not disassembled as part of this removal and re-installation.
After re-installation, on October 23, 2013, the valves were tested in accordance with EST-135-2, Local Leak Rate Test of Vacuum Relief Valves. The test pressure of 44 psig could not be obtained due to excessive leakage through V12-12. A visual inspection of the valve identified that the vane was not completely closed, leaving a gap to the valve seat-sealing ring.
The V12-12 seal ring was replaced and the travel stops were adjusted to ensure proper closure in accordance with maintenance procedure CM-M-GNRL-VLV-0001. Post maintenance LLRT Surveillance was performed with satisfactory results of 53 sccm.
3.4     NRC Information Notices (INs) 3.4.1   IN 2010-12, Containment Liner Corrosion This IN provides examples of containment liner degradation caused by corrosion. Concrete reactor containments are typically lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions. The reactor containment is required to be operable as specified in plant technical specifications to limit the leakage of fission product radioactivity from the containment to the environment. The regulations at 10 CFR 50.55a, Codes and Standards, require the use of Subsection IWE of ASME Section XI to perform inservice inspections of containment components. The required inservice inspections include periodic visual examinations and limited volumetric examinations using ultrasonic thickness measurements. The containment components include the steel containment liner and integral

Enclosure Page 39 of 60 attachments for the concrete containment, containment personnel airlock and equipment hatch, penetration sleeves, moisture barriers, and pressure-retaining bolting. The NRC also requires licensees to perform leak rate testing of the containment pressure-retaining components and isolation valves according to 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, as specified in plant technical specifications. This operating experience highlights the importance of good quality assurance, housekeeping and high quality construction practices during construction operations in accordance with 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
Corrosion to the containment liner is not a new industry issue. Programs and procedures are in-place to inspect the containment liner and would identify any areas subject to corrosion.
Containment Liner Corrosion Operating Experience Summary, Technical Letter Report Revision 1, dated August 2, 2011, Section 2.4, Previous Assessments and Containment Operating Experience provided the following descriptions of previously identified containment liner and moisture barrier degradation for HBRSEP2:
* Robinson Unit 2 and Beaver Valley Unit 1 had areas of bulging and spot corrosion of the liner plate and degradation of the liner coatings. Surface cracking of concrete and deterioration of earlier patched concrete was observed (Ashar and Bagchi, 1995).
* In April 1992, Robinson Unit 2 was observed to have discoloration of the vertical portion of the containment liner indicating possible corrosion of the liner at an insulation joint (Ashar and Bagchi, 1995). This was also addressed in Information Notice 97-19, Liner Plate Corrosion in Concrete Containments.
* In December 1996, H.B. Robinson Unit 2 was found to have degraded caulking and insulation sheathing panels during a containment walkdown. The vertical portion of the containment liner at Robinson is protected by a polyvinyl chloride insulation material and a metal sheathing material. The licensee determined that a portion of this insulation sheathing material was loose and that some of the caulking between the sheathing panels was deteriorated. After examination during subsequent refueling outages, they determined that the protective coating for the containment liner was degraded and that while some corrosion of the containment liner had occurred, the liner still met design requirements (Braverman, et al., 2000).
* During the First Ten-Year Interval, 100% of the moisture barrier was inspected. A review of the data associated with the First Ten-Year Interval for the moisture barrier inspections indicated a degradation of the moisture barrier in some locations, which required removal and reapplication. Visual examination of the liner after moisture barrier removal revealed that the minimum wall thickness of the liner behind the moisture barrier was not violated and was acceptable to the procedural criteria. Liner coatings and the moisture barrier were reapplied and as-left examinations were performed prior to panel insulation and sheathing replacement.
* During the First Ten-Year Interval, inspections were performed under 108 insulation panels that were removed. This included the planned removal of the entire inventory of 62 panels at the base mat elevation, for which both liner and moisture barrier inspections

Enclosure Page 40 of 60 were performed. A total of 46 panels at higher elevations were removed for various reasons. Once removed, a liner inspection was performed.
A review of the data associated with the First Ten-Year Interval for the liner inspections indicated a degradation of the coating, which required coating removal and reapplication.
Subsequent ultrasonic and visual examination after coating removal revealed the minimum liner wall thickness was not violated and was acceptable to the procedural criteria. Liner coatings were reapplied and as-left examinations were performed prior to panel insulation and sheathing replacement.
The results of recent Second Ten-Year Interval inspections are described as follows:
* IWE Inspections of the CV dome produced a visual examination report of degraded coatings of the dome liner. CV dome degraded coatings documented by the IWE Inspections during R028 confirm degraded coatings found during coatings assessment walkdowns of previous outages. A minor increase in the amount of degraded coatings was identified. Most all of the degraded areas observed in the IWE Inspection of R028 have been previously documented by the coatings program. All degraded coatings on the dome liner must be documented and Incorporated Into the coatings exempt log to monitor the containment exempt coatings margin, and all coatings from the IWE Inspection not already contained In the coatings exempt log will be added.
The coatings program will continue to monitor and trend the containment dome degraded coatings and will Incorporate IWE Inspection data into the coatings exempt log.
* IWE Inspections of the CV wall liner during RO-26, RO-27 and RO-28 produced visual examination reports of degraded coatings, excessive corrosion/pitting, broken pins, gouging, bulging, and abnormal wear/discoloration of the wall liner. All degraded coating were restored. All VT-3 indications were dispositioned by evaluation. No base metal repairs were required.
3.4.2  IN 2014-07, Degradation of Leak Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner This IN was issued to inform addressees of issues identified by the NRC staff concerning degradation of floor weld leak-chase channel systems of steel containment shell and concrete containment metallic liner that could affect leak-tightness and aging management of containment structures.
IN 2014-07 described the leak chase channel system as follows:
Consists of steel channel sections that are fillet welded continuously over the entire bottom shell or liner seam welds and subdivided into zones, each zone with a test connection. Each test connection consists of a small carbon or stainless steel tube (less than 1-inch (2.5 centimeters) diameter) that penetrates through the back of the channel and is seal-welded to the channel steel. The tube extends up through the concrete floor slab to a small steel access (junction) box embedded in the floor slab. The steel tube, which may be encased in a pipe, projects up through the bottom of the access box with a threaded coupling connection welded to the top of the tube, allowing for pressurization

Enclosure Page 41 of 60 of the leak-chase channel.
IN 2014-07 describes a recessed box with a cover plate at floor level that allows for water to pool inside the recessed box and cause degradation. The HBRSEP2 system is not the same as the cited systems since it uses manifolds 3.5 feet above the floor instead of a capped floor access box. The manifolds are vented to containment atmosphere, however they are sealed passages and not conducive to flow to transport moisture.
The general review of NRC IN 2014-07 did identify a gap in the IWE program as described below:
The piping/tubing runs that go through the concrete have not been previously included in the IWE inspection program to verify that there is no corrosion or a breach near the floor level which would allow moisture to enter in a manner like the plants described in the IN.
The PPS to the Weld Channels was abandoned in the mid 1970s and per the 10 CFR 50.55(b)(2)(ix)(A) requirement; the piping/tubing should be examined closely enough to say that there is not a breach.
To provide reasonable assurance that aging effects of the containment liner are being managed, the IWE program was revised to perform visual inspections of accessible tubing in the PPS system from manifold to floor of the CV.
3.4.3    IN 92-20, Inadequate Local Leak Rate Testing NRC IN 92-20 was issued to alert licensees to problems with local leak rate testing of two-ply stainless steel bellows used on piping penetrations at some plants. Specifically, local leak rate testing could not be relied upon to accurately measure the leakage rate that would occur under accident conditions since, during testing, the two plies in the bellows were in contact with each other, restricting the flow of the test medium to the crack locations. Any two-ply bellows of similar construction may be susceptible to this problem.
This is not applicable to HBRSEP2. There are both single and two-ply bellows assemblies, which are also Containment isolation barriers, installed at HBRSEP2. The difference is that the bellows assemblies are tested using the original Penetration Pressurization System (PPS).
ESR/MOD 95-00888 removed the automatic continuous pressurization and monitoring features of this system. It is now only used during power operation to test the personnel hatch and during outages to test containment penetrations (local leak rate tests) such as the penetration bellows assemblies.
The PPS test connection is located on the outboard penetration sleeve for each bellows assembly and not between the plies on the two-ply assemblies so there is no restriction in airflow between the inboard and outboard bellows assemblies.
The leakage rate performance of the bellows assemblies subject to Type B testing is shown in Table 3.4.3-1 below:

Enclosure Page 42 of 60 Table 3.4.3-1, HBRSEP2 Type B Bellows As-Left Trend Summary RO-23         RO-24       RO-25      RO-26       RO-27      RO-28       RO-29 Fall        Spring        Fall    Summer       Spring       Fall     Summer 2005          2007        2008       2010        2012       2013        2015 0        1355.46     3002.01    2723.21     2478.37      2054        1687 sccm         sccm        sccm        sccm        sccm       sccm       sccm 3.5      Supplemental Inspections In the Safety Evaluation Report for NEI 94-01 Revision 2-A, the NRC stated the following requirement for the performance of Supplemental Visual Inspections in SER Section 3.1.1.3, Adequacy of Pre-Test Inspections (Visual Examinations):
Subsections IWE and IWL (References 13 and 14) of the ASME Code, Section XI, as incorporated by reference in 10 CFR 50.55a, require general visual examinations two times within a 10-year interval for concrete components (Subsection IWL), and three times within a 10-year interval for steel components (Subsection IWE). To avoid duplication or deletion of examinations, licensees using NEI TR 94-01, Revision 2, have to develop a schedule for containment inspections that satisfy the provisions of Section 9.2.3.2 of this TR and ASME Code, Section XI, Subsection IWE and IWL requirements.
The performance of a supplemental inspection is a requirement of the HBRSEP2 Containment Integrated Leak Test surveillance procedure. A pretest visual inspection of accessible interior and exterior surfaces of Containment structures and components for evidence of deterioration is completed and documented in the surveillance. The purpose of this inspection is to identify evidence of structural deterioration that might affect either the primary containment structural integrity or leak lightness. Any irregularities such as cracking, peeling, delamination, corrosion, and structural deterioration shall be recorded and evaluated, or repaired as required.
This inspection requirement may be met in part OR its entirety by completing the containment building IWE and IWL inspections OR by walking down the Containment per the instructions contained in the Containment Integrated Leak Test surveillance procedure.
3.6      Limitations and Conditions 3.6.1    Limitations and Conditions Applicable to NEI 94-01, Revision 2-A The NRC staff found that the use of NEI TR 94-01, Revision 2, was acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT surveillance interval to 15 years, provided the following conditions as listed in Table 3.6-1 were satisfied:
Table 3.6-1, NEI 94-01 Revision 2-A Limitations and Conditions

Enclosure Page 43 of 60 Limitation/Condition (From Section 4.0 of SE)                        HBRSEP2 Response For calculating the Type A leakage rate, the    HBRSEP2 will utilize the definition in licensee should use the definition in the NEI   NEI 94-01 Revision 2-A, Section 5.0.
TR 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002. (Refer to SE                Reference Table 3.2.3-1, ILRT Test Section 3.1.1.1.)                               Results, Notes (3) and (4)
The licensee submits a schedule of              Reference Tables 3.2.2-1, 3.2.2-2 and containment inspections to be performed prior   3.5-1 of this submittal.
to and between Type A tests. (Refer to SE Section 3.1.1.3.)
The licensee addresses the areas of the         Reference Section 3.2.2, Inaccessible containment structure potentially subjected to  Class MC Areas, of this submittal.
degradation. (Refer to SE Section 3.1.3.)
Reference Section 3.2.2, IWE/IWL augmented Examinations and Attachment 5, E-C AUGMENT Plan Tables for the Second Ten-Year Interval.
The licensee addresses any tests and            HBRSEP2 steam generator inspections performed following major           replacements were completed in 1984 modifications to the containment structure, as  using the equipment hatch.
applicable. (Refer to SE Section 3.1.4.)
There are no planned modifications for HBRSEP2 that will require a Type A test prior to the next scheduled Type A test proposed under this LAR.
There is no anticipated addition or removal of plant hardware within the containment building, which could affect its leak-tightness.
The normal Type A test interval should be less  HBRSEP2 will follow the requirements than 15 years. If a licensee has to utilize the of NEI 94-01 Revision 2-A, Section 9.1.
provision of Section 9.1 of NEI TR 94-01,       In accordance with the requirements of Revision 2, related to extending the ILRT       94-01 Revision 2-A, SER Section interval beyond 15 years, the licensee must     3.1.1.2, HBRSEP2 will also demonstrate demonstrate to the NRC staff that it is an      to the NRC staff that an unforeseen unforeseen emergent condition. (Refer to SE     emergent condition exists in the event Section 3.1.1.2.)                               an extension beyond the 15-year interval is required.

Enclosure Page 44 of 60 For plants licensed under 10 CFR Part 52,           Not applicable. HBRSEP2 was not applications requesting a permanent extension       licensed under 10 CFR Part 52.
of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.
3.6.2    Limitations and Conditions Applicable to NEI 94-01 Revision 3-A The NRC staff found that the guidance in NEI TR 94-01, Revision 3, was acceptable for referencing by licensees in the implementation for the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J. However, the NRC staff identified two conditions on the use of NEI TR 94-01, Revision 3 (Reference NEI 94-01 Revision 3-A, NRC SER 4.0, Limitations and Conditions):
Topical Report Condition 1 NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI TR 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR MSIVs), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.
Response to Condition 1 Condition 1 presents three (3) separate issues that are required to be addressed as follows:
* ISSUE 1 - The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.
* ISSUE 2 - In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.
* ISSUE 3 - Use of the allowed 9-month extension for eligible Type C valves is only authorized for non-routine emergent conditions.

Enclosure Page 45 of 60 Response to Condition 1, Issue 1 The post-outage report shall include the margin between the Type B and Type C Minimum Pathway Leak Rate (MNPLR) summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.60 La.
Response to Condition 1, Issue 2 When the potential leakage understatement adjusted Types B and C MNPLR total is greater than the HBRSEP2 administrative leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the HBRSEP2 administrative limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.
Response to Condition 1, Issue 3 HBRSEP2 will apply the 9-month grace period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.
Topical Report Condition 2 The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition-monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 refueling outages. Type C tests involve valves, which in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for.
Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.
When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total, and must be included in a

Enclosure Page 46 of 60 licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.
Response to Condition 2 Condition 2 presents two (2) separate issues that are required to be addressed as follows:
* ISSUE 1 - Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.
* ISSUE 2 - When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.
Response to Condition 2, Issue 1 The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25%
in the LLRT periodicity. As such, HBRSEP2 will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate, which will increase the As-Left leakage total 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, if the MNPLR is greater than the HBRSEP2 administrative 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 HBRSEP2 administrative leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.
Response to Condition 2, Issue 2 If the potential leakage understatement adjusted leak rate MNPLR is less than the HBRSEP2 administrative leakage summation limit of 0.50 La, then the acceptability of the greater than a 60 month test interval up to the 75-month LLRT extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

Enclosure Page 47 of 60 In addition to Condition 1, Parts 1 and 2, which deal with the MNPLR Type B and C summation margin, NEI 94-01, Revision 3-A also has a margin related requirement as contained in Section 12.1, Report Requirements.
A post-outage report shall be prepared presenting results of the previous 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.
At HBRSEP2, in the event an adverse trend in the aforementioned potential leakage understatement adjusted Type B and C summation is identified, and then an analysis and determination of a corrective action plan shall be prepared to restore the trend and associated margin to an acceptable level. The corrective action plan shall focus on those components which have contributed the most to the adverse trend in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.
At HBRSEP2, an adverse trend is defined as three (3) consecutive increases in the final pre-RCS Mode Change Type B and C MNPLR leakage summation values, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.
3.7     Evaluation of Risk Impact 3.7.1   Methodology 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 HBRSEP2. 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 RG 1.200 as applied to ILRT interval extensions, and risk insights in support of a request for a plants licensing basis as outlined in 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 1009325, Revision 2-A of EPRI 1018243
[Reference 24].
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, is acceptable for referencing by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the

Enclosure Page 48 of 60 limitations and conditions noted in Section 4.0 of the SE. Table 3.7.1-1 addresses each of the four limitations and conditions for the use of EPRI 1009325, Revision 2.
Table 3.7.1-1, EPRI Report No.TR-1009325 Revision 2 Limitations and Conditions Limitation/Condition                          HBRSEP2 Response (From Section 4.2 of SE) 1.The licensee submits documentation          The technical adequacy of the HBRSEP2 indicating that the technical adequacy      PRA is consistent with the requirements of of their PRA is consistent with the         Regulatory Guide 1.200 as is relevant to requirements of RG 1.200 relevant to        this ILRT interval extension, as detailed in the ILRT extension.                         Attachment 4 of this submittal, PRA Risk Assessment for Extending ILRT Interval to 15 Years, Attachment 1.
: 2. The licensee submits documentation         EPRI Report No. 1009325, Revision 2-A, indicating that the estimated risk          incorporates these population dose and increase associated with permanently        CCFP acceptance guidelines, and these extending the ILRT surveillance             guidelines have been used for the interval to 15 years is small, and          HBRSEP2 plant specific assessments.
consistent with the clarification provided in Section 3.2.4.5 of this SE.
Specifically, a small increase in The increase in population dose is 0.020 population dose should be defined as person-rem/year.
an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive.
In addition, a small increase in CCFP      The increase in CCFP is 0.829%. The should be defined as a value               increase proved to be below 1.5 marginally greater than that accepted      percentage points and thus is considered in a previous one-time ILRT extension      to be small.
requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point.

Enclosure Page 49 of 60
: 3. The methodology in EPRI Report No.          EPRI Report No. 1009325, Revision 2-A, 1009325, Revision 2, is acceptable           incorporated the use of 100 La as the except for the calculation of the            average leak rate for the pre-existing increase in expected population dose         containment large leakage rate accident (per year of reactor operation). In order    case (accident case 3b), and this value to make the methodology acceptable,          has been used in the HBRSEP2 plant the average leak rate accident case          specific risk assessment.
(accident case 3b) used by the licensees shall be 100 La instead of 35 La
: 4. A licensee amendment request (LAR)          For HBRSEP2, containment over-pressure is required in instances where               is NOT relied upon for emergency core containment over-pressure is relied          cooling system (ECCS) performance.
upon for emergency core cooling system (ECCS) performance.
3.7.2    Summary of Internal Events PRA Quality Statement for Permanent 15-Year ILRT Extension Internal Events PRA Model The HBRSEP2 Internal Events PRA model RNP_12 is utilized to calculate CDF, LERF, and PDSs for the permanent 15-year ILRT extension. Any elements of the supporting requirements detailed in ASME/ANS RA-Sa-2009 that could be significantly affected by the application are required to meet Capability Category II requirements.
The Internal Events PRA provides an adequate base model for the development of the permanent 15-year ILRT extension. While there has been one instance of a change to the internal events model since the 2010 peer review that could be considered a model upgrade, this change is specifically applicable to the loss of component cooling water (CCW) initiating event and does not affect the permanent 15-year ILRT extension results.
To ensure that the current Internal Events PRA model remains an accurate reflection of the as-built, as-operated plant, the following configuration control activities are routinely performed:
* Design changes and procedure changes are reviewed for their impact on the Internal Events PRA model. PRA screening is required for all design and procedure changes.
* New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.
* Plant-specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated based upon reviews of plant program data, particularly data supporting the Maintenance Rule.

Enclosure Page 50 of 60 The HBRSEP2 Internal Events PRA is based on a detailed model of the plant developed from the Individual Plant Examination for Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities." The model is maintained and updated in accordance with HBRSEP2 procedures, and has been updated to meet the ASME PRA Standard and Revision 2 of RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities."
In accordance with RG 1.200, Rev. 2 and ASME/ANS RA-Sa-2009, a full scope Peer Review of the Internal Events model of the HBRSEP2 PRA was conducted in May 2010 by six PRA experts (Attachment U of Reference 29). The review provided findings and suggestions regarding the model and identified 25 supporting requirements within the internal events portion of the model that did not meet Capability Category II. These findings were either resolved by additional analysis and included in the quantitative results, or evaluated with their impact on the applicable risk evaluation. The dispositions for these findings are presented in Table 1, Internal Events PRA Peer Review - Facts and Observations. All findings, which had significant impact on this analysis, have been addressed. The ILRT application was determined to be an application requiring a Capability Category II PRA model per the RG 1.200 criteria, Revision 2.
This is based on the requirement for numerical results for CDF and LERF to determine the risk impact of the requested change and the fact that this change is risk-informed, not risk based. of this submittal, PRA Risk Assessment for Extending ILRT Interval to 15 Years, , Table 1 includes discussion of all findings from the industry peer review along with the assessment and evaluation of the finding that shows that they have either been addressed or have no material impact on the ILRT interval extension request.
Following disposition, all of the SRs evaluated Met Capability Category II or better.
Fire PRA Quality Statement for Permanent 15-Year ILRT Extension In accordance with RG 1.205 position 4.3: The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard supporting requirements important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable.
The HBRSEP2 Internal Events model is also updated to support the HBRSEP2 Fire PRA. The HBRSEP2 Combined Internal Events and Fire PRA was peer reviewed during the period of March 2013 (Attachment V of Reference 29). The peer review was conducted by a team of industry personnel (utility and vendor). The Westinghouse Owners Group performed the review and has documented the outcome via LTR-RAM-13-06, Fire PRA Peer Review of the H. B.
Robinson Nuclear Plant Fire Probabilistic Risk Assessment against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard. As noted in LTR-RAM-13-06, the HBRSEP2 Fire PRA was found to be consistent with the ASME/ANS Standard and is suitable for supporting risk-informed applications.
The peer review team noted a number of Facts and Observations (F&Os). As documented in LTR-RAM-13-06, 85% of the Supporting Requirements (SRs) were assessed at Capability Category II or higher. Approximately eighteen Finding level and nine Suggestion level F&Os were identified during the peer review conducted in March 2013. Duke Energy recognized that CDF and LERF were relatively high, as noted in LTR-RAM-13-06. Based on the CDF and LERF values at the time of the initial peer review, coupled with the number of findings associated with

Enclosure Page 51 of 60 the Fire Scenario Selection (FSS) Technical Element (18), Duke Energy decided to have a focused peer review.
The focused peer review was conducted during the period of July 2013 and evaluated the FSS Technical Element based on refinements to approved methodologies and updated documentation. The focused peer review was conducted by Frederick Mowrer (C P Fire, LLC) and Bijan Najafi (JENSEN HUGHES, Inc.) and is documented via JENSEN HUGHES Calculation No. 0004-0042-415-RPT-001, Robinson Nuclear Plant Fire PRA Focused Peer Review, Revision 0. As noted in LTR-RAM-13-06 and JENSEN HUGHES Calculation No. 0004-0042-415-RPT-001, the Fire PRA does apply the methodologies outlined in NUREG/CR-6850 correctly, is consistent with the ASME/ANS Standard and is applicable for supporting risk informed applications. Although several of the initial F&Os were resolved, seven new findings and three new suggestions were identified during the focused-scope peer review. of this submittal, PRA Risk Assessment for Extending ILRT Interval to 15 Years, , Table 2 documents the Finding level F&Os associated with both the initial and focused peer reviews. All but two findings have been dispositioned as meeting Capability Category II or better. FSS-E3-01, which was assessed as meeting Category I, involved providing a mean value of, and statistical representation of, the uncertainty intervals for the parameters used for modeling the significant fire scenarios. Although no change has yet been made that would improve the Capability Category assessment, HBRSEP2 considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because documenting the statistical representation of uncertainty intervals will not change the quantified risk metrics.
Supporting Requirement FSS-E1-01 has not been resolved, but this is a documentation issue that does not impact quantification. These two SRs not being resolved does not impact the ILRT extension analysis. The Fire PRA is adequate to support the ILRT extension.
3.7.3   Summary of Plant-Specific Risk Assessment Results The risk impact of permanently extending the Type A ILRT test frequency to once in fifteen years is as follows:
* RG.174 (Reference 4) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of Core Damage Frequency (CDF) 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 Large Early Release Frequency (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 9.36E-8/year using the EPRI guidance. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174.
* 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 0.020 person-rem/year. EPRI Report No. 1009325, Revision 2-A (Reference 20) 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

Enclosure Page 52 of 60 when compared to other severe accident risks is negligible.
* The increase in the conditional containment failure from the 3-in-10-year interval to 1-in-15-year interval is 0.545%. EPRI Report No. 1009325, Revision 2-A (Reference 20) states that increases in Conditional Containment Failure Probability (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 HBRSEP2 risk profile.
3.7.4   Previous Assessments The NRC in NUREG-1493 (Reference 7) has previously concluded that:
* Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.
* Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.
The findings for HBRSEP2 confirm these general findings on a plant-specific basis considering the severe accidents evaluated for HBRSEP2, the HBRSEP2 containment failure modes, and the local population surrounding HBRSEP2.
Details of the HBRSEP2 risk assessment are contained in Attachment 4 of this submittal.

==4.0     REGULATORY EVALUATION==

4.1     Applicable Regulatory Requirements/ Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. 10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR Part 50, "Leakage Rate Testing of Containment of Water Cooled Nuclear Power Plants." Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.
The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B and Type C testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of 10 CFR

Enclosure Page 53 of 60 Part 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type B and Type C test frequencies will not directly result in an increase in containment leakage.
EPRI TR-1009325, Revision 2, provided a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance. NEI 94 01, Revision 3-A, Section 9.2.3.1 states that Type A ILRT intervals of up to 15 years are allowed by this guideline. The Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI Report 1018243 (Formerly TR-1009325, Revision 2) indicates that, in general, the risk impact associated with ILRT interval extensions for intervals up to 15 years is small.
However, plant-specific confirmatory analyses are required.
The NRC staff reviewed NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2.
For NEI TR 94-01, Revision 2, the NRC staff determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163. The NRC staff finds that the Type A testing methodology as described in ANSI/ ANS-56.8-2002, and the modified testing frequencies recommended by NEI TR 94- 01, Revision 2, serves to ensure continued leakage integrity of the containment structure. Type B and Type C testing ensures that individual penetrations are essentially leak tight.
In addition, aggregate Type B and Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.
For EPRI Report No. 1009325, Revision 2, a risk-informed methodology using plant specific risk insights and industry ILRT performance data to revise ILRT surveillance frequencies, the NRC staff finds that the proposed methodology satisfies the key principles of risk-informed decision making applied to changes to TSs as delineated in RG 1.177 and RG 1.174. The NRC staff, therefore, found that this guidance was acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.2 of the Safety Evaluation Report (SER).
The NRC staff reviewed NEI TR 94-01, Revision 3, and determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J, as modified by the conditions and limitations summarized in Section 4.0 of the associated Safety Evaluation. This guidance included provisions for extending Type C LLRT intervals up to 75 months. Type C testing ensures that individual containment isolation valves are essentially leak tight. In addition, aggregate Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The NRC staff, therefore, found that this guidance, as modified to include two limitations and conditions, was acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing. Any applicant may reference NEI TR 94-01, Revision 3, as modified by the associated SER and approved by the NRC, and the conditions and limitations specified in NEI 94- 01, Revision 2-A, dated October 2008, in a licensing action to satisfy the requirements of Option B to 10 CFR Part 50, Appendix J.
4.2      Precedent

Enclosure Page 54 of 60 This LAR is similar in nature to the following license amendments previously authorized by the NRC to extend the Type A test frequency to 15 years and the Type C test frequency to 75 months:
* Surry Power Station, Units 1 and 2 (Reference 21)
* Donald C. Cook Nuclear Plant, Units 1 and 2 (Reference 22)
* Beaver Valley Power Station, Unit Nos. 1 and 2 (Reference 23)
* Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (Reference 24)
* Peach Bottom Atomic Power Station, Units 2 and 3 (Reference 27)
This license amendment request is similar in nature to the following license amendment previously authorized by the NRC to adopt TSTF Technical Change Traveler 52, Revision 3, to implement Option B of 10 CFR 50 Appendix J:
* Oconee Nuclear Station, Units 1, 2, and 3 (Reference 25) 4.3     Significant Hazards Consideration Duke Energy Progress has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
: 1.      Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment to the Technical Specifications (TS) involves the extension of the H. B. Robinson Steam Electric Plant Unit No. 2 (HBRSEP2) Type A containment test interval to 15 years, the extension of the Type B test intervals to 120 months for selected components, and the extension of the Type C test interval to 75 months for selected components. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test.
The current Type B test interval of each reactor shutdown for refueling but in no case at intervals greater than 2 years would be extended on a performance basis to no longer than 120 months. The current Type C test interval of each reactor shutdown for refueling but in no case at intervals greater than 2 years would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. The containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen years, measured, as an increase

Enclosure Page 55 of 60 to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.020 person-rem/year. The Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2-A, states that a very small population dose is defined as an increase of  1.0 person-rem per year, or  1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended integrated leak rate test (ILRT) intervals. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.
As documented in NUREG-1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The HBRSEP2 Type A test history supports this conclusion.
The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) activity based, and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with the American Society of Mechanical Engineers (ASME) Section XI, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated.
The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.
Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment to the TS involves the extension of the HBRSEP2 Type A containment test interval to 15 years, the Type B test interval to 120 months for selected components and the extension of the Type C test interval to 75 months for selected components. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

Enclosure Page 56 of 60 The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3.      Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment to TS 5.5.16 involves the extension of the HBRSEP2 Type A containment test interval to 15 years, the Type B test interval to 120 months for selected components and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.
The proposed change involves only the extension of the interval between Type A containment leak rate tests, Type B tests and Type C tests for HBRSEP2. The proposed surveillance interval extension is bounded by the 15-year ILRT interval, the 120-month Type B interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Types B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI, TS and the Maintenance Rule serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing.
The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A, Type B and Type C test intervals.
The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Duke Energy Progress concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c),
and, accordingly, a finding of no significant hazards consideration is justified.

Enclosure Page 57 of 60 4.4      Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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, describe an NRC-accepted approach for implementing the performance-based requirements of 10 CFR Part 50, Appendix J, Option B. It incorporated the regulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals to 15 years, Type B test intervals to 120 months, and Type C test intervals to 75 months. NEI 94-01, Revision 3-A delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies. HBRSEP2 is adopting the guidance of NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, for the HBRSEP2, 10 CFR Part 50, Appendix J Testing Program Plan.
Based on the previous ILRT tests conducted at HBRSEP2, it may be concluded that the permanent extension of the containment ILRT interval from 10 to 15 years represents minimal risk to increased leakage. The risk is minimized by continued Type B and Type C testing performed in accordance with Option B of 10 CFR Part 50, Appendix J and the overlapping inspection activities performed as part of the following HBRSEP2 inspection programs:
* Containment Inservice Inspection Program (IWE/IWL)
* Primary Containment Coatings Condition Assessment This experience is supplemented by risk analysis studies, including the HBRSEP2 risk analysis provided in Attachment 4. The findings of the risk assessment confirm the general findings of previous studies, on a plant-specific basis, that extending the ILRT interval from 10 to 15 years results in a very small change to the HBRSEP2 risk profiles.

==5.0      ENVIRONMENTAL CONSIDERATION==

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

==6.0      REFERENCES==

Enclosure Page 58 of 60
: 1. RG 1.163, Performance-Based Containment Leak-Test Program, September 1995.
: 2. NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 2012.
: 3. NEI 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, October 2008.
: 4. RG 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, May 2011.
: 5. RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009.
: 6. NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 1995.
: 7. NUREG-1493, Performance-Based Containment Leak-Test Program, January 1995.
: 8. EPRI TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, August 1994.
: 9. Letter from M. J. Maxin (NRC) to J. C. Butler (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (TAC No.
MC9663), dated June 25, 2008.
: 10. Letter from S. Bahadur (NRC) to B. Bradley (NEI), Final Safety Evaluation of Nuclear Energy Institute (NEI) Report 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (TAC No. ME2164), dated June 8, 2012.
: 11. Industry/TSTF Standard Technical Specification Change Traveler, Implement 10 CFR 50, Appendix J, Option B, TSTF-52 Revision 3 (ML040400371)
: 12. Letter from NRC (Ram Subbaratnam) to Carolina Power & Light (J. Moyer), H. B.
Robinson Steam Electric Plant Unit 2 - Issuance of Amendment [No. 193] - Technical Specification Change Regarding One-Time Extension of Containment Type A Test Interval, H. B. Robinson Steam Electric Plant, Unit No. 2 (TAC No. MB4658) dated September 16, 2002 (ML022690765)
: 13. Letter to Caroliina Power & Light (J. W. Moyer) from NRC (C. Patel) dated February 11, 2004. (ML040430023). H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of an Amendment Re: Containment Integrated Leak Rate Test (TAC NO. MB9662).
: 14. ML071070170, Letter to T. Walt from C. Patel (NRC) dated June 15, 2007. H. B.
Robinson Steam Electric Plant, Unit No. 2  Issuance of an Amendment for Technical

Enclosure Page 59 of 60 Specifications Changes Related to Containment Peak Pressure (TAC NO. MD2682)
: 15. ML0082210549, Letter to T. Walt from M. Vaaler (NRC) dated October 3, 2008. H. B.
Robinson Steam Electric Plant, Unit No. 2 -Issuance of Amendment Regarding Changes to the Technical Specifications Related to the Isolation Valve Seal Water System (TAC NO. MD7469).
: 16. ML12174A010, Letter to W. Gideon from D. Broaddus (NRC) dated July 13, 2012. H. B.
Robinson Steam Electric Plant, Unit No. 2 - Relief Request-3 for the Fifth 10-Year Interval Inservice Testing Program Plan (TAC NO. ME8260).
: 17. EPRI Report 1003102, Guideline on Nuclear Safety-Related Coatings, Revision 1 (formerly TR-109937).
: 18. Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Rev.
4, Developed for NEI by EPRI and Data Systems and Solutions, November 2001.
: 19. Letter from Calvert Cliffs Nuclear Power Plant (Mr. C. H. Cruse) to NRC (Document Control Desk), Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Docket No. 50-317, dated March 27, 2002. (ML020920100)
: 20. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA: 2008.
: 21. ML14148A235, Letter to D. Heacock from S. Williams (NRC) dated July 3, 2014. Surry Power Station, Units 1 And 2- Issuance of Amendment Regarding the Containment Type A And Type C Leak Rate Tests.
: 22. ML15072A264, Letter to L. Weber from A. Dietrich (NRC) dated March 30, 2015. Donald C. Cook Nuclear Plant, Units 1 And 2 -Issuance of Amendments Re: Containment Leakage Rate Testing Program..
: 23. ML15078A058, Letter to E. Larson from T. Lamb (NRC) dated April 8, 2015. Beaver Valley Power Station, Unit Nos. 1 And 2 -Issuance of Amendment Re: License Amendment Request to Extend Containment Leakage Rate Test Frequency.
: 24. ML15154A661, Letter to G. Gellrich from A. Chereskin (NRC) dated July 16, 2015.
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 And 2 -Issuance of Amendments Re:
Extension Of Containment Leakage Rate Testing Frequency.
: 25. ML11186A906, Letter to P. Gillespie from J. Stang (NRC) dated July 28, 2011. Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding A Proposed Change To The Technical Specifications To Adopt Technical Specification Task Force (TSTF) Technical Change Traveler 52, Revision 3, To Implement Option B of Appendix J To Title 10 of the Code of Federal Regulations, Part 50 (TAC NOS. ME4557, ME4558, AND ME4559).

Enclosure Page 60 of 60
: 26. EPRI 1018243, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, Revision 2-A of 1009325.
: 27. ML15196A559, Letter to B. Hanson from R. Ennis (NRC) dated September 8, 2015.
Peach Bottom Atomic Power Station, Units 2 And 3 - Issuance of Amendments Re:
Extension of Type A and Type C Leak Rate Test Frequencies (TAC NOS. MF5172 AND MF5173).
: 28. Letter to J. Jones (CP&L) from A Schwencer (NRC) dated April 23, 1979. Safety Evaluation Report of the Isolation Valve Seal Water System, H. B. Robinson Unit 2 Docket No. 50-261.
: 29. Letter from W. R. Gideon (Duke Energy, H. B. Robinson) to U. S. Nuclear Regulatory Commission, Docket No. 50-261, ML13267A211, September 16, 2013. License Amendment Request (LAR) To Adopt NFPA 805 Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants (2001 Edition).

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 5 Pages (including this cover page)
PROPOSED TECHNICAL SPECIFICATION CHANGES

Containment 3.6.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                       FREQUENCY SR 3.6.1.1      Perform required Type B and C leakage rate testing       --------NOTE--------
except for containment air lock testing, in accordance   SR 3.0.2 is not Insert 1       with 10 CFR 50, Appendix J, Option A, as modified        applicable by approved exemptions.                                  ------------------------
Insert 2       The leakage rate acceptance criterion is  1.0 L a . In accordance with However, during the first unit startup following testing 10 CFR 50, performed in accordance with 10 CFR 50, Appendix         Appendix J, J, Option A, as modified by approved exemptions,         Option A, as              Insert 2 the leakage rate acceptance criterion is < 0.6 L a for   modified by the Type B and Type C tests.                             approved exemptions SR 3.6.1.2      Verify containment structural integrity in accordance    In accordance with the Containment Tendon Surveillance Program.        with the Containment Tendon Surveillance Program SR 3.6.1.3      Perform required visual examinations and Type A          In accordance with leakage rate testing, in accordance with the             the Containment Containment Leakage Rate Testing Program.                Leakage Rate Testing Program Insert 1:
visual examinations and Insert 2:
the Containment Leakage Rate Testing Program HBRSEP Unit No. 2                           3.6-2                         Amendment No. 176

Containment Air Lock 3.6.2 ACTIONS CONDITION             REQUIRED ACTION                    COMPLETION TIME B.  (continued)            B.3    ----------NOTE--------------
Air lock doors in high radiation areas may be verified locked closed by administrative means.
Verify an OPERABLE               Once per 31 days door is locked closed.
C. Containment air lock   C.1I   Initiate action to               Immediately inoperable for reasons        evaluate overall other than Condition A        containment leakage or B.                         rate per LCO 3.6.1.
Delete Typo            AND C.2    Verify a door is closed in       1 hour the air lock.
AND C.3    Restore air lock to              24 hours OPERABLE status.
D. Required Action and    D.1    Be in MODE 3.                    6 hours associated Completion Time not met.          AND D.2    Be in MODE 5.                    36 hours HBRSEP Unit No. 2                  3.6-5                              Amendment No. 176

Containment Air Lock 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                   FREQUENCY SR 3.6.2.1      ---------------------------NOTES---------------------------------
: 1.      An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
Insert 1
: 2.      Results shall be evaluated against acceptance criteria of SR 3.6.1.1, in accordance with 10                   ---------NOTE-------
CFR 50, Appendix J, Option A, as modified by                    SR 3.0.2 is not approved exemptions.                                            applicable Perform required air lock leakage rate testing in                       In accordance with accordance with 10 CFR 50, Appendix J, Option A,                        10 CFR 50, as modified by approved exemptions.                                     Appendix J, Insert 2                                                                                Option A, as             Insert 2 modified by approved exemptions.
SR 3.6.2.2      Verify only one door in the air lock can be                             24 months opened at a time.
Insert 1:
applicable to Insert 2:
the Containment Leakage Rate Testing Program.
HBRSEP Unit No. 2                                  3.6-6                                 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16        Containment Leakage Rate Testing Program A program      This program provides controls for implementation of the leakage rate testing of shall          the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, establish      Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception:
Insert 1
: a. NEI 94 1995, Section 9.3.2: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than April 9, 2007.
Type B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix J, Option A.
The peak calculated containment internal pressure for the design basis loss of The            coolant accident, P a , is 42 psig, which exceeds the peak calculated containment containment    internal pressure for the design basis loss of coolant accident.
design pressure is    The maximum allowable containment leakage rate, L a , at P a , shall be 0.1% of the 42 psig.       containment air weight per day.
Leakage rate acceptance criteria are:
: a. 1. Containment leakage rate acceptance criteria is < 1.0 L a . During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 L a for the Type B and Type C tests, and
                              < 0.75 L a for Type A tests.
Insert 2 The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
Insert 3 5.5.17        Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be implemented to ensure that, with an OPERABLE Control Room Emergency Filtration System, CRE occupants can control the nuclear power unit safely following a radiological event, hazardous chemical release, or a smoke challenge. The program shall include the following elements:
Insert 1: NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.
Insert 2: 2. Air lock testing acceptance criteria are:
: a. Overall air lock leakage rate is < 0.05 La when tested at > Pa.
: b. For each door, leakage rate is < 0.01 La when pressurized to > 42 psig            (continued)
HBRSEP                                           5.0-22                       Amendment No. 219 Insert 3: Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 5 Pages (including this cover page)
REVISED TECHNICAL SPECIFICATION PAGES

Containment 3.6.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.6.1.1      Perform required visual examinations and leakage         In accordance with rate testing except for containment air lock testing, in the Containment accordance with the Containment Leakage Rate             Leakage Rate Testing Program.                                         Testing Program SR 3.6.1.2      Verify containment structural integrity in accordance    In accordance with the Containment Tendon Surveillance Program.        with the Containment Tendon Surveillance Program HBRSEP Unit No. 2                           3.6-2                             Amendment No.

Containment Air Lock 3.6.2 ACTIONS CONDITION            REQUIRED ACTION                    COMPLETION TIME B.  (continued)            B.3   ----------NOTE--------------
Air lock doors in high radiation areas may be verified locked closed by administrative means.
Verify an OPERABLE               Once per 31 days door is locked closed.
C. Containment air lock   C.1   Initiate action to               Immediately inoperable for reasons       evaluate overall other than Condition A       containment leakage or B.                        rate per LCO 3.6.1.
AND C.2   Verify a door is closed in       1 hour the air lock.
AND C.3   Restore air lock to              24 hours OPERABLE status.
D. Required Action and    D.1   Be in MODE 3.                    6 hours associated Completion Time not met.          AND D.2   Be in MODE 5.                    36 hours HBRSEP Unit No. 2                 3.6-5                                  Amendment No.

Containment Air Lock 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                   FREQUENCY SR 3.6.2.1      ---------------------------NOTES---------------------------------
: 1.      An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
: 2.      Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.
Perform required air lock leakage rate testing in                       In accordance with accordance with the Containment Leakage Rate                            the Containment Testing Program.                                                        Leakage Rate Testing Program.
SR 3.6.2.2      Verify only one door in the air lock can be                             24 months opened at a time.
HBRSEP Unit No. 2                                  3.6-6                                     Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16       Containment Leakage Rate Testing Program A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42 psig. The containment design pressure is 42 psig.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of the containment air weight per day.
Leakage rate acceptance criteria are:
: 1. Containment leakage rate acceptance criteria is < 1.0 La. During the first unit 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 Type A tests.
: 2. Air lock testing acceptance criteria are:
: a. Overall air lock leakage rate is  0.05 La when tested at  Pa.
: b. For each door, leakage rate is  0.05 La when pressurized to  42 psig.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.
5.5.17       Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be implemented to ensure that, with an OPERABLE Control Room Emergency Filtration System, CRE occupants can control the nuclear power unit safely following a radiological event, hazardous chemical release, or a smoke challenge. The program shall include the following elements:
(continued)
HBRSEP                                         5.0-22                       Amendment No.

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 11 Pages (including this cover page)
PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES

SR Applicability B 3.0 BASES SR 3.0.1          Surveillances, including Surveillances invoked by Required (continued)
Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.
Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.
SR 3.0.2          SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that equires the periodic performance of the Required Action on a once per . . . interval.
SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).
The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.
These exceptions are stated in the individual Specifications. An example of where SR 3.0.2 does not apply is a Surveillance with a Frequency of in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.
(continued)
HBRSEP Unit No. 2                         B 3.0-14                             Revision No. 52

An example of where SR 3.0.2 does not apply is in the Containment Leakage Rate Testing Program.                  SR Applicability This program establishes testing requirements and                        B 3.0 Frequencies in accordance with the requirements of BASES             regulations.
SR 3.0.2           The requirements of regulations take precedence over the TS.
(continued)       The TS cannot in and of themselves extend a test interval specified in the regulations.
Therefore, there is a Note in the Frequency stating, "SR 3.0.2 is not applicable."
As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ..." basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.
The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.
SR 3.0.3           SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.
This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.
The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, (continued)
HBRSEP Unit No. 2                         B 3.0-15                               Revision No. 52

Containment B 3.6.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment BASES BACKGROUND          The containment consists of the concrete reactor building, its steel liner, and the penetrations through this structure. The structure is designed to contain radioactive material that may be released from the reactor core design basis        following a Design Basis Accident (DBA). Additionally, this structure loss of             provides shielding from the fission products that may be present in the coolant             containment atmosphere following accident conditions.
accident (LOCA)              The containment is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow dome roof. The inside surface of the containment is lined with a stainless steel liner to ensure a high degree of leak tightness during operating and accident conditions.
The cylinder wall is prestressed with a post tensioning system in the vertical direction.
Design Basis Accident (DBA)
The concrete reactor building is required for structural integrity of the SR 3.6.1.1 leakage    containment under DBA conditions. The steel liner and its penetrations rate requirements     establish the leakage limiting boundary of the containment. Maintaining comply with 10 CFR    the containment OPERABLE limits the leakage of fission product 50, Appendix J,       radioactivity from the containment to the environment.
Option B (Ref. 1), as modified by approved  The isolation devices for the penetrations in the containment boundary exemptions.           are a part of the containment leak tight barrier. To maintain this leak tight barrier:
: a.       All penetrations required to be closed during accident conditions are either:
: 1.      capable of being closed by an OPERABLE automatic containment isolation system, or
: 2.      closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.3, "Containment Isolation Valves";
(continued)
HBRSEP Unit No. 2                            B 3.6-1                              Revision No. 0

Containment B 3.6.1 BASES BACKGROUND        b.       The air lock is OPERABLE, except as provided in (continued)               LCO 3.6.2, "Containment Air Lock";
: c.       The equipment hatch is closed and sealed; and
: d.       The Isolation Valve Seal Water (IVSW) sytem is OPERABLE, except as provided in LCO 3.6.8.
APPLICABLE        The safety design basis for the containment is that the SAFETY ANALYSES containment must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.
The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA) and a steam line break (Ref. 2). In addition, release of significant fission product radioactivity within containment can occur from a LOCA. In the LOCA analyses, it is assumed that the containment is OPERABLE such that, for the LOCA, the release to the environment is controlled by the rate of containment leakage. The containment has an allowable leakage rate of 0.1% of containment air weight per day (Ref. 2). This leakage rate, used to evaluate offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Ref. 1), as L a : the maximum allowable
   , Option B containment leakage rate at the calculated peak containment internal pressure (P a ) resulting from the LOCA. At HBRSEP, Unit 2, P a is design basis specified as the containment design pressure of 42 psig, which exceeds the calculated peak LOCA containment pressure. The allowable leakage rate represented by L a forms the basis for the acceptance criteria imposed on all containment leakage rate testing. L a is assumed to be 0.1% per day in the safety analysis at P a = 42 psig (Ref. 2).
Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY.
The containment satisfies Criterion 3 of the NRC Policy Statement.
(continued)
HBRSEP Unit No. 2                          B 3.6-2                            Revision No. 34

Containment B 3.6.1 BASES (continued)
LCO               Containment OPERABILITY is maintained by limiting leakage to  1.0 L a ,
except prior to the first startup after performing a required 10 CFR 50, Containment      Appendix J, leakage test. At this time, the combined Type B and C Leakage Rate     leakage must be < 0.6 L a and the overall Type A leakage must be Testing          < 0.75 L a . applicable leakage limits must be met.
Program Compliance with this LCO will ensure a containment configuration, including the equipment hatch, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.
Individual leakage rates specified for the containment air lock are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J.
Therefore, leakage rates exceeding these individual limits only result in the containment being inoperable when the leakage results in exceeding overall          the acceptance criteria of Appendix J. 1.0 La APPLICABILITY     In MODES 1, 2, 3, and 4, a LOCA could cause a release of radioactive material into containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, containment is not required to be OPERABLE in MODE 5 to prevent leakage of radioactive material from containment. The requirements for containment during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations."
ACTIONS           A.1 In the event containment is inoperable, containment must be restored to OPERABLE status within 1 hour. The 1 hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining containment OPERABLE during MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring during periods when containment is inoperable is minimal.
(continued)
HBRSEP Unit No. 2                          B 3.6-3                               Revision No. 0

Containment B 3.6.1 BASES ACTIONS               B.1 and B.2 (continued)
If containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE          SR 3.6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 1), Option Insert 1              A, as modified by approved exemptions. Air lock leakage is not acceptable if its contribution to overall Type B, and C leakage causes overall Type B and C leakage to exceed limits. As left leakage prior to the first startup after performing a required 10 CFR 50, Appendix J,         Insert 2 leakage test is required to be < 0.6 L a for combined Type B and C Insert 3              leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall leakage limit of  1.0 L a . At 1.0 L a the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by Appendix J, Insert 1              Option A. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.
Reviewer's Note: NEI 94-01 includes acceptance criteria for as-left and as-found Type A leakage rates and combined Type B and C leakage SR 3.6.1.2 rates, which may be reflected in the Bases.
This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program.
Insert 1: the Containment Leakage Rate testing Program.
Insert 2: Containment Leakage Rate testing Program Insert 3: , < 0.60 La for the Type B and Type C tests, and < 0.75 La for Type A tests.
(continued)
HBRSEP Unit No. 2                              B 3.6-4                             Revision No. 0

Containment B 3.6.1 BASES SURVEILLANCE      SR 3.6.1.3 REQUIREMENTS (continued)      Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program applicable to Type A leakage rate tests. Air lock leakage is not acceptable if its contribution to overall Type A leakage causes overall Type A leakage to exceed limits.
As left leakage after performing a required 10 CFR 50, Appendix J, leakage test is required to be < 0.75 L a for overall Type A leakage. At 1.0 L a the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by the Containment Leakage Rate Testing Program. This periodic testing requirement verifies that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.
REFERENCES        1.       10 CFR 50, Appendix J. , Option B
: 2.       UFSAR, Section 6.2.
HBRSEP Unit No. 2                         B 3.6-5                               Revision No. 0

Containment Air Lock B 3.6.2 BASES SAFETY ANALYSES leakage. The containment has an allowable leakage rate of 0.1% of (continued)      containment air weight per day at 42 psig (Ref. 2).
The containment air lock satisfies Criterion 3 of the NRC Policy Statement.
LCO               The containment air lock forms part of the containment pressure boundary. As part of containment, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.
The air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE. Closure of a single door in the air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.
or APPLICABILITY     In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment air locks are not required in MODE 5 to prevent leakage of radioactive material from containment. The requirements for the containment air locks during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations."
(continued)
HBRSEP Unit No. 2                          B 3.6-7                                 Revision No.34

Containment Air Lock B 3.6.2 BASES ACTIONS           C.1, C.2, and C.3 (continued) inoperable air lock to OPERABLE status, assuming that at least one door is maintained closed in the air lock.
D.1 and D.2 If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE      SR 3.6.2.1 REQUIREMENTS Maintaining the containment air lock OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 1), Option A, as modified by approved exemptions. This SR reflects the the leakage rate testing requirements with regard to air lock leakage Containment       (Type B leakage tests). The periodic testing requirements verify that the Leakage Rate      air lock leakage does not exceed the allowed fraction of the overall Testing           containment leakage rate. The Frequency is required by Appendix J Program           (Ref. 1), Option A, as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply.
The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria of SR 3.6.1.1. This ensures that air lock leakage is properly accounted for which is         in determining the overall containment leakage rate.
applicable to combined Type B and C (continued)
HBRSEP Unit No. 2                        B 3.6-11                              Revision No. 0

Containment Air Lock B 3.6.2 BASES SURVEILLANCE      SR 3.6.2.2 REQUIREMENTS (continued)      The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur. Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is not normally challenged when the containment air lock door is used for entry and exit (procedures require strict adherence to single door opening), this test is only required to be performed every 24 months. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, and the potential for loss of containment OPERABILITY if the surveillance were performed with the reactor at power. The 24 month Frequency for the interlock is justified based on generic operating experience. The 24 month Frequency is based on engineering judgment and is considered adequate given that the interlock is not challenged during the use of the interlock.
REFERENCES        1.      10 CFR 50, Appendix J. , Option B
: 2.      UFSAR, Paragraph 6.9.2.
HBRSEP Unit No. 2                          B 3.6-12                             Revision No. 0

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 80 Pages (including this cover page)
EVALUATION OF RISK SIGNIFICANCE OF PERMANENT ILRT EXTENSION

Robinson Nuclear Plant:
Evaluation of Risk Significance of Permanent ILRT Extension 54002-CALC-01 Prepared for:
Robinson Nuclear Plant Project

==Title:==
Permanent ILRT Extension Revision: 0A Name and Date Preparer: Justin Sattler Reviewer: Matthew Johnson Review Method                  Design Review  Alternate Calculation Approved by: Nicholas Lovelace Revision 0A                                                                Page 1 of 79

54002-CALC-01       Evaluation of Risk Significance of Permanent ILRT Extension REVISION RECORD 

==SUMMARY==

Revision                    Revision Summary 0A                          Initial Issue Revision 0A                                                         Page 2 of 79

54002-CALC-01                                   Evaluation of Risk Significance of Permanent ILRT Extension 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 .................................................................................................................20 5.2.4 Step 4 - Determine the Change in Risk in Terms of LERF...................................22 5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability
               ..........................................................................................................................25 5.2.6 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage
               ..........................................................................................................................25 5.3 Sensitivities .................................................................................................................28 5.3.1 Potential Impact from External Events Contribution .............................................28 5.3.2 Potential Impact from Steel Liner Corrosion Likelihood ........................................30 5.3.3 Expert Elicitation Sensitivity .................................................................................32 5.3.4 Large Leak Probability Sensitivity Study ..............................................................33 6.0    RESULTS ......................................................................................................................34

==7.0    CONCLUSION==
S AND RECOMMENDATIONS ..............................................................35 A. Attachment 1 .................................................................................................................36 Revision 0A                                                                                                               Page 3 of 79

54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension 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 Robinson Nuclear Plant (RNP). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1], the methodology used in EPRI TR-104285 [Reference 2], the NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals from November 2001 [Reference 3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 as applied to ILRT interval extensions, and risk insights in support of a request for a 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 1009325, Revision 2-A of EPRI 1018243 [Reference 24].
2.0      SCOPE Revisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing frequency requirement from three in ten years to at least once in ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage rate was less than limiting containment leakage rate of 1La.
The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, September 1995 [Reference 6], provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessment of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the 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 RNP.
NEI 94-01 Revision 2-A contains a Safety Evaluation Report that supports using EPRI Report No. 1009325 Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, for performing risk impact assessments in support of ILRT extensions [Reference 24].
The Guidance provided in Appendix H of EPRI Report No. 1009325 Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.
It should be noted that containment leak-tight integrity is also verified through periodic in-service Revision 0A                                                                               Page 4 of 79

54002-CALC-01                          Evaluation of Risk Significance of Permanent ILRT Extension inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.
The acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 10-6 per reactor year. RG 1.174 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the Conditional Containment Failure Probability (CCFP), which helps ensure the defense-in-depth philosophy is maintained, is also calculated.
Regarding CCFP, changes of up to 1.1% have been accepted by the NRC for the one-time requests for extension of ILRT intervals. In context, it is noted that a CCFP of 1/10 (10%) has been approved for application to evolutionary light water designs. Given these perspectives, a change in the CCFP of up to 1.5% is assumed to be small.
In 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.
For those plants that credit containment overpressure for the mitigation of design basis accidents, a brief description of whether overpressure is required should be included in this section. In addition, if overpressure is included in the assessment, other risk metrics such as CDF should be described and reported.
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54002-CALC-01                         Evaluation of Risk Significance of Permanent ILRT Extension

==3.0     REFERENCES==

The following references were used in this calculation:
: 1. Revision 3-A to Industry Guideline for Implementing Performance-Based Option of 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: Surry Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 3, Revision 1, Part 1, October 1990.
: 8. Letter from R. J. Barrett (Entergy) to U. S. Nuclear Regulatory Commission, IPN-01-007, January 18, 2001.
: 9. United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB0178), April 17, 2001.
: 10. Impact of 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. Calculation No. RNP-F/PSA-0077, Revision 2, Robinson Nuclear Plant, RNP QUANTIFICATION CALCULATION, September 2012.
: 18. Calculation No. RNP-F/PSA-0048, Revision 2, H.B. Robinson Steam Electric Plant, PSA Model Section 9.0 - Source Term Development.
: 19. Robinson Nuclear Plant License Renewal Application, 2004.
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54002-CALC-01                         Evaluation of Risk Significance of Permanent ILRT Extension
: 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. Procedure EST-085, Revision 17, H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2, CONTAINMENT INTEGRATED LEAK RATE TEST.
: 28. Letter L-14 -121, ML14111A291, FENOC Evaluation of the Proposed Amendment, Beaver Valley Power Station, Unit Nos. 1 and 2, April 2014.
: 29. Technical Letter Report ML112070867, Containment Liner Corrosion Operating Experience Summary, Revision 1, August 2011.
: 30. Calculation No. RNP-RA/02-0028, Revision 0, H.B. Robinson Steam Electric Plant, One-Time Extension of Containment Type A Test Interval, March 2002.
: 31. Armstrong, J., Simplified Level 2 Modeling Guidelines: WOG PROJECT: PA-RMSC-0088, Westinghouse, WCAP-16341-P, November 2005.
: 32. Calculation No. RNP-F/PSA-0046, Revision 1, H.B. Robinson Steam Electric Plant, PSA Model Section 7.0 - Containment Safeguards and Plant Damage States.
: 33. Evacuation Time Estimates for the RNP Steam Electric Plant, Plume Exposure Pathway, Emergency Planning Zone, HMM Associates, June 1987.
: 34. Harrison, D., Generic Component Fragilities for the GE Advanced BWR Seismic Analysis, International Technology Corporation, September 1988.
: 35. IPEEE, H. B. Robinson Steam Electric Plant, Unit No. 2, Individual Plant Examination Submittal, August 1992.
: 36. Calculation No. RNP-F/PSA-0047, Revision 2, H.B. Robinson Steam Electric Plant, PSA Model Section 8.0 - Containment Response Assessment.
: 37. Calculation No. 4150-Seismic-005, Revision 0, AMEC Foster Wheeler Environment &
Infrastructure, Inc., Soil Hazard and Smoothed GMRS at Elevation 226 ft for the Robinson Plant Area.
: 38. Letter from W. R. Gideon (Duke Energy, H. B. Robinson) to U. S. Nuclear Regulatory Commission, Docket No. 50-261, ML13267A211, September 16, 2013.
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54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension
: 39. Calculation No. RNP-F/PSA-0054, Revision 1, H.B. Robinson Steam Electric Plant, Unit No. 2, PSA Appendix J - Containment Overpressure Capacity.
: 40. Calculation No. RNP-F/PSA-0050, Revision 0, H.B. Robinson Steam Electric Plant, Unit No. 2, Estimate of 50 Mile Population Dose from Design Basis Containment Leakage Following a Core Melt Accident.
: 41. 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.
: 42. The Nuclear Energy Institute - Seismic Risk Evaluations for Plants in the Central and Eastern United States, ML14083A596, March 2014.
4.0     ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculation:
The technical adequacy of the RNP PRA is consistent with the requirements of Regulatory Guide 1.200 as is relevant to this ILRT interval extension, as detailed in Attachment 1.
The RNP Level 1 and Level 2 internal events PRA models provide representative results.
It is appropriate to use the RNP 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 analysis and the detailed Fire PRA (model RNP_12Fh_s1) are used for this sensitivity analysis. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose) will not substantially differ if detailed analysis of seismic and 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.
The representative containment leakage for Class 3a sequences is 10La based on the previously approved methodology performed for Indian Point Unit 3 [Reference 8, Reference 9].
The representative containment leakage for Class 3b sequences is 100La based on the guidance provided in EPRI Report No. 1009325, Revision 2-A (EPRI 1018243)
[Reference 24].
The Class 3b can be very conservatively categorized as LERF based on the previously approved methodology [Reference 8, Reference 9].
The impact on population doses from containment bypass scenarios is not altered by the proposed ILRT extension, but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes on the conclusions from this analysis will result from this separate categorization.
The reduction in ILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal.
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54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension 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 LLRT test intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the basis for the consequence analysis of the ILRT interval extension for RNP. 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 Revision 0A                                                                              Page 9 of 79

54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension 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:
Sthe 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.
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 Revision 0A                                                                             Page 10 of 79

54002-CALC-01                         Evaluation of Risk Significance of Permanent ILRT Extension
: 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:
Sthe 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 yearS 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 Surry. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the RNP Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent RNP. (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 methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms was factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete base-mat, each with a steel liner.
EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals [Reference 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 RNP 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.
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54002-CALC-01                              Evaluation of Risk Significance of Permanent ILRT Extension 5.1.2   Plant Specific Inputs The plant-specific information used to perform the RNP ILRT Extension Risk Assessment includes the following:
Level 1 Model results [Reference 17]
Level 2 Model results [Reference 17, Reference 18, Reference 19]
Release category definitions used in the Level 2 Model [Reference 18, Reference 19]
Dose within a 50-mile radius [Reference 19]
ILRT results to demonstrate adequacy of the administrative and hardware issues
[Reference 30]
Containment failure probability data [Reference 18, Reference 32, Reference 33]
RNP Model The Internal Events PRA Model that is used for RNP is characteristic of the as-built plant. The current Level 1 model (RNP PRA Model Version RNP_12) [Reference 17] is a linked fault tree model. The total CDF is 1.05E-5/year [Table 18 of Reference 17]. When the #PDS_TOP gate is quantified to use flag events to classify the cutsets into plant damage states, the calculated CDF is 1.72E-5/year [Table 14 of Reference 17]. Since this calculation uses the results from the PDS quantification that contains double counting of scenarios risk by classifying the same cutset in multiple PDSs and the CAFTA software uses the min-cut upper bound quantification approach, results of this analysis are conservative. Table 5-1 and Table 5-2 provides a summary of the Internal Events CDF and LERF results for RNP PRA Model Version RNP_12. The total LERF when quantified with PDSs is 1.67E-6/year.
The total Fire CDF is 1.80E-5/year, and the total Fire LERF is 3.76E-6/year (RNP Fire PRA Model Version RNP_12Fh_s1). The Seismic PRA results from the IPEEE Seismic Margins Analysis do not result in estimate of CDF [Reference 35]. Refer to Section 5.3.1 for further details on external events as they pertain to this analysis.
Table 5 Internal Events CDF (RNP PRA Model Version RNP_12)
Internal Events                  Frequency (per year)
ISLOCA                                          2.83E-08 LOCA                                            1.64E-06 SGTR                                            1.12E-07 Internal Flood                                  1.92E-06 TransientLOCA                                  3.38E-06 TransientLoss of DHR                           9.10E-06 ATWS                                            1.01E-06 Total Internal Events CDF                       1.72E-05 Total Internal Events CDF 1.71E-05 (Excluding ISLOCA & SGTR)
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54002-CALC-01                               Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Internal Events LERF (RNP PRA Model Version RNP_12)
Internal Events                        Frequency (per year)
ISLOCA                                                  1.60E-09 LOCA                                                    3.71E-08 SGTR                                                    1.05E-07 Internal Flood                                          3.55E-07 TransientLOCA                                          3.53E-07 TransientLoss of DHR                                   7.59E-07 ATWS                                                    5.55E-08 Total Internal Events LERF                              1.67E-06 Population Dose Calculations The population dose calculation was reported in the One-Time Extension of Containment Type A Test Interval and Estimate of 50 Mile Population Dose from Design Basis Containment Leakage Following a Core Melt Accident [References 30 and 40]. Table 5-3 presents dose exposures calculated from methodology described in Reference 1 and data from Reference 30.
Reference 30 provides the population dose (person-rem) for Classes 1, 2, 6, 7, and 8; Class 3a and 3b population dose values are calculated from the Class 1 population dose and represented as 10La and 100L a, respectively, as guidance in Reference 1 dictates.
Table 5 Population Dose Accident Class                             Description                               Release (person-rem) 1                                 Containment Remains Intact                               1.56E+03 2                                Containment Isolation Failures                            2.22E+06 3a                     Independent or Random Isolation Failures SMALL                     1.56E+041 3b                     Independent or Random Isolation Failures LARGE                     1.56E+052 Isolation Failure in which pre-existing leakage is not                  n/a 4
dependent on sequence progression. Type B test Failures Isolation Failure in which pre-existing leakage is not                  n/a 5
dependent on sequence progression. Type C test Failures 6                Isolation Failure that can be verified by IST/IS or surveillance          8.85E+06 7                       Containment Failure induced by severe accident                     2.62E+06 8                         Accidents in which containment is by-passed                      2.45E+07
: 1. 10
* La
: 2. 100
* La Release Category Definitions Table 5-4 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.
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54002-CALC-01                                   Evaluation of Risk Significance of Permanent ILRT Extension Table 5 EPRI Containment Failure Classification [Reference 2]
Class                                                      Description Containment remains intact including accident sequences that do not lead to containment failure in the 1       long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant.
Containment isolation failures (as reported in the Individual Plant Examinations) including those accidents 2
in which there is a failure to isolate the containment.
Independent (or random) isolation failures include those accidents in which the pre-existing isolation 3
failure to seal (i.e., provide a leak-tight containment) is not dependent on the sequence in progress.
Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 3 isolation 4
failures, but is applicable to sequences involving Type B tests and their potential failures. These are the Type B-tested components that have isolated, but exhibit excessive leakage.
Independent (or random) isolation failures including those accidents in which the pre-existing isolation 5       failure to seal is not dependent on the sequence in progress. This class is similar to Class 4 isolation failures, but is applicable to sequences involving Type C test and their potential failures.
Containment isolation failures including those leak paths covered in the plant test and maintenance 6
requirements or verified per in-service inspection and testing (ISI/IST) program.
Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J 7
testing requirements do not impact these accidents.
Accidents in which the containment is bypassed (either as an initial condition or induced by phenomena) 8 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 5-4, 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 delta LERF values for several plants may fall above the very small change guidelines of the NRC Regulatory Guide 1.174
[Reference 4]. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for 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 Revision 0A                                                                                              Page 14 of 79

54002-CALC-01                          Evaluation of Risk Significance of Permanent ILRT Extension 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 RNP, as detailed in Section 5.2, involves the following:
The Class 2 and Class 8 sequences and some Class 6 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. The subtracted Class 6 sequences involve large early containment failure. These sequences are already considered to contribute to LERF in the RNP 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 5-5.
The analysis performed examined RNP-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 Revision 0A                                                                                Page 15 of 79

54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension evaluation as part of the baseline risk profile. However, they are not affected by the ILRT frequency change.
Class 4 and 5 sequences are impacted by changes in Type B and C test intervals; therefore, changes in the Type A test interval do not impact these sequences.
Table 5 EPRI Accident Class Definitions Accident Classes                                             Description (Containment Release Type) 1                                              No Containment Failure 2                                     Large Isolation Failures (Failure to Close) 3a                                      Small Isolation Failures (Liner Breach) 3b                                      Large Isolation Failures (Liner Breach) 4                                Small Isolation Failures (Failure to Seal - Type B) 5                                Small Isolation Failures (Failure to Seal - Type C) 6                                Other Isolation Failures (e.g., Dependent Failures) 7                                 Failures Induced by Phenomena (Early and Late) 8                                        Bypass (Interfacing System LOCA)
CDF                           All CET End States (Including Very Low and No Release)
The steps taken to perform this risk assessment evaluation are as follows:
Step 1 - Quantify the baseline risk in terms of frequency per reactor year for each of the accident classes presented in Table 5-5.
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 5-5 were developed for RNP by first determining the frequencies for Classes 1, 2, 6, 7, and 8. Table 5-6 presents the grouping of each release category in EPRI Classes based on the associated description. Table 5-7 presents the release categories, frequency, and EPRI category for each sequence and the totals of each EPRI classification. Table 5-8 provides a summary of the accident sequence Revision 0A                                                                                         Page 16 of 79

54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension frequencies that can lead to radionuclide release to the public and have been derived consistent with the definitions of accident classes defined in EPRI TR-104285 [Reference 2], the NEI Interim Guidance [Reference 3], and guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24]. Adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion of the steel liner per the methodology described in Section 5.2.6. 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
         =         = 0.0092 217 Multiplying the CDF by the probability of a Class 3a leak yields the Class 3a frequency contribution in accordance with guidance provided in Reference 24. As described in Section 5.1.3, additional consideration is made to not apply failure probabilities on those cases that are already LERF scenarios (i.e., release categories RC-2, RC-2B, RC-4, RC-4C, RC-5, and RC-5C
[Reference 18]; values for these release categories are shown in Table 5-6). Therefore, these LERF contributions from CDF are removed. Therefore, the frequency of a Class 3a failure is calculated by the following equation:
               =                    2     2        4    4       5    5
         =        1.72E 5.80E 3.82E 0 - 1.15E 0 - 5.88E-7 = 1.43E-7 In the database of 217 ILRTs, there are zero containment leakage events that could result in a large early release. Therefore, the Jeffreys non-informative prior is used to estimate a failure rate and is illustrated in the following equations:
5678 9: ;<=6 > + 1/2 Jeffreys Failure Probability =
5678 9: A >B> + 1 0 + 1/2 C =             = 0.0023 217 + 1 The frequency of a Class 3b failure is calculated by the following equation:
C =         C        2 2  4 4  5 5
             .D
         =        1.72E 5.80E 3.82E 0 - 1.15E 0 - 5.88E-7 = 3.56E-8 E
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 Revision 0A                                                                             Page 17 of 79

54002-CALC-01                                Evaluation of Risk Significance of Permanent ILRT Extension per year is initially determined from the EPRI Accident Class 1 frequency listed in Table 5-7 and then subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF), calculated below:
            =                                              C Class 2 Sequences. This group consists of core damage accident progression bins with large containment isolation failures. The frequency per year for these sequences is obtained from the EPRI Accident Class 2 frequency listed in Table 5-7.
Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis, consistent with approved methodology.
Class 5 Sequences. This group consists of all core damage accident progression bins for which a containment isolation failure-to-seal of Type C test components. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis, consistent with approved methodology.
Class 6 Sequences. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs.
These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution. For RNP, this class is defined as the release categories RC-2, RC-2B, and RC-3. The frequency per year for these sequences is obtained from the EPRI Accident Class 6 frequency listed in Table 5-7.
Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (e.g., overpressure). For this analysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table 5-7.
Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass or SGTR occurs. For this analysis, the frequency is determined from the EPRI Accident Class 8 frequency listed in Table 5-7.
LERF quantification is distributed into EPRI categories based on release categories. Table 5-6 shows this distribution.
Table 5 Release Category Frequencies Release                         Description of Outcome                       EPRI   Frequency (/yr)
Category                                                                   Category IC-1                              Containment intact                         1         1.41E-05 RC-1       Late containment failure caused by gradual overpressurization     7         0.00E+00 Late containment failure caused by gradual overpressurization RC-1A                                                                         7         0.00E+00 with revaporization Late containment failure caused by gradual overpressurization RC-1B                                                                         7         1.16E-06 without scrubbing Late containment failure caused by gradual overpressurization RC-1BA                                                                        7         4.66E-08 with revaporization, without scrubbing RC-2                       Large early containment failure                   6         5.80E-07 RC-2B              Large early containment failure without scrubbing          6         3.82E-07 Early containment isolation failure RC-3                                                                         6         1.84E-07 with a small leakage rate Early containment isolation failure RC-3B                                                                         7         2.31E-08 with a small leakage rate without scrubbing RC-4               Containment bypass with a small leakage rate              2         0.00E+00 RC-4C      Containment bypass with a small leakage rate without scrubbing     8         1.15E-07 Revision 0A                                                                                  Page 18 of 79

54002-CALC-01                                  Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Release Category Frequencies Release                           Description of Outcome                            EPRI      Frequency (/yr)
Category                                                                          Category Containment bypass with a large RC-5                                                                                8           0.00E+00 leakage rate Containment bypass with a large RC-5C                                                                                8           5.88E-07 leakage rate without scrubbing Contribution to EPRI Classification 6                                  1.15E-06 Contribution to EPRI Classification 7                                  1.23E-06 Contribution to EPRI Classification 8                                  7.03E-07 Table 5 Accident Class Frequencies Release Categories                      EPRI Category              Frequency (/yr)
IC-1                                Class 1                  1.41E-05 RC-4                                Class 2                  0.00E+00 RC-2, RC-2B, RC-3                           Class 6                  1.15E-06 RC-1, RC-1A, RC-1B, RC-1BA, RC-3B                    Class 7                  1.23E-06 RC-4C, RC-5, RC-5C                           Class 8                  7.03E-07 Total (CDF)                               N/A                    1.72E-05 Table 5 Baseline Risk Profile Class                                    Description                                  Frequency (/yr) 1                               No containment failure                                1.39E-052 2                      Large containment isolation failures                           0.00E+00 3a                      Small isolation failures (liner breach)                        1.43E-07 3b                      Large isolation failures (liner breach)                        3.56E-08 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)          1.15E-06 7          Severe accident phenomena induced failure (early and late)                 1.23E-06 8                                Containment bypass                                   7.03E-07 Total                                       1.72E-05
: 1. represents a probabilistically insignificant value.
: 2. The Class 3a and 3b frequencies are subtracted from Class 1 to preserve total CDF.
5.2.2    Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose)
Plant-specific release analyses were performed to estimate the person-rem doses to the population within a 50-mile radius from the plant [References 30 and 40]. The releases are based on RNP-specific dose calculations summarized on Table 5-3. Table 5-9 provides a correlation of RNP population dose to EPRI Accident Class. Table 5-10 provides population dose for each EPRI accident class.
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54002-CALC-01                                Evaluation of Risk Significance of Permanent ILRT Extension 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:
G     H =;>> 3; 9I6=;B<9J 9> = 10  1.56G+3 = 1.56G+4 G     H =;>> 38 9I6=;B<9J 9> = 100  1.56G+3 = 1.56G+5 Table 5 Mapping of Population Dose to EPRI Accident Class Release Category           EPRI Category            Frequency (/yr) Dose (person-rem)
IC-1                     Class 1                 1.41E-05         1.56E+03 RC-4                     Class 2                 0.00E+00         2.22E+06 RC-2, RC-2B, RC-3               Class 6                 1.15E-06         8.85E+06 RC-1, RC-1A, RC-Class 7                 1.23E-06         2.62E+06 1B, RC-1BA, RC-3B RC-4C, RC-5, RC-Class 8                 7.03E-07         2.45E+07 5C Table 5 Baseline Population Doses Class                                Description                                 Population Dose (person-rem) 1                             No containment failure                                     1.56E+03 2                     Large containment isolation failures                               2.22E+06 3a                     Small isolation failures (liner breach)                           1.56E+041 3b                     Large isolation failures (liner breach)                           1.56E+052 4                Small isolation failures - failure to seal (type B)                        N/A 5                Small isolation failures - failure to seal (type C)                        N/A 6      Containment isolation failures (dependent failure, personnel errors)              8.85E+06 7          Severe accident phenomena induced failure (early and late)                    2.62E+06 8                              Containment bypass                                        2.45E+07
: 1. 10*La
: 2. 100*La 5.2.3    Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Years The next step is to evaluate the risk impact of extending the test interval from its current 10-year interval to a 15-year interval. To do this, an evaluation must first be made of the risk associated with the 10-year interval, since the base case applies to 3-year interval (i.e., a simplified representation of a 3-to-10 interval).
Risk Impact Due to 10-Year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and Class 3b sequences is impacted. The risk contribution is changed based on the NEI guidance as described in Section 5.1.3 by a factor of 10/3 compared to the base case values. The Class 3a and 3b frequencies are calculated as follows:
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54002-CALC-01                                   Evaluation of Risk Significance of Permanent ILRT Extension 10 2 K        LMN   =                         2         2     4      4       5     5 3 217 L
        =              1.55E-5 = 4.77E-7 10 . 5 K     C LMN    =                         2         2     4      4       5     5 3 218 L     .D
        =              1.55E-5 = 1.19E-7 E
The results of the calculation for a 10-year interval are presented in Table 5-11.
Table 5 Risk Profile for Once in 10 Year ILRT Class                 Description                 Frequency     Contribution     Population      Population
(/yr)         (%)        Dose (person-      Dose Rate rem)       (person-rem/yr) 1             No containment failure2            1.35E-05       78.62%          1.56E+03         2.11E-02 2     Large containment isolation failures       0.00E+00       0.00%           2.22E+06        0.00E+00 Small isolation failures (liner 3a                                                4.77E-07       2.77%           1.56E+04         7.44E-03 breach)
Large isolation failures (liner 3b                                                1.19E-07       0.69%           1.56E+05         1.85E-02 breach)
Small isolation failures - failure to 4                                                     1           1               1               1 seal (type B)
Small isolation failures - failure to 5                                                     1           1               1               1 seal (type C)
Containment isolation failures 6         (dependent failure, personnel          1.15E-06       6.66%           8.85E+06        1.01E+01 errors)
Severe accident phenomena 7                                                1.23E-06       7.16%           2.62E+06        3.22E+00 induced failure (early and late) 8              Containment bypass                7.03E-07       4.09%           2.45E+07        1.73E+01 Total                    1.72E-05                                       3.07E+01
: 1. represents a probabilistically insignificant value.
: 2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.
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:
15 2 K        DMN   =                         2         2     4      4       5     5 3 217
        =5            1.55E-5 = 7.16E-7 15 . 5 K     C DMN    =                         2         2     4      4       5     5 3 218
                  .D
        =5            1.55E-5 = 1.78E-7 E
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54002-CALC-01                                  Evaluation of Risk Significance of Permanent ILRT Extension The results of the calculation for a 15-year interval are presented in Table 5-12.
Table 5 Risk Profile for Once in 15 Year ILRT Class              Description               Frequency       Contribution   Population Dose       Population
(/yr)          (%)         (person-rem)          Dose Rate (person-rem/yr) 1          No containment failure2          1.32E-05         76.89%           1.56E+03            2.06E-02 Large containment isolation 2                                           0.00E+00         0.00%            2.22E+06           0.00E+00 failures Small isolation failures (liner 3a                                          7.16E-07         4.16%            1.56E+04            1.12E-02 breach)
Large isolation failures (liner 3b                                          1.78E-07         1.04%            1.56E+05            2.78E-02 breach)
Small isolation failures - failure 4                                                 1            1                1                  1 to seal (type B)
Small isolation failures - failure 5                                                 1            1                1                  1 to seal (type C)
Containment isolation failures 6       (dependent failure, personnel       1.15E-06         6.66%            8.85E+06           1.01E+01 errors)
Severe accident phenomena 7                                           1.23E-06         7.16%            2.62E+06           3.22E+00 induced failure (early and late) 8           Containment bypass              7.03E-07         4.09%            2.45E+07           1.73E+01 Total                 1.72E-05                                             3.07E+01
: 1. represents a probabilistically insignificant value.
: 2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.
5.2.4    Step 4 - Determine the Change in Risk in Terms of LERF The risk increase associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from an intact containment could, in fact, result in a larger release due to the increase in probability of failure to detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.
Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 [Reference 4] defines very small changes in risk as resulting in increases of CDF less than 10-6/year and increases in LERF less than 10-7/year, and small changes in LERF as less than 10-6/year. Since containment overpressure is not required in support of ECCS performance to mitigate design basis accidents at RNP, the ILRT extension does not impact CDF. Therefore, the relevant risk-impact metric is LERF.
For RNP, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a 10-year test interval from Table 5-11, the Class 3b frequency is 1.19E-7/year; based on a 15-year test interval from Table 5-12, the Class 3b frequency is 1.78E-7/year. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is 1.42E-7/year. Similarly, the increase due to increasing the interval from 10 to 15 years is 5.94E-8/year. As can be seen, even with the conservatisms included in the evaluation (per Revision 0A                                                                                         Page 22 of 79

54002-CALC-01                              Evaluation of Risk Significance of Permanent ILRT Extension the EPRI methodology), the estimated change in LERF is below the threshold criteria for a small change when comparing the 15-year results to the current 10-year requirement, and slightly greater than the criteria when compared to the original 3-year requirement. Table 5-13 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)              3.56E-08            1.19E-07         1.78E-07 LERF (3 year baseline)                                 8.31E-08         1.42E-07 LERF (10 year baseline)                                                   5.94E-08 The increase in the overall probability of LERF due to Class 3b sequences being slightly greater than 10-7 is not unexpected. Since the target is exceeded, some refinement is necessary. One method to remove some conservatism is to examine the source term expected to be available for release during the accident sequence. The source term is greatly reduced if the debris expelled from the reactor remains covered with water. Therefore, if the accident sequence contains containment spray success, the source term is not considered to lead to a large early release. Excluding scenarios where containment spray is credited and therefore scrubbing the source term release results in a frequency reduction. Section 7.5 in Reference 32 describes the RNP plant damage states and the containment systems that are failed and successful in each end state. Containment safeguards event tree end states A, B, C, G, H, and I encapsulate scenarios that have successful containment spray during both the injection and recirculation phases of the accident. Since release classes RC-2, RC-2B, RC-4, RC-4C, RC-5, and RC-5C have already been subtracted from CDF when calculating Class 3 frequencies, frequency was not redundantly removed in these release categories for containment spray success. To refine the Class 3 calculations by removing the frequency of scenarios with containment spray success, the frequency of the scenarios in release categories IC-1, RC-1, RC-1A, RC-1B, RC-1BA, RC-3, and RC-3B and end states A, B, C, G, H, and I are summed and removed from the Class 3 frequencies. Table 5-14 describes each end state.
Table 5 Summary of RNP Containment Safeguards Event Tree End States End State    Containment Isolation      Containment Sprays      Containment Fans A         Successful isolation          Functioning          Functioning B        Small isolation failure        Functioning          Functioning C        Large isolation failure        Functioning          Functioning D         Successful isolation      Fail in recirculation    Functioning E        Small isolation failure    Fail in recirculation    Functioning F        Large isolation failure    Fail in recirculation    Functioning G         Successful isolation          Functioning             Failed H        Small isolation failure        Functioning             Failed I        Large isolation failure        Functioning             Failed J         Successful isolation      Fail in recirculation       Failed K        Small isolation failure    Fail in recirculation       Failed L        Large isolation failure    Fail in recirculation       Failed Revision 0A                                                                                     Page 23 of 79

54002-CALC-01                                 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Summary of RNP Containment Safeguards Event Tree End States End State    Containment Isolation          Containment Sprays       Containment Fans M          Successful isolation            Fail in injection         Functioning N         Small isolation failure          Fail in injection         Functioning O         Large isolation failure          Fail in injection         Functioning P          Successful isolation            Fail in injection           Failed Q         Small isolation failure          Fail in injection           Failed R         Large isolation failure          Fail in injection           Failed Table 5-15 shows adjusted release category frequencies after some conservatisms from containment spray success excluded.
Table 5 Adjusted Class Frequencies Class                                 Description                                          Frequency (/yr) 1                            No containment failure1                                       1.40E-05 2                      Large containment isolation failures                                0.00E+00 3a                     Small isolation failures (liner breach)                              9.41E-08 3b                     Large isolation failures (liner breach)                              2.34E-08 4                Small isolation failures - failure to seal (type B)                       0.00E+00 5                Small isolation failures - failure to seal (type C)                       0.00E+00 Containment isolation failures (dependent failure, personnel 6                                                                                          1.15E-06 errors) 7          Severe accident phenomena induced failure (early and late)                      1.23E-06 8                              Containment bypass                                          7.03E-07 Total (CDF)                                                                                      1.72E-05
: 1. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.
Substituting these values into the previously defined equations and calculation method yields the final results displayed in Table 5-16.
Table 5 Impact on LERF due to Extended Type A Testing Intervals with Adjusted CDF ILRT Inspection Interval             3 Years (baseline)          10 Years         15 Years Class 3b (Type A LERF)                    2.34E-08              7.80E-08         1.17E-07 LERF (3 year baseline)                                         5.46E-08         9.36E-08 LERF (10 year baseline)                                                           3.91E-08 The adjusted containment spray inputs allow the values to be less than the 1E-7 LERF metric.
The delta LERF between the 3 years and the 15 years is 9.36E-8/yr. This value shows that the proposed extension meets the definition of a very small change in risk as defined in Regulatory Guide 1.174. Also, as explained in Section 5.2.1, since this calculation uses the results from the PDS quantification that contains double counting of risk, results are conservative.
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54002-CALC-01                             Evaluation of Risk Significance of Permanent ILRT Extension 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:
: J :
       =1 where f(ncf) is the frequency of those sequences that do not result in containment failure; this frequency is determined by summing the Class 1 and Class 3a results.
Since CCFP is only concerned with a containment failure and not whether the release is small or large, the Class 1 results without refinement must be used to calculate the CCFP. Table 5-17 shows the steps and results of this calculation. The difference in CCFP between the 3-year test interval and 15-year test interval is 0.829%.
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)             1.41E-05         1.40E-05          1.39E-05 f(ncf)/CDF                  0.819            0.814             0.810 CCFP                     0.181            0.186             0.190 CCFP (3 year baseline)                            0.483%            0.829%
CCFP (10 year baseline)                                              0.345%
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 0.829%. 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 base-mat, each with a steel liner.
The following approach is used to determine the change in likelihood, due to extending the ILRT, of detecting corrosion of the containment steel liner. This likelihood is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:
Differences between the containment base-mat and the containment cylinder and dome:
The historical steel liner flaw likelihood due to concealed corrosion The impact of aging The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw Revision 0A                                                                                 Page 25 of 79

54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension Assumptions Consistent with the Calvert Cliffs analysis, a half failure is assumed for base-mat concealed liner corrosion due to the lack of identified failures (See Table 5-18, 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 5-18, 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 5-18, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.
In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere, given that a liner flaw exists, was estimated as 1.1% for the cylinder and dome, and 0.11% (10% of the cylinder failure probability) for the base-mat. These values were determined from an assessment of the probability versus containment pressure. For RNP, the ILRT maximum pressure is 42 psig [References 27] and ultimate pressure of 130 psig [Reference 39]. Probabilities of 1% for the cylinder and dome, and 0.1% for the base-mat are used in this analysis, and sensitivity studies are included in Section 5.3.2 (See Table 5-18, Step 4).
Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the base-mat region is considered to be less likely than the containment cylinder and dome region (See Table 5-18, Step 4).
Although the Calvert Cliffs analysis uses a 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10%, the RNP containment is insulated such that complete visual inspection is not possible.
Approximately 74% of the RNP containment is covered by insulation and is not readily visible. The remaining 26% which is visible is examined as required by ASME Section XI, Subsection IWE. Therefore, a 77% visual inspection failure likelihood is used (See Table 5-18, 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.
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54002-CALC-01                                 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Steel Liner Corrosion Base Case Step                 Description                      Containment Cylinder and             Containment Basemat Dome (85%)                             (15%)
Historical liner flaw likelihood           Events: 2                             Events: 0 Failure data: containment location         (Brunswick 2 and North Anna 2)        Assume a half failure specific                                   2 / (70 x 5.5) = 5.19E-03             0.5 / (70 x 5.5) = 1.30E-03 1     Success data: based on 70 steel-lined containments and 5.5 years since the 10CFR 50.55a requirements of periodic visual inspections of containment surfaces Year                   Failure rate   Year               Failure rate Aged adjusted liner flaw likelihood During the 15-year interval, assume        1                      2.05E-03       1                  5.13E-04 failure rate doubles every five years      average 5-10           5.19E-03       average 5-10       1.30E-03 2                                                15                     1.43E-02       15                 3.57E-03 (14.9% increase per year). The average for the 5th to 10th year set to the historical failure rate.            15 year average = 6.44E-03            15 year average = 1.61E-03 Increase in flaw likelihood between 3 and 15 years Uses aged adjusted                   0.73% (1 to 3 years)               0.18% (1 to 3 years) 3     liner flaw likelihood (Step 2),                    4.18% (1 to 10 years)              1.04% (1 to 10 years) assuming failure rate doubles every                9.66% (1 to 15 years)              2.41% (1 to 15 years) five years.
Likelihood of breach in containment 4                                                                  1%                                0.1%
given liner flaw 77%
Approximately 74% of the containment cylinder and dome liner is covered by insulation and is not readily visible. It is assumed the remaining approximately 26%
Visual inspection detection failure        which is visible would be visually                 100%
5 likelihood                                 examined. A 5% probability is         Cannot be visually inspected then assumed that visible failures would not be detected, and a 5%
probability was assumed that flaws could exist but would not be visible.
0.74 + (0.26 * (0.05 + 0.05))
0.0056% (3 years)                     0.000180% (3 years) 0.73% x 1% x 77%                      0.18% x 0.1% x 100%
Likelihood of non-detected                 0.0322% (10 years)                    0.00104% (10 years) 6     containment leakage (Steps 3 x 4 x
: 5)                                         4.18% x 1% x 77%                      1.04% x 0.1% x 100%
0.0744% (15 years)                    0.00241% (15 years) 9.66% x 1% x 77%                      2.41% 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 base-mat, as summarized below for RNP.
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54002-CALC-01                              Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for RNP Description At 3 years: 0.00562% + 0.000180% = 0.00580%
At 10 years: 0.03279% + 0.00104% = 0.03323%
At 15 years: 0.07438% + 0.00241% = 0.07679%
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 a 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 NRG 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
[Reference 28]. 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.
Robinson is transitioning to NFPA 805 licensing basis for fire protection and submitted a License Amendment Request (LAR) on September 16, 2013 [Reference 38]. This transition included performing a Fire PRA and installing modifications to reduce the fire-induced CDF and LERF to those reported in the NFPA 805 LAR. It is anticipated that many, but not all, of the NFPA 805 modifications will be completed by the next scheduled ILRT. All the modifications will be installed prior to coming out of refueling outage 31 in 2018; the next scheduled ILRT is Revision 0A                                                                               Page 28 of 79

54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension February 2017. All of the modifications that have an appreciable impact on reducing overall fire CDF have been installed. The remaining modifications are of lower risk significance [Attachment S of Reference 38]. Compensatory measures have been implemented to reduce the fire risk until the modifications are implemented. These measures may be actions to reduce fire initiating event probabilities, actions to improve suppression probability, and/or actions to recover or protect systems that mitigate core damage and large early release accident sequences.
Therefore, the Fire PRA model is deemed applicable for this calculation.
The Fire PRA model RNP_12Fh_s1 was used to obtain the fire CDF and LERF values. To reduce conservatism in the model, the methodology of subtracting existing LERF from CDF is also applied to the Fire PRA model. The following shows the calculation for Class 3b:
0.5 C =         C          OG     =       1.80G-5  3.76G-6 = 3.26G-8 218 L                                 L     L.D C LMN =             C        OG      =              1.80G-5  3.76G-6 = 1.09E-7 E
D                                     L.D P      C DMN =              C       OG       =5            1.80G-5  3.76G-6 = 1.63E-7 E
The Seismic PRA results from the IPEEE Seismic Margins Analysis do not result in estimate of CDF [Reference 35]. The 2014 Seismic Reevaluations for operating reactor sites [Reference 42] state the conclusions reached in 2010 by GI-199 [Reference 41] remain valid for estimating Seismic CDF at plants in the Central and Eastern United States, which includes RNP. The most conservative Seismic CDF reported in Table D-1 of Reference 41 is 1.4E-5. Applying the internal event LERF/CDF ratio to the seismic CDF yields an estimated seismic LERF of 1.36E-6, as shown by the equation below.
LERFSeismic  CDFSeismic
* LERFIE / CDFIE = 1.4E-5
* 1.67E-6 / 1.72E-5 = 1.36E-6 Again subtracting LERF from CDF, the Class 3b frequency can be calculated by the following formulas:
L.D C =         C          OG     =   E 1.4E-5-1.36E-6 = 2.90E-8 L                               L    L.D C LMN =           C        OG      =           1.4E-5-1.36E-6 = 6.67E-8 E
D                               D    L.D C DMN =           C        OG      =           1.4E-5-1.36E-6 = 1.45E-7 E
The High Wind PRA results from the IPEEE [Section 5.3.6 of Reference 35] report a CDF of 2E-
: 6. Applying the internal event LERF/CDF ratio to the high wind CDF yields an estimated high wind LERF of 1.94E-7. Again subtracting LERF from CDF, the Class 3b frequency can be calculated by the following formulas:
L.D C =         C          OG     =      2.0E-6-1.94E-7 = 4.14E-9 E
L                               L    L.D C LMN =           C        OG      =           2.0E-6-1.94E-7 = 1.38E-8 E
D                               D    L.D C DMN =           C        OG      =           2.0E-6-1.94E-7 = 2.07E-8 E
The fire, seismic, and high wind contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The change in LERF is Revision 0A                                                                                 Page 29 of 79

54002-CALC-01                                 Evaluation of Risk Significance of Permanent ILRT Extension calculated for the 1 in 10 year and 1 in 15 year cases and the change defined for the external events in Table 5-20.
Table 5 RNP 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            6.58E-08              2.19E-07               3.29E-07               2.63E-07 Internal Events           2.34E-08              7.80E-08               1.17E-07               9.36E-08 Combined               8.92E-08              2.97E-07               4.46E-07               3.57E-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 3.57E-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 + LERFHW + LERFclass3Bincrease
         = 1.67E-6/yr + 3.76E-6/yr + 1.36E-6/yr + 1.94/yr + 3.40E-7/yr = 7.33E-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.
5.3.2      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 5-23. Since CCFP is only concerned with a containment failure and not whether the release is small or large, the Class 1 results without containment spray refinement is used to calculate the CCFP. The Annual Dose Rate calculations are performed using the containment spray adjustments. 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               3b              3b             LERF            LERF             LERF Frequency         Frequency       Frequency         Increase        Increase         Increase (3-per-10        (1-per-10       (1-per-15      (3-per-10 to    (3-per-10 to     (1-per-10 to year ILRT)       year ILRT)      year ILRT)        1-per-10)       1-per-15)        1-per-15)
Internal Event 3B           2.34E-08         7.81E-08        1.17E-07         5.47E-08        9.38E-08         3.91E-08 Contribution Corrosion Likelihood         2.52E-08         1.11E-07        2.33E-07         8.63E-08        2.08E-07         1.21E-07 X 1000 Revision 0A                                                                                          Page 30 of 79

54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Steel Liner Corrosion Sensitivity Case: 3B Contribution 3b               3b               3b            LERF           LERF          LERF Frequency        Frequency       Frequency         Increase       Increase      Increase (3-per-10        (1-per-10        (1-per-15     (3-per-10 to   (3-per-10 to  (1-per-10 to year ILRT)       year ILRT)       year ILRT)       1-per-10)      1-per-15)     1-per-15)
Corrosion Likelihood    4.09E-08         4.12E-07         1.28E-06        3.71E-07       1.24E-06      8.64E-07 X 10000 Corrosion Likelihood    1.99E-07         3.42E-06         1.17E-05        3.22E-06       1.15E-05      8.29E-06 X 100000 Table 5 Steel Liner Corrosion Sensitivity: CCFP CCFP           CCFP          CCFP CCFP             CCFP             CCFP          Increase       Increase      Increase (3-per-10        (1-per-10        (1-per-15     (3-per-10 to   (3-per-10 to  (1-per-10 to year ILRT)       year ILRT)       year ILRT)       1-per-10)      1-per-15)     1-per-15)
Baseline 1.81E-01         1.86E-01         1.90E-01        4.83E-03       8.29E-03      3.45E-03 CCFP Corrosion Likelihood    1.81E-01         1.87E-01         1.90E-01        5.19E-03       8.91E-03      3.71E-03 X 1000 Corrosion Likelihood    1.83E-01         1.91E-01         1.97E-01        8.45E-03       1.45E-02      6.03E-03 X 10000 Corrosion Likelihood    1.97E-01         2.38E-01         2.67E-01        4.10E-02       7.03E-02      2.93E-02 X 100000 Table 5-23 -Steel Liner Corrosion Sensitivity: Does Rate Dose Rate      Dose Rate     Dose Rate Dose Rate      Dose Rate        Dose Rate        Increase       Increase      Increase (3-per-10      (1-per-10        (1-per-15     (3-per-10 to   (3-per-10 to  (1-per-10 to year ILRT)     year ILRT)       year ILRT)       1-per-10)      1-per-15)     1-per-15)
Dose Rate       3.65E-03       1.22E-02         1.83E-02        8.53E-03       1.46E-02      6.10E-03 Corrosion Likelihood X    3.93E-03       1.74E-02         3.63E-02        1.35E-02       3.24E-02      1.90E-02 1000 Corrosion Likelihood X    6.38E-03       6.43E-02         1.99E-01        5.79E-02       1.93E-01      1.35E-01 10000 Corrosion Likelihood X    3.10E-02       5.34E-01         1.83E+00        5.03E-01       1.80E+00     1.29E+00 100000 Revision 0A                                                                                Page 31 of 79

54002-CALC-01                             Evaluation of Risk Significance of Permanent ILRT Extension 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 5-24 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 RNP Summary of ILRT Extension Using Expert Elicitation Values (from Reference 24)
Leakage Size (La)     Jeffreys Non-Informative      Expert Elicitation Mean         Percent Reduction Prior              Probability of Occurrence 10                    2.70E-02                     3.88E-03                       86%
100                    2.70E-03                     9.86E-04                       64%
Taking the baseline analysis and using the values provided in Tables 5-12 and 5-13 for the expert elicitation sensitivity yields the results in Table 5-25.
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54002-CALC-01                                 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 RNP Summary of ILRT Extension Using Expert Elicitation Values Accident                                                  ILRT Interval Class 3 per 10 Years                         1 per 10 Years             1 per 15 Years Base         Adjusted       Dose        Dose      Frequency       Dose      Frequency        Dose Frequency         Base       (person-       Rate                      Rate                       Rate Frequency        rem)      (person-                  (person-                   (person-rem/yr)                   rem/yr)                    rem/yr) 1        1.41E-05       1.40E-05      1.56E+03     2.19E-02    1.38E-05      2.16E-02     1.37E-05      2.14E-02 2        0.00E+00      0.00E+00       2.22E+06    0.00E+00     0.00E+00     0.00E+00      0.00E+00      0.00E+00 3a          N/A         6.40E-08      1.56E+04     9.98E-04    2.13E-07      3.33E-03     3.20E-07      4.99E-03 3b          N/A         1.63E-08      1.56E+05     2.54E-03    5.42E-08      8.45E-03     8.13E-08      1.27E-02 6        1.15E-06       1.15E-06      8.85E+06    1.01E+01     1.15E-06     1.01E+01      1.15E-06      1.01E+01 7        1.23E-06       1.23E-06      2.62E+06    3.22E+00     1.23E-06     3.22E+00      1.23E-06      3.22E+00 8        7.03E-07       7.03E-07      2.45E+07    1.73E+01     7.03E-07     1.73E+01      7.03E-07      1.73E+01 Totals     1.72E-05       1.72E-05      3.84E+07    3.07E+01     1.72E-05     3.07E+01      1.72E-05      3.07E+01 LERF (3 per 10                            N/A                                 3.79E-08                   6.50E-08 yrs base)
LERF (1 per 10                            N/A                                    N/A                     2.71E-08 yrs base)
CCFP                            18.01%                                  18.23%                     18.39%
The results illustrate how the expert elicitation reduces the overall change in LERF and the overall results are more favorable with regard to the change in risk.
5.3.4     Large Leak Probability Sensitivity Study The large leak probability is a vital portion of determining the Class 3b frequency. A sensitivity is done by using the WCAP method as an alternative to the EPRI method for calculating the large leak probability. Table 5-26 presents the large leak probabilities for the baseline test, 10 year test interval, and 15 year test interval. Table 5-27 was developed using the same process as to calculate Class 3b.
Table 5 RNP Large Leak Probabilities Using the WCAP Method Test Interval        WCAP Large Leak Probability           EPRI Accident Class 3b Frequency 3 per 10 years                    2.47E-4                                  3.46E-09 10 years                       7.41E-4                                  1.02E-08 15 years                       1.11E-3                                  1.50E-08 Using the same EPRI approach, but with an updated Class 3b frequency calculated from the WCAP large leak probability data, Table 5-27 contains the final results.
Table 5 Impact on LERF due to Extended Type A Testing Intervals with WCAP CDF ILRT Inspection Interval           3 Years (baseline)       10 Years         15 Years Class 3b (Type A LERF)                  3.46E-09           1.02E-08         1.50E-08 LERF (3 year baseline)                                    6.71E-09         1.16E-08 LERF (10 year baseline)                                                     4.85E-09 These results demonstrate that the EPRI methodology is conservative when used to calculate a large leak probability as compared to the WCAP method.
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54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension 6.0      RESULTS The results from this ILRT extension risk assessment for RNP are summarized in Table 6-1.
Table 6-1 -ILRT Extension Summary Class       Dose              Base Case                    Extend to              Extend to (person-           3 in 10 Years               1 in 10 Years          1 in 15 Years rem)
CDF/Year          Person-      CDF/Year      Person-  CDF/Year      Person-Rem/Year                   Rem/Year               Rem/Year 1       1.56E+03     1.40E-05         2.18E-02     1.37E-05      2.14E-02 1.35E-05     2.11E-02 2       2.22E+06    0.00E+00         0.00E+00      0.00E+00      0.00E+00 0.00E+00     0.00E+00 3a      1.56E+04     9.41E-08         1.47E-03     3.14E-07      4.89E-03 4.70E-07     7.34E-03 3b      1.56E+05     2.34E-08         3.65E-03     7.80E-08      1.22E-02 1.17E-07     1.83E-02 6       0.00E+00     1.15E-06        1.01E+01      1.15E-06      1.01E+01 1.15E-06     1.01E+01 7       0.00E+00     1.23E-06        3.22E+00      1.23E-06      3.22E+00 1.23E-06     3.22E+00 8       8.85E+06     7.03E-07        1.73E+01      7.03E-07      1.73E+01 7.03E-07     1.73E+01 Total                  1.72E-05        3.07E+01      1.72E-05      3.07E+01 1.72E-05     3.07E+01 ILRT Dose Rate from 3a and 3b From 3 N/A                       1.15E-02               1.97E-02 Total      Years Dose Rate    From 10 N/A                          N/A                 8.23E-03 Years From 3 N/A                       0.0376%                0.0644%
   %Dose       Years Rate      From 10 N/A                          N/A                 0.0268%
Years 3b Frequency (LERF)
From 3 N/A                       5.46E-08               9.36E-08 Years LERF From 10 N/A                          N/A                 3.90E-08 Years CCFP %
From 3 N/A                        0.318%                 0.545%
Years CCFP%
From 10 N/A                          N/A                  0.227%
Years Revision 0A                                                                               Page 34 of 79

54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension

==7.0      CONCLUSION==
S AND RECOMMENDATIONS Based on the results from Section 5.2 and the sensitivity calculations presented in Section 5.3, the following conclusions regarding the assessment of the plant risk are associated with extending the Type A ILRT test frequency to 15 years:
Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.0E-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 9.36E-8/year using the EPRI guidance. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].
The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.020 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 from the 3 in 10 year interval to 1 in 15 year interval is 0.545%. 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 RNP risk profile.
Previous Assessments The NRC in NUREG-1493 [Reference 6] has previously concluded that:
Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.
Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.
The findings for RNP confirm these general findings on a plant-specific basis considering the severe accidents evaluated for RNP, the RNP containment failure modes, and the local population surrounding RNP.
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54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension A.       ATTACHMENT 1 Internal Events PRA Quality Statement for Permanent 15-Year ILRT Extension The RNP Internal Events PRA model RNP_12 is utilized to calculate CDF, LERF, and PDSs for the permanent 15-year ILRT extension. Any elements of the supporting requirements detailed in ASME/ANS RA-Sa-2009 that could be significantly affected by the application are required to meet Capability Category II requirements.
The Internal Events PRA provides an adequate base model for the development of the permanent 15-year ILRT extension. While there has been one instance of a change to the internal events model since the 2010 peer review that could be considered a model upgrade, this change is specifically applicable to the loss of CCW initiating event and does not affect the permanent 15-year ILRT extension results.
To ensure that the current Internal Events PRA model remains an accurate reflection of the as-built, as-operated plant, the following configuration control activities are routinely performed:
Design changes and procedure changes are reviewed for their impact on the Internal Events PRA model. PRA screening is required for all design and procedure changes.
New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.
Plant-specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated based upon reviews of plant program data, particularly data supporting the Maintenance Rule.
The Robinson Internal Events PRA is based on a detailed model of the plant developed from the Individual Plant Examination for Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities." The model is maintained and updated in accordance with Robinson procedures, and has been updated to meet the ASME PRA Standard and Revision 2 of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities."
In accordance with RG 1.200, Rev. 2 and ASME/ANS RA-Sa-2009, a full scope Peer Review of the Internal Events model of the RNP PRA was conducted in May 2010 by six PRA experts (Attachment U of Reference 38). The review provided findings and suggestions regarding the model and identified 25 supporting requirements within the internal events portion of the model that did not meet Capability Category II. These findings were either resolved by additional analysis and included in the quantitative results, or evaluated with their impact on the applicable risk evaluation. The dispositions for these findings are presented in Table 1, Internal Events PRA Peer Review - Facts and Observations. All findings which had significant impact on this analysis have been addressed. The ILRT application was determined to be an application requiring a Capability Category II PRA model per the Regulatory Guide 1.200 criteria, Revision
: 2. This is based on the requirement for numerical results for CDF and LERF to determine the risk impact of the requested change and the fact that this change is risk-informed, not risk-based. Table 1 includes discussion of all findings from the industry peer review along with the assessment and evaluation of the finding that shows that they have either been addressed or have no material impact on the ILRT interval extension request.
Following disposition, all of the SRs evaluated Met Capability Category II or better.
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54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension Fire PRA Quality Statement for Permanent 15-Year ILRT Extension In accordance with RG 1.205 position 4.3:
The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard supporting requirements important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable.
The Robinson Internal Events model is also updated to support the Robinson Fire PRA. The RNP Combined Internal Events and Fire PRA was peer reviewed during the period of March 2013 (Attachment V of Reference 38). The peer review was conducted by a team of industry personnel (utility and vendor). The Westinghouse Owners Group performed the review and has documented the outcome via LTR-RAM-13-06 Fire PRA Peer Review of the H. B. Robinson Nuclear Plant Fire Probabilistic Risk Assessment against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard. As noted in LTR-RAM-13-06, the RNP Fire PRA was found to be consistent with the ASME/ANS Standard and is suitable for supporting risk-informed applications.
The peer review team noted a number of Facts and Observations (F&Os). As documented in LTR-RAM-13-06, 85% of the Supporting Requirements (SRs) were assessed at Capability Category II or higher. Approximately eighteen Finding level and nine Suggestion level F&Os were identified during the peer review conducted in March 2013. Duke Energy recognized that CDF and LERF were relatively high, as noted in LTR-RAM-13-06. Based on the CDF and LERF values at the time of the initial peer review, coupled with the number of findings associated with the Fire Scenario Selection (FSS) Technical Element (18), Duke Energy decided to have a focused peer review.
The focused peer review was conducted during the period of July 2013 and evaluated the FSS Technical Element based on refinements to approved methodologies and updated documentation. The focused peer review was conducted by Frederick Mowrer (C P Fire, LLC) and Bijan Najafi (JENSEN HUGHES, Inc.) and is documented via JENSEN HUGHES Calculation No. 0004-0042-415-RPT-001, Robinson Nuclear Plant Fire PRA Focused Peer Review, Revision 0. As noted in LTR-RAM-13-06 and JENSEN HUGHES Calculation No. 0004-0042-415-RPT-001, the Fire PRA does apply the methodologies outlined in NUREG/CR-6850 correctly, is consistent with the ASME/ANS Standard and is applicable for supporting risk-informed applications. Although several of the initial F&Os were resolved, seven new findings and three new suggestions were identified during the focused-scope peer review.
Table 2 documents the Finding level F&Os associated with both the initial and focused peer reviews. All but two findings have been dispositioned as meeting Capability Category II or better. FSS-E3-01, which was assessed as meeting Category I, involved providing a mean value of, and statistical representation of, the uncertainty intervals for the parameters used for modeling the significant fire scenarios. Although no change has yet been made that would improve the Capability Category assessment, RNP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because documenting the statistical representation of uncertainty intervals will not change the quantified risk metrics. Supporting Requirement FSS-E1-01 has not been resolved, but this is a documentation issue that does not impact quantification. These two SRs not being resolved does not impact the ILRT extension analysis. The Fire PRA is adequate to support the ILRT extension.
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54002-CALC-01                                                                        Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic           Status                    Finding/Observation                                        Disposition ID)
AS-A5 (1)          Define the Accident  Dispositioned RNP-F/PSA-0043, Tables 4.2, 4.4, 4.5, 4.6, 4.7,     This is consistent with FRP-H.1 where Step 73a
[Cat I-III is MET] Sequence Model                     4.8, & 4.10 correlate safety function and           checks for SI pumps running and if not, the procedure success criteria to procedures and supporting       requires to go to Step 76 to verify pressurizer PORVs documents; The transient event tree (Figure         are all closed. If PORVs are not closed, the procedure 4.3) maps the progression of plant response to      directs to close associated PORV Block Valves.
an initiating event as is presented in the ESD. Therefore, description of sequence TBU in RNP-F/PSA-0043 section 4.2.3.5 is re-written to remove assumption.
Description of sequence TBU, in 0042, Section 4.2.3.5, "For sequences involving injection failure, it is assumed that the operators do not open the PORVs following the failure of safety injection. The RCS pressure increases until the pressurizer relief valves lift. The valves cycle to relieve pressure but successfully reclose. The potential for valve failure is addressed in the next sequence." Is this consistent with procedures and bounding for consequences?
AS-A5 (2)          Define the Accident  Dispositioned RNP-F/PSA-0043, Tables 4.2, 4.4, 4.5, 4.6, 4.7,     Transient-induced RCP seal failure results in transient
[Cat I-III is MET] Sequence Model                     4.8, & 4.10 correlate safety function and           similar to an S1 LOCA and is usually limiting in that an success criteria to procedures and supporting       S1 LOCA requires inventory makeup and secondary documents; The transient event tree (Figure         side heat removal. It is assumed that a small 4.3) maps the progression of plant response to      percentage of seal LOCA develops that would require an initiating event as is presented in the ESD. inventory makeup and heat removal using Safety Injection because of high leak rates, but the large breaks would occur after at least a 15 minute delay Describe how the range of seal LOCA break           resulting in reduction in transient severity. Section 4.4A flows, which can exceed small LOCA break flow       of RNP-F/PSA-0043 has been revised to explicitly were modeled.                                       include this assumption. Additionally, this assumption and uncertainty is captured in RNP-F/PSA-0074, RNP Uncertainty Analysis, for risk impact determination.
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54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic            Status                   Finding/Observation                                          Disposition ID)
AS-A5 (3)          Define the Accident    Dispositioned RNP-F/PSA-0043, Section 4.8, Consequential          An Induced SGTR model, based on the guidance in
[Cat I-III is MET] Sequence Model                       Pressure-induced SGTR not considered.               NUREG-1570, Risk Assessment of Severe "Consequential tube ruptures due to high            Accident-Induced Steam Generator Tube Rupture,
[Associated with                                        primary-to-secondary differential pressures         USNRC, March 1998, was developed and documented AS-B3]                                                  (e.g., due to secondary line breaks or ATWS         in RNP-F/PSA-0047. Sections 4.8 of RNP-F/PSA-0043 events) are not explicitly considered in the        and the Level 2 results have been updated to reflect model. Secondary line breaks can result in a        the induced SGTR model and results. More detailed primary-to-secondary differential pressure          modeling of Induced SGTR was added to the level 2 equivalent to RCS pressure."                        analysis as described in section 8.2.2 of RNP-F/PSA-0047, Rev. 2.
AS-B3 (1)          Identify               Dispositioned Section 7.5 of the system notebook addresses        A review of the systems and the impact of the initiating
[Cat I-III is NOT  phenomenological                     the requirements of SRs AS-B1 and AS-B3.            events on those systems were performed to ensure MET]               conditions created by                Based on a review of these section of several       that the impacts are captured in the model as the accident                         system notebooks, this reviewer did not identify    appropriate. A statement discussing the impact of the
[Associated with   progression                          any components or operator action that have         initiators on the system was added to section 7.1 of the AS-B1 and SY-                                           been considered to be impacted by any               system notebook for each system.
B14]                                                    initiating event (or accident sequences associated with the initiating event). Based on a   RESOLUTION OF CAPABILITY CATEGORY discussion with the Duke Energy's Lead PRA          CLASSIFICATION engineer, it is understood that the potential       Based on the disposition above, SR AS-B3 is impact of an IE on the systems credited to          considered to be MET at CAT I-III.
mitigate the consequences of the IE have been addressed. However, the discussions provided in the pertinent sections of the system notebooks are not adequate to provide a defensible and easily traceable documentation of the justification of 'no impact.' Given the relatively compact physical structure of the plant, housing of redundant components in the same location, and relatively high dependency on the operator action, a better documentation should be provided to support the 'no impact' assertion. For example, the potential impact of the SLB outside containment IE on the AFW function and the non-emergency buses should Revision 0A                                                                                                                                    Page 39 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic              Status                   Finding/Observation                                         Disposition ID) be included. Another potentially vulnerable system is the Fire Protection system, where the component of such a system is usually are not credited to work in harsh environment. The system notebook and the plant walk down documentation are thought to provide justification that this issue has been looked at.
Better documentation of the supporting analyses is needed to support the assertion of
                                                         'no impact'.
AS-C1 (1)          Correlation of system   Dispositioned RNP-F/PSA-0043 provides a reasonable               For each accident sequence events in
[Cat I-III is MET] top events and                        notebook with clear tables and discussions of      RNP-F/PSA-0043 success criteria tables, system top Accident Sequence                     work performed. However, it is difficult to        event gate names are added in the Supporting safety function events                correlate between system top events and            Documentation column.
accident sequence safety function events.
AS-C2 (1)          Correlation of Type B   Dispositioned RNP-F/PSA-0006 on Initiating Events identifies     RNP-F/PSA-0014 is added to Document Indexing
[Cat I-III is MET] & C HRA and                           transients and LOCAs covered in Accident           Table and Section 4.0, References, of Accident Sequence                     Sequence Analysis. Operator actions are            RNP-F/PSA-0006.
Analysis                              discussed in tables and narrative, but HRA Type B & C Notebook RNP-F/PSA-0014 is not on reference list. A tie to the HRA notebook is needed for clarity.
AS-C2 (2)          Accident sequence       Dispositioned RNP-F/PSA-0006 on Initiating Events identifies     The MTC values selected from WCAP-15831-P and
[Cat I-III is MET] development                           transients and LOCAs covered in Accident           listed in Table 4.9 of RNP-F/PSA-0043 for the ATWS Sequence Analysis. Operator actions are            UET analysis are for a Low Reactivity Core Model with discussed in tables and narrative.                 a maximum HZP MTC of +3.5 pcm/&#xba; F (HFP MTC WCAP-15831-P is used as the basis for              equivalent of -7.42 pcm/&#xba; F). Per ANP-2887 -cycle Moderator Temperature Coefficient (MTC)            reload report for cycle 27, the RNP MTC at HZP is -
Unfavorable Exposure Time (UET). No                0.35 pcm/&#xba; F at BOC and -11.29 pcm/&#xba; F at HFP. The explanation provided as to how this data reflects  WCAP analysis used a bounding MTC that was current plant operation, refueling cycles, or fuel positive. For RNP, the MTC is negative, therefore configuration.                                     bounded by the MTC values of the Low Reactivity Core Model used in the WCAP. Additionally, PSA is on the Revision 0A                                                                                                                                  Page 40 of 79

54002-CALC-01                                                                       Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic           Status                   Finding/Observation                                           Disposition ID)
Reload Review panel and will be evaluating current plant data to existing PSA analysis as indicated in Note under Table 4.9 of RNP-F/PSA-0043.
DA-A2 (1 and 2)   Definitions for data  Dispositioned 1) SSC boundaries are defined in                       Consideration of the AFW room cooling as an air handling
[Cat I-III is NOT analysis                            RNP-F/PSA-0077, Attachment 2. No boundary              unit does not fit the manner in which the components have MET]                                                  definition is defined for HVAC air handling units or   been modeled. Given the system configuration in the plant coolers. The AFW coolers are modeled as                the separation of the components is more appropriate.
separate fans and heat exchangers. These are           NUREG-CR/6928 provides values for the as modeled more appropriately grouped as an air handling unit     configuration and these are utilized.
per NUREG-CR-9628. The failure boundary for the Diesel Generator output breaker is not             The definition of the EDG component boundary is correct consistent with the system analysis modeling and       in that it includes the output breaker. The event the generic failure data. Attachment 2 has the         representing the failure of the output breakers breaker within the component boundary, but the         (PCB5217BNN and PCB5227BNN) to close will be set to breaker is modeled separately. Control room hand       zero in the model since failures of this component are switches are included in the component boundary        rolled into failures of the EDG.
per Attachment 2, even though they are not local as described in NUREG-CR/6968, this appears to         The control circuitry which is included in the NUREG-be a grey area, though. Logic for control valve        CR/6928 component boundary definitions is interpreted to automation is not within the component boundary,       include the control room hand switches as well as local though NUREG-CR/6968 states local                      switches. Therefore the definition of component boundaries instrumentation and controls are within the            as defined in Attachment 2 of the RNP data calculation boundary. For pumps, it appears that separate          RNP-F/PSA-0072 are consistent with the boundary cooling and lubrication systems are not within         definitions identified in NUREG-CR/6928.
scope, though these systems would be included in NUREG-CR/6928 if local. The peer review did not        Component boundaries for control valves as defined in include a comprehensive review of component            Attachment 2 of the RNP data calculation RNP-F/PSA-boundaries. The issues found, though individually      0072 are consistent with the boundary definitions identified not a finding, show a trend of inconsistencies         in NUREG-CR/6928 in that they specify that local switches, between the plant model and component                  contacts, relays, and control circuitry are included in the boundary definitions. 2) No definitions of failure     boundary. Logic for an automatic actuation is not identified modes were found. 3) No definitions of success         in NUREG-CR/6928 definitions and therefore is defined to criteria were found.                                   be outside of the control valve boundary in Attachment 2.
Attachment 2 of the RNP data calculation RNP-F/PSA-0072 specifies that, with regard to pumps, if a Revision 0A                                                                                                                                    Page 41 of 79

54002-CALC-01                                                  Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation      Topic    Status                Finding/Observation                                     Disposition ID) different system is providing additional cooling or lubrication then this component is considered to be separate and not within the boundaries of the pump being considered. NUREG-CR/6928 identifies that only local lubrication or cooling systems are to be included in pump boundaries and Attachment 2 boundary definitions are in agreement with this specification.
With regard to the battery chargers the input and output breakers are modeled separately due to the fact that plant specific data is utilized in the model for battery chargers.
Thus the manner in which the component is modeled is consistent with those definitions put forth in Attachment 2 of the RNP data calculation RNP-F/PSA-0072. Generic data values are utilized to model the breakers.
A review was made of other component boundaries and it was determined the boundaries detailed in Attachment 2 of the data calculation (RNP-F/PSA-0072) and in the model are appropriate when compared to those set forth in NUREG-CR/6928.
Wording has been added to Section 4.3 of the data calculation (RNP-F/PSA-0072) to provide more detailed information on the identification of events as failures. There are definitions in the RNP scoping documents that are used in the evaluation of functional failures for the Maintenance Rule. The MR database is then used to capture all failure events. These events are then evaluated and screening comments provided to determine if they count as failures towards the PSA.
Success criteria are comprehensively defined for each modeled system in the respective system notebooks (RNP-F/PSA-0018).
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54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic            Status                  Finding/Observation                                          Disposition ID)
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR DA-A2 is considered to be MET at CAT I-III.
DA-C1 (1)          Parameter estimation  Dispositioned Some generic data collected from a recognized      The spreadsheet which documents and aggregates the
[Cat I-III is MET]                                     source, NUREG-CR/6928. Other generic data          generic data (GENDATA RNP 2011.xlsx) identifies collected from a Progress Energy source called     which data source contributes for each component and
                                                       'Generic Data Aggregation.' Unable to find         failure mode. This includes the numerical values that documentation of process and methodology           are used for each source. The Section 3.2 of the data used for data collection and analysis for Generic  calculation (RNP-F/PSA-0072) provides references for Data Aggregation.' Generic data is not used for    each contributor to the generic data aggregation.
unavailability. A cursory review of Generic Data Aggregation showed that the failure rates were within reason based on similar failure rates in NUREG-CR/6928.
DA-C14 (1)         Intersystem           Dispositioned No inter-system coincident maintenance             PRA personnel reviewed EOOS entries for instances in
[Cat I-III is MET] Unavailability                      analysis was found to be performed.                which intersystem coincident maintenance occurred.
While examples of such events could be found, none were the result of planned, repetitive maintenance activities. Therefore, using the guidance of the ASME standard (ASME/ANS RA-Sa-2009) it was determined no additional events should be created for these occurrences.
DA-C3 (1)          Plant data screening  Dispositioned Plant specific data originates from Maintenance    The wording in Section 4.3 of the data calculation
[Cat I-III is MET]                                     Rule functional failure data. Justifications are   (RNP-F/PSA-0072) has been adjusted to articulate that provided for screening and disregarding plant-     the MR database is used to capture all failure events.
specific data in RNP_Fail_Data.xls. Section 4.3    These events are then evaluated and screening of RNP-F/PSA-0072 states that "A failure would     comments provided to determine if they count as be counted if a component tripped for no           failures towards the PSA. The statement from Section apparent cause and was later restarted with no     4.3 of the data calculation quoted in this F&O has been corrective activity. This type of event may not be removed from the calculation due to an inaccurate counted as a MR functional failure..." However, Revision 0A                                                                                                                                   Page 43 of 79

54002-CALC-01                                                                      Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic        Status                   Finding/Observation                                         Disposition ID) this type of failure would be screened out by the  representation of the process used in evaluating failure identification process that starts with MR failures.
functional failures.
DA-C4 (1)          Plant component    Dispositioned For maintenance rule functional failures and       This is primarily a documentation issue and wording
[Cat I-III is NOT  failure data                     unavailability events, the reviewer was unable     has been added to Section 4.3 of the data calculation MET]                                                to find system or component specific definitions   (RNP-F/PSA-0072) to provide more detailed of these MR performance criteria. ADM-NGGC-        information on the identification of events as failures.
0101 states that PSA should be considered as       There are definitions in the RNP scoping documents an input into MR performance criteria              that are used in the evaluation of functional failures for development. No evidence was found that MR         the Maintenance Rule. The MR database is then used System functional failure and unavailability       to capture all failure events. These events are then definitions have been compared to definitions or   evaluated and screening comments provided to scoping in PSA. This SR is not met.                determine if they count as failures towards the PSA.
RESOLUTION OF CAPABILITY CATEGORY No clear basis was discovered for identification   CLASSIFICATION of events as failures.                             The documentation has been updated and addressed accordingly. Therefore, SR DA-C4 is considered MET at CAT I-III.
DA-C6 (1)          Plant-specific     Dispositioned In practice, PI (computer point) data is used for  Further documentation has been added to Section 4.4
[Cat I-III is MET] component demands                estimating demands on standby components,          of the data calculation (RNP-F/PSA-0072) to where the data points exist. According to          demonstrate that post maintenance testing is excluded Section 4.4 of RNP-F/PSA-0072, estimates of        from the collection of demand data. The component operation based on normal operating practices,     data based on OSI-PI utilizes a one hour screening surveillance procedures (SPs), and System          value. Other demands are based on test procedures or Engineer input were used for the remaining         system engineer input and only AOVs and MOVs, Batteries, Battery Chargers,        include those demands specified during the test or Air Compressors, Dedicated Shutdown Diesel         during normal operations.
Generator, Diesel Driven Fire Pump, Motor Driven Fire Pump, and Deepwell Pumps. There is no evidence that demands from post maintenance testing are excluded. No unusual Revision 0A                                                                                                                                Page 44 of 79

54002-CALC-01                                                                  Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation             Topic         Status                  Finding/Observation                                         Disposition ID) number of demands as multiple demands in an hour are excluded.
DA-C8 (1)         Component          Dispositioned The flag events that represent the time that      Data was obtained and analyzed to determine the
[Cat I is MET]    configurations in                components were configured in their standby       runtime percentages of all possible alignment standby                          status do not appear to use actual plant data as  configurations for select plant systems. These plant a primary source.                                 systems were selected based on those for which data was available and those which utilized alignment flags in previous model revisions. New basic events were put into the model to replace the former alignment flags. The new events represent each possible running configuration and have a value equal to the percentage that the alignment was found to be operating.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With the incorporation of plant data into this analysis, SR DA-C8 is considered to be MET at CAT II/III is met.
DA-D4 (1)         Bayesian approach  Dispositioned No discussion is provided to demonstrate the      Section 5.1 of the data calculation (RNP-F/PSA-0072)
[Cat I is MET]                                     posterior distributions are reasonable given the  provides identification of ways in which the Bayesian relative weight of evidence provided by the prior results have been validated using guidance from the and the plant-specific data.                      standard. Both generic and plant specific data was reviewed and spreadsheet (RNP_Bayes_Data_R1.xlsx) calculations analyzed for appropriateness in the performance of the Bayesian update. Results of the update were checked for appropriateness using at least one of the methods put forth in the standard.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With the incorporation of the posterior distributions to the Bayesian approach, CAT II/III is considered to be MET.
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54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic             Status                  Finding/Observation                                            Disposition ID)
DA-D6 (1)           Common Cause           Dispositioned No evidence that a review was performed of           Performed a review for plant experience with common
[Cat I is MET]      Failure probabilities                generic common cause failure probabilities to        cause failures, and updated RNP-F/PSA-0073 Section be consistent with available plant experience.       2.0 to include this discussion.
There is no discussion of available plant experience with respect to common cause              RESOLUTION OF CAPABILITY CATEGORY failures in RNP-F/PSA-0073.                          CLASSIFICATION Incorporation of plant experience qualifies for meeting CAT II requirements for DA-D6.
DA-E2 (1)           Data analysis process  Dispositioned The PRA documentation does not show the              Development and discussion of the occurrences
[Cat I-III is MET]  documentation                        probabilities or basis for the probabilities for the represented by the basic events cited in the F&O may following basic events (note that the basic          be found in the Westinghouse Owners Group events are assigned probabilities in the fault       document WOG Risk-Informed ATWS Assessment tree): #ACRDMF (control rod fails due to             and Licensing Implementation Process (WCAP-mechanical binding), #CRDMF (insufficient rod        15831-P) which is cited in the RNP Success Criteria insertion), #RPS (failure of reactor trip),          calculation (RNP-F/PSA-0075). Additionally, EAMSAC (AMSAC failure), ESFAS (ESFAS                 information regarding to the performance of a fails), CCVENT (performing CV purge), and            containment purge may be found in the Technical GINRDOORSL (personnel hatch door gasket).            Specifications for RNP, Sections 3.3.6, 3.6, and 3.9.3.
No evidence was found that these basic event values are not appropriate.
HR-H1           (1) Operator recovery      Dispositioned Section E.5 RNP-F/PSA-0014 Rev. 1 discusses          As indicated in Section E.5 of RNP-F/PSA-0014, the
[Cat II is MET]     actions                              TYPE Cr: POST-INITIATOR RECOVERY                     only non-proceduralized (Type CR) human interactions ACTIONS; states that the only recovery human         included in the Robinson PSA model are in the Plant actions included in the Robinson PSA model           flooding analysis (RNP-F/PSA-0009). OPER-48 and are in the Plant flooding analysis, but Table A-1    OPER-51 are not Type CR.
indicates OPER-48 recovery of the loss of IA to the SG PORVs via the use of the Steam Dump N2 accumulator and OPER-51 Failure to recover RCP seal injection with alternate filter line. Documentation of operator recovery actions is not clear and consistent.
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54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation                Topic            Status                   Finding/Observation                                         Disposition ID)
HR-I1 (1)          HRA documentation       Dispositioned The Type A pre-initiator process is unique and     Updated Figure 1-1, Pre-Initiator Human Failure Events
[Cat I-III is MET]                                       therefore requires some detailed review to         Flow Chart of RNP-F/PSA-0071 Rev. 2 to provide identify how HFEs were                             additional detail about how HFEs are identified.
identified for inclusion in the models.
IE-A6 (1)          Initiating events from  Dispositioned Several IE's are evaluated with a fault tree that  Section 1.1.4.10 and 1.1.4.11 of RNP-F/PSA-0006,
[Cat II is MET]    multiple failures                     models the loss of a system. Such fault trees      Rev 3 has been updated to include consideration.
typically include common cause failures and various system alignments caused by test and maintenance unavailability. No consideration is documented for potential unique initiating events for common cause failures of both DC busses or both AC busses. The consequences of such initiating events would be more severe than an initiating event where a single bus fails.
IE-C3 (1)          Recovery action credit  Dispositioned The screening of the PORV opening due to           Should pressure transmitter PT-445 fail high, a
[Cat I-III is MET]                                       failure of PT-445 needs to be investigated as a    demand for the PCV-456 will be generated. In order for potential unique initiator.                        the valve to open, the P-11 interlock must be defeated (i.e., RCS pressure > 2000 psi). Should this occur, the opening of the valve will result in a decrease in RCS pressure and a reactor trip on low pressurizer pressure. This event is similar to a small LOCA with a frequency of 5E-4 per year. However, it can be mitigated if the operator closes either the PORV or its associated block valve. When mitigated, it would be similar to the reactor trip initiator of 2.53E-01 per year.
When the potential for operator recovery is included, the initiating frequency for this event is much less than the small LOCA initiating event frequency: From RNP-F/PSA-0072, RNP Failure Data Analysis, the probability of PORV Spurious Operation is 3.31E-07 per hour, PORV Fails to Close is 6.70E-4 per demand, PORV Block Valve Fails to Close is 1.64E-3 per demand and from RNP-F/PSA-0014 the potential for operator recovery is 2.6E-3. The frequency of a Revision 0A                                                                                                                                       Page 47 of 79

54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic          Status                  Finding/Observation                                            Disposition ID) spurious PORV opening and failure to reclose is thus 1.5E-5 per year (Note: 2 PORVs) which is small compared to the small LOCA frequency of 5E-4 per year. In addition, it does not result in any other significant impacts on safety systems. Therefore, a spurious PORV opening is not considered to require separate consideration in the RNP PRA model. RNP-F/PSA-0006, Section 1.1.4.6 is updated to include this justification.
IE-D3 (1)         Initiating event      Dispositioned A qualitative assessment of identified IE            Rationale for using lognormal distribution is
[Cat I-III is NOT analysis                            uncertainties is presented in RNP-F/PSA-0074.        documented in RNP-F/PSA-0049, Section 6.5 and MET]                                                  No basis provided for selection of uncertainty       Table 12 note 1 in RNP-F/PSA-0006 is updated distributions.                                       accordingly.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated to provide the appropriate basis for this methodology and addressed accordingly. Therefore, SR IE-D3 is considered to be MET at CAT I-III.
LE-C11 (1)        Containment failure   Dispositioned RNP-F/PSA-0047 credits containment spray             Level 2 operator actions were assessed in more detail
[Cat I is MET]    impacts on equipment                operation following containment failure for small    using industry-standard methodologies via the EPRI and operator actions                and late releases, but not for large early           HRA calculator. As part of that assessment,
[Associated with                                      releases. Manual isolation operator actions are      adjustments were made as appropriate to the LE-C12]                                               described in Sections 8.3.6 and 8.3.7. No            environmental conditions under the actions that would evidence of justification of credit for equipment    be performed. In this context, assessment of survivability or justification of credit for human   Containment Spray success is used to determine actions that could potentially be impacted by        whether a containment release (non-ISLOCA or SGTR) containment failure.                                 is scrubbed or unscrubbed. In many end-states, in the Containment Event Tree, this determines whether the release is a Large Early release or not. The success or failure of the Containment Spray system to perform its credited functions is determined through the Revision 0A                                                                                                                                    Page 48 of 79

54002-CALC-01                                                                     Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation             Topic          Status                  Finding/Observation                                          Disposition ID) development of failure and unavailability data using industry-approved sources and/or plant-specific data.
Furthermore, the Containment Spray system is an ECCS mitigation system and thus designed to sustain the radiation and temperature effects of post-accident conditions postulated in the design basis. No environmental conditions were identified which required the Containment Spray system to operate beyond their design basis. Thus, it is reasonable to conclude that the Containment Spray system would be available (within the constraints of the failure and unavailability data) during a post-accident containment failure.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With adequate justification for credit of equipment survivability and human actions impacted by containment failure, this SR is considered to be MET at CAT II/III.
LE-C2 (1)         Operator actions    Dispositioned LERF-related operator actions following the         Level 2 operator actions were assessed in more detail
[Cat I is MET]    following onset of                onset of core damage are described in the HRA       using industry-standard methodologies via the EPRI core damage                       notebook (RNP-F/PSA-0014, Annex B) and              HRA calculator.
RNP-F/PSA-0047. Screening values were set to 0.1 to see if there was sensitivity of results. For RESOLUTION OF CAPABILITY CATEGORY example, operator action to mitigate ISLOCA is      CLASSIFICATION evaluated in Section 8.2 of RNP-F/PSA-0047.         Through analysis of human operator actions with HRA The probability of failure is assumed to be 0.1     calculator as a more thorough approach, CAT II/III is based on "available guidance and time               considered to be MET.
available." The scenarios for these operator actions were not adequately documented as consistent with applicable procedures. Need to ensure applicability to the current as built/as operated plant.
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54002-CALC-01                                                                        Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic           Status                   Finding/Observation                                          Disposition ID)
LE-C4 (1)         LERF model logic for  Dispositioned The accident sequences are based on generic         Assessment performed in RNP-F/PSA-0062.
[Cat I is MET]    accident progression                references and on plant specific MAAP analysis. Important mitigation actions in           Based on this assessment, and based on the fact that significant accident progression sequences,         SI-864A and SI-864B are in series in the RWST supply such as SG isolation for SGTR are not               line and are in close proximity to one another, it is modeled. Evidence of technical justification for    reasonable to conclude that the ISLOCA pathway to demonstrating the feasibility of mitigating         the RWST could, at worst case, be isolated by closing actions was not provided. Scrubbing is              one valve until it can be closed no further and then discussed for release categories in                 closing the other until it can be closed no further, and RNP-F/PSA-0048 and brief rationale is provided      so on going back and forth between the valves until the when it is credited. Inclusion of beneficial        line is isolated.
failures was not observed.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The assessment and justification provided above are considered to meet CAT II for SR LE-C4.
LE-C9 (1)         Human actions under   Dispositioned Four LERF HFEs documented in Sections 8.2,          Level 2 operator actions were assessed in more detail
[Cat II-III is    adverse environments                8.3.3, 8.3.6, 8.3.7, and 8.3.11 of                  using industry-standard methodologies via the EPRI MET]                                                  RNP-F/PSA-0047 are modeled. Evidence of             HRA calculator. As part of that assessment, justification for human actions under adverse       adjustments were made as appropriate to the environments was not found in the LERF HFE          environmental conditions under the actions would be documentation. One of the HFEs, OPER-ISOL,          performed.
represents local closure of manual valves given failure of containment isolation. The environmental conditions associated with these valves could potentially have an impact on the operator's ability to locally close the valves and needs to be addressed. Although this action was not credited, the documentation should be clear about this.
LE-D5 (1)         Secondary side        Dispositioned Based on a review of a number of calculation        Assumption of 10 cycles for SG PORVs is not used in
[Cat I is NOT     isolation capability                files, SGTR event tree, and fault tree for #RW, it  the Level 2 analysis. For induced SGTR analysis that MET]              analysis                            appears that a realistic secondary side isolation   was added to the Level 2 analysis, an assumption of capability analysis is performed. However, there    40 cycles was used. Since one stuck-open SG SRV is Revision 0A                                                                                                                                  Page 50 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation             Topic             Status                    Finding/Observation                                          Disposition ID) are a number of important issues that their           considered sufficient for a large early release, the resolutions may impact this conclusion. These         modeling of multiple SRVs sticking open is extraneous.
issues include a) A justification for the assumed number of cycles for the PORVs on the faulted         RESOLUTION OF CAPABILITY CATEGORY SG (which is 10) is not provided. Also, it seems      CLASSIFICATION that only 'fail to close' of one SRV is considered    Based on the above disposition LE-D5 is considered to in the fault tree. b) Additionally, it appears that   be MET at CAT II.
Operator action to isolate the faulted SG is not included as a potential contributor to the secondary side isolation failure probability.
Given that in the current model 1) fail to isolate probability is dominated by the probability of a PORV sticking open, and 2) the SGTR IE is one of the highest contributors to the LERF figure of merit, it is important to ensure that the above apparent issues are fully resolved.
LE-D6 (1)          Thermally-Induced     Dispositioned Based on a review of Sections 8.2.4 and 8.3.7         More detailed modeling of Induced SGTR was added
[Cat I is MET]     Steam Generator                     of RNP-F/PSA-0047 a simple representation of          to the level 2 analysis as described in section 8.2.2 of (SG) Tube Rupture                   thermally induced SGTR is performed on a              RNP-F/PSA-0047.
Analysis                            generic basis. Note that 1.200 Gap Analysis comment with respect to ISGTR is a valid              RESOLUTION OF CAPABILITY CATEGORY comment. The response to the comment seems            CLASSIFICATION to be adequate but could not be verified.             Based on the above disposition LE-D6 is considered to be MET at CAT II.
LE-E1 (1)          Parameter values for  Dispositioned The system level data are documented in the           Description of development of input parameters is
[Cat I-III is MET] equipment and                       Level 1 analysis and some Level 2-specific            provided in RNP-F/PSA-0047. In many cases, they are operator response                   parameters are documented in Attachment 2 of          based on expert judgment.
RNP-F/PSA-0047. The basis for some of the values was subjective (e.g., see Table 8-10) and others are based primarily on the IPE submittal and supporting MAAP runs for the Level 2 analysis (e.g., Table 8- 35). Although it is understood that the approach for estimating probabilities for some Level 2 parameters is Revision 0A                                                                                                                                     Page 51 of 79

54002-CALC-01                                                                        Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic            Status                   Finding/Observation                                        Disposition ID) soft, using IPE results without additional justification may not be as realistic as required for a CC II classification.
LE-F1 (1)          Parameter estimates   Dispositioned Relative contribution to LERF is presented in      Induced SGTR has been added to the Level 2 model
[Cat II-III is     for accident                        the quantification notebook (RNP-F/PSA-0077)       for RNP. It is the dominant contributor for large bypass MET]               progression                         and the source term development notebook           containment failure sequences. Dominant contributors phenomena                           (RNP-F/PSA-0048). Induced SGTR contribution        to LERF and containment failure are described in was not presented among the significant LERF       sections 3.16 and 3.17 of the quantification calculation contributor results presented in Section 3.16 of   (RNP-F/PSA-0077).
RNP-F/PSA-0077.
LE-F2 (1)          Documentation of      Dispositioned The updated quantification notebook                Review of LERF contributors and results of that review
[Cat I-III is NOT  Contributors to LERF                (RNP-F/PSA-0077) was examined to find              are described in section 3.3 of the quantification MET]               analysis                            evidence of a review of the LERF contributors      calculation.
for reasonableness. All that was found were entries for "Level 2 HNPSUM Changes" toward        RESOLUTION OF CAPABILITY CATEGORY the end of the Appendix A "Change Log for          CLASSIFICATION RNP MOR09." Clear documentation is needed          Based on the above disposition LE-F2 is considered to to know what LERF reasonableness reviews           be MET at CAT I-III.
were performed and what were the results of the reviews to ensure that the SR is fully met.
LE-F3 (1)          Documentation of      Dispositioned The uncertainty analysis, as documented in         Assumptions and Uncertainties related to LERF are
[Cat I-III is MET] LERF analysis                       RNP-F/PSA-0074, contains very little               revised and documented in RNP-F/PSA-0074.
documentation of identification of LERF
[Associated with                                       uncertainties. Only about 10% of all the LE-G4]                                                 assumptions and uncertainties documented in Table 1 of RNP-F/PSA-0074, seem to be LERF related, even though LERF analysis as whole is supported by many conservative analysis assumptions. Additionally most of the identified uncertainties are not related to the biggest contributor to the LERF figure or merit (i.e., a SGTR event). This reviewer found only one (Item 195).
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54002-CALC-01                                                                        Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation                Topic          Status                  Finding/Observation                                        Disposition ID)
LE-G5           (1) LERF analysis        Dispositioned Limitations in the LERF analysis that would       RNP-F/PSA-0047, Rev. 2, Section 8.4.2 states that No
[Cat I-III is NOT   limitations                        impact applications are not identified.           limitations were identified in Level 2 analysis (including MET]                                                                                                     LERF) that would impact applications. However, any assumptions and uncertainties identified in the analysis are addressed in RNP-F/PSA-0074, RNP Uncertainty Analysis.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR LE-G5 is considered to be MET at CAT I-III.
QU-B3 (1)           Truncation analysis  Dispositioned The current truncation levels of 1E-10 for CDF    An appropriate lower truncation is used in
[Cat I-III is NOT                                      and LERF are insufficient to estimate the mean    RNP-F/PSA-0077, Rev. 2, and Section 2.2 truncation MET]                                                   CDF or LERF required by QU-A3 and LE-E4.          study shows convergence for both CDF and LERF.
The truncation study in Section 2.2 of RNP-F/PSA-0077 shows that truncation occurs for       RESOLUTION OF CAPABILITY CATEGORY CDF, but does not establish that it occurs at 1E- CLASSIFICATION
: 10. The study does not show convergence for       Based on the above disposition, SR QU-B3 is LERF. SR QU-B3 describes an acceptable            considered to be MET at CAT I-III.
process for demonstrating convergence, where a successive decade reduction in truncation results in a CDF or LERF change of less than 5%. The RNP results show a decade reduction from 1E-10 to 1E-11 results in a change of approximately 10% for CDF and 30% for LERF, which does not meet the example process in the SR.
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54002-CALC-01                                                                    Evaluation of Risk Significance of Permanent ILRT Extension QU-D1 (1)         Review significant Dispositioned Documented cutset reviews were inadequate.        The following has been added to section 3.3 of the
[Cat I-III is NOT cutsets to verify                Section 3.3 of RNP-F/PSA-0077 explains that       quantification calculation (RNP-F/PSA-0077): Due to MET]              model logic                      the top 50 cutsets were reviewed for CDF and      the low truncation settings and large quantity of cutsets LERF. Section 1.1 defines a significant cutsets   generated it is not practical to evaluate each cutset in as one of the top 95% of cutsets or a cutset that the 95th percentile. (Approximately 3800 cutsets contributes more than 1% to CDF. The top 50       comprise the top 95 %.) All of the top fifty cutsets were cutsets for CDF are near 50% for all CDF.         reviewed followed by a review of select cutsets at LERF cutsets are assumed to have a similar        varying intervals leading up to the 95% level. The top limitation. Review of non-significant cutsets     fifteen cutsets represent 50% of the CDF, and cutset appeared to be adequate. RNP-F/PSA-0077           17 is the point at which each subsequent cutsets states that 'A sample of non-significant cutsets  contribute less than 1% to CDF. Cutsets at the bottom listed in Table B-2 (lower than the 95% of the    of the top 50 each only contribute approximately 0.2%
CDF) was reviewed for reasonableness and to       to total CDF. In addition to the top 50, the bottom 50 ensure recoveries were being applied.'            cutsets were evaluated to increase confidence that the generated MOR12 was yielding logical results. A list of the 50 top core damage cutsets is provided in Table B-1 in Appendix B. These cutsets are presented in order of descending frequency. A sample of non-significant cutsets listed in Table B-2(lower than the 95% of the CDF) was reviewed for reasonableness and to ensure recoveries were being applied. Part of this review for reasonableness was reviewing which accident sequences were contributing to CDF and LERF, comparing those results to the previous model with consideration of what model changes were made, and performing sensitivity testing as needed to determine the reason for the differences between the previous model and this one.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-D1 is considered to be MET at CAT I-III.
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54002-CALC-01                                                                       Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic            Status                   Finding/Observation                                        Disposition ID)
QU-D3 (1)         Review cutsets for     Dispositioned Section 3.3 of RNP-F/PSA-0077, Cutset Results      Section 3.3 of RNP-F/PSA-0077, states that flag
[Cat I-III is NOT flag events, mutually                Review (Significant & Non-significant) indicates   events settings, mutually exclusive event rules and MET]              exclusive events, and                that "a panel of PRA experts reviewed the          recovery rules are reviewed for correctness and logical recovery rule                        cutsets for modeling accuracy, operational         results.
accuracy                             consistency and to ensure the correct recoveries were being applied." But only found     RESOLUTION OF CAPABILITY CATEGORY limited evidence that results were reviewed to     CLASSIFICATION determine that flag event settings and mutually    Based on the disposition above, SR QU-D3 is exclusive event rules yield logical results. Found considered to be MET at CAT I-III.
model changes listed in Table A-1 of RNP-F/PSA-0077 that indicated that results had been reviewed.
QU-D4 (1)         Compare results to     Dispositioned RNP-F/PSA-0077 Section 3.4, Similar Plant          Section 3.4 of RNP-F/PSA-0077, is updated to provide
[Cat I is MET]    similar plants                       Review, Table 7 shows a comparison of              adequate comparison. Comparison with the LERF was Robinson CDF and LERF values to Turkey             conducted in the cut-set review process. However, this Point and Beaver Valley. Plant systems             detailed review was not documented in the current differences are compared. However, the             MOR.
comparison is inadequate and sources of specific differences (e.g., LOCA contribution)     RESOLUTION OF CAPABILITY CATEGORY are not identified.                                CLASSIFICATION Based on the above disposition, SR QU-D4 is considered to be MET at CAT II-III.
QU-D7 (1)         Review                 Dispositioned Importance was calculated at basic event level,    Section 3.13 of RNP-F/PSA-0077, documents
[Cat I-III is NOT component/basic                      but could not be found at component level.         Component Importance.
MET]              event importance                     Review did not find documentation to substantiate review of component importance.       RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-D7 is considered to be MET at CAT I-III.
QU-F2 (1)         Model integration      Dispositioned Several items identified as typically documented   1. System successes are listed in the system notebook
[Cat I-III is NOT process                              for SR QU-F2 were not adequately                   calculation (RNP-F/PSA-0018) and described in the MET]                                                   documented. Missing or inadequate                  success criteria notebook (RNP-F/PSA-0075).
documentation was identified for the following     2. A discussion of the sensitivity of accident sequences Revision 0A                                                                                                                                   Page 55 of 79

54002-CALC-01                                                                      Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation             Topic            Status                  Finding/Observation                                         Disposition ID) items: 1) description of the process used to      to operator actions has been added to section 2.7 of account for system successes, 2) discussion of    the quantification calculation (RNP-F/PSA-0077).
human actions that are the key factors in         3. Asymmetries for RNP are discussed in the system causing the accidents to be non-dominant, 3)      notebook calculation (RNP-F/PSA-0018). Asymmetries discussion of asymmetries in the model, 4) list   exist in particular on the AC Power system (and of mutually exclusive results, 5) cutset review   associated systems like DGs and DC power).
process, and 6) discussion of RRW for             4. Mutually-exclusive events are listed in the equipment or operator errors.                     RNP_MTX_12.xls file, and there is a cutset file which lists all the mutually-exclusive combinations in the fault tree. Both of these files are part of the quantification calculation (RNP-F/PSA-0077).
: 5. Cutset review process is described in section 3.3
: 6. No specific standard requirement to discussion review of RRW, but the review of F-V importance is discussed in the quantification calculation and this is considered adequate for RRW as well.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-F2 is considered to be MET at CAT I-III.
QU-F3 (1)         Document significant  Dispositioned Important basic events are documented in          Discussion of top accident sequences for SGTR,
[Cat II-III is    contributors to CDF                 Table 10 of RNP-F/PSA-0077. Important IE          ATWS and LOCA have been added to section 3.15 of MET]              results                             importances are documented Figure 18.             the quantification calculation (RNP-F/PSA-0077).
Detailed accident sequence descriptions are provided in Section 3.15. Some, but not all, of these accident sequences describe the top sequences. For example, no top sequence was described for SGTR, ATWS, or LOCA. Top event tree sequences are discussed in general, but the top cutsets within the event tree sequences are not described. Detailed description of significant accident sequences or functional failure groups could be improved to Revision 0A                                                                                                                                 Page 56 of 79

54002-CALC-01                                                                             Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic               Status                  Finding/Observation                                           Disposition ID) provide more useful detail for all significant contributors to CDF.
QU-F5 (1)          Document limitations      Dispositioned Some software and hardware limitations are         Section 1.3 of RNP-F/PSA-0077, describes limitations.
[Cat I-III is MET] of the quantification                   cited in RNP-F/PSA-0077 Section 2.2, but are       The limitations of the hardware and software have process                                 not described.                                     been discussed in section 2.2. One example of limitations is documented as being the truncation level and a discussion is included as to how better confidence is obtained in the truncation level.
QU-F6 (1)          Document                  Dispositioned Significant' definitions are in Section 1.1 of     Justification for not performing a review of all cutsets in
[Cat I-III is MET] quantitative definitions                RNP-F/PSA-0077. However, the definitions           the top 95% was provided in section 3.3 of the for significant basic                   were not consistently applied. The cutset results  quantification calculation (RNP-F/PSA-0077).
events, cutsets and                     review documented in Section 3.3 of RNP-accident sequences                      F/PSA-0077 was not performed for the top 95%,
as discussed in SR QU-D1. No justification provided for not performing a review of significant cutsets. Also, as discussed in QU-D1, it was not clear that RNP-F/PSA-0077 Table B-4 cutsets included those lower than 95% of the LERF.
SC-A1 (1)          Definition of core        Dispositioned The success criteria definitions are documented    Success criteria definitions documented in calculations
[Cat I-III is NOT  damage                                  in RNP-F/PSA-0075, RNP-F/PSA-0061,                 RNP-F/PSA-0075, RNP-F/PSA-0043, MET]                                                       RNP-F/PSA-0043 and RNP-F/PSA-0049. There           RNP-F/PSA-0061 and RNP-F/PSA-0049 have been is some inconsistency in the success criteria      revised to provide consistent statements throughout.
[Associated with                                           definitions used for MAAP analyses.                Definition for "briefly exceeding 1800 &deg;F" has been SC-A2]                                                                                                        revised to read if the temperature exceeds 1800 &deg;F for a short time (i.e., ~1 min) in RNP-F/PSA-0043 and RNP-F/PSA-0049 based on the MAAP case MLOCA_8-HPI.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the disposition above, SR SC-A1 is considered to be MET at CAT I-III.
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54002-CALC-01                                                                          Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation                Topic            Status                  Finding/Observation                                         Disposition ID)
SC-B3           (1) Definition of Success  Dispositioned RNP Success Criteria calculation                  The success criteria selected for the PSA for the
[Cat I-III is MET]  Criteria                             RNP-F/PSA-0075 and Accident Sequence              LBLOCAs is based on the success criteria specified for calculation RNP-F/PSA-0043 references the         design basis LOCAs in the UFSAR. This includes the
[Associated with                                         supporting thermal hydraulic analyses and/or      use of a single train of equipment for injection and SC-C2]                                                   other applicable analyses. MAAP was used to       recirculation, and injection from the 2 accumulators on confirm the success criteria for large break      the non-broken loops. To provide a sense of assurance LOCA in the PSA, which is based on the            that the UFSAR success criterion was applicable to the success criteria specified for design basis       PSA, the MAAP code was looked at but not credited to LBLOCAs in the UFSAR, according to                evaluate LBLOCA scenarios at RNP. Last paragraph in RNP-F/PSA-0075, Section D.3.1. Although the       Section D.3.1. of RNP-F/PSA-0075 is re-written to LBLOCA MAAP run was described as only a           clarify this.
confirmation of the assumed success criteria, MAAP is not appropriate for analysis of large break LOCAs. MAAP does not conserve momentum and can provide misleading unrealistic results for LBLOCA.
SC-C3 (1)           Document model         Dispositioned Sources of model uncertainty and related          The success criteria calc, RNP-F/PSA-0075, has been
[Cat I-III is NOT   uncertainty                          assumptions are inadequately identified and       reviewed to identify source of model uncertainties and MET]                                                     associated with the development of success        related assumptions. The identified assumptions and criteria.                                         uncertainties have been added in Table 1 of calc RNP-F/PSA-0074, RNP Uncertainty Analysis.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated and addressed accordingly. Therefore, SR SC-C3 is considered to be MET at CAT I-III.
SY-A11          (1) Failures in system     Dispositioned Based on a review of a number of system fault     A review of the systems revealed that an RWST
[Cat I-III is MET]  model affecting                      trees and system notebooks, it appears that       rupture event was not included in the model. Event operability                          most of the pertinent equipment failures are      HTKRWSTFN was developed and added to the model.
included in the system models. However, pipe      An evaluation of pipe rupture events was performed Revision 0A                                                                                                                                    Page 58 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic            Status                   Finding/Observation                                          Disposition ID) ruptures are not included as potential              using the criteria in SR SY-A15, and it was determined equipment failures based on low probability of      that there is adequate justification to discount pipe pipe rupture. This exclusion is acceptable per      failure events for those systems where pipe failures SY-A15 but two additional points need to be         were previously assumed to have a negligible impact covered: 1) pipe failure probabilities for events   on risk. A statement to that effect has been added to that result is the total loss of a system need to   the applicable system notebooks.
be compared with the random probability of system failure not only one component failure probability, 2) if the same fault tree is used for quantifying the IE frequency, the low failure probability/frequency may not be true any longer.
SY-A11 (2)         Failures in system     Dispositioned RWST tank rupture not accounted for in the          Added basic event HTKRWSTFN under gates J150
[Cat I-III is MET] model affecting                      fault tree.                                         and H044 for RWST rupture.
operability SY-A2 (1)          Accuracy of system     Dispositioned Based on a review of RNP-F/PSA-0018                 SR SY-A2 requires the following: COLLECT pertinent
[Cat I-III is MET] analysis compared to                 attachments, (e.g., Attachment A.1, A.9, and        information to ensure that the systems analysis as-built, as-operated                A.5), it is concluded that pertinent data appears   appropriately reflects the as-built and as-operated
[Associated with   plant                                to have been used. Additionally, each system        systems. Examples of such information include system SY-A3]                                                  notebook appears to include a list of all the       P&IDs, one-line diagrams, instrumentation and control pertinent references. However, a direct             drawings, spatial layout drawings, system operating relationship between the information in the         procedures, abnormal operating procedures, system notebook and the source of information       emergency procedures, success criteria calculations, is often not provided (e.g., The source             the final or updated SAR, technical specifications, document for Figure A.1.1 is not provided. Or       training information, system descriptions and related source of information for instrumentation and       design documents, actual system operating control is not stated).                             experience, and interviews with system engineers and operators.
While it could be considered a good practice, there is no standard requirement to provide direct references within the body of the system notebook documentation to support the statements made therein. Additionally, there is no standard requirement to differentiate Revision 0A                                                                                                                                      Page 59 of 79

54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation                Topic           Status                   Finding/Observation                                          Disposition ID) between informational and direct references. All references are included in section 8.0 of each system notebook appendix.
It is acknowledged that traceability of system engineer review comments was deficient. In subsequent system notebook updates, system engineers were assigned NTMs in Passport in which to record their review comments.
As such, no changes are required to meet this SR.
SY-A23 (1)          Develop consistent    Dispositioned The RNP utilizes CAFTA type codes such that a      Added Section 2.8 Basic Events Naming Convention to
[Cat I-III is MET]  model nomenclature                  component failure mode and its associated          RNP-F/PSA-0049.
failure rate are applied consistently in multiple systems. Need documentation of the naming convention for basic event names.
SY-A8           (1) Establish component   Dispositioned The boundaries of the modeled components           The definition of the EDG component boundary is
[Cat I-III is MET]  boundaries,                         appear to match the definitions used to            correct in that it includes the output breaker. The event definitions and data                establish the component failure data. However,     representing the failure of the output breakers there is some evidence that the data boundary      (PCB5217BNN and PCB5227BNN) to close will be set definitions are not consistent with the definition to zero in the model since failures of this component of component boundary. For example, the            are rolled into failures of the EDG. A review was made definition of the component boundary for EDGs      of other component boundaries and it was determined does not include output breaker but the            the boundaries detailed in Attachment 2 of the data definition of the component boundary in the        calculation (RNP-F/PSA-0072) and in the model are data analysis includes output breaker.             appropriate when compared to those set forth in NUREG-CR/6928.
SY-B8           (1) Documentation of      Dispositioned The spatial hazards that may impact multiple       Additional discussion added to Section 3.2 of the
[Cat I-III is NOT   Spatial Hazards                     systems or redundant components in the same        System Notebook Appendices, as appropriate.
MET]                                                    system are not clearly identified.                 RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated and addressed accordingly. Therefore, SR SY-B8 is considered to be MET at CAT I-III.
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54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                           Topic                Status                            Finding                                          Disposition CF-A1-01          REVIEW the conditional failure     Dispositioned RNP-0153 is Attachment 9 of RNP-F/PRA-0094         RNP-0153 (Attachment 7 of P2217-1021-01-03)
(CAT III)         probabilities for fire-induced                   and describes cable failure analyzed for the RNP   has been updated to reflect all RNP information.
circuit failures                                 Fire PRA, RNP-0153. The third bullet of RNP-       Additionally, Reference 5.1 has been updated to
[Associated       and                                              0153, Section 3.0 states, "All cables are assumed  RNP Fire Safe Shutdown Program Database, with CF-A1]       ASSIGN the appropriate industry-                 to be Thermoset. This is consistent with the cable Rev 19.
wide generic values for risk-                    specifications for safety related cable at BNP and significant contributors based on                there are no substantive differences in failure the specific circuit configuration               probabilities for the two cable types for the under consideration                              purposes of this analysis." Reference 5.1 of RNP-0153 is "BNP Fire Safe Shutdown Program Database, Rev. 26." In response to a Peer Review question, it was advised that BNP should be RNP, that cable should be Thermo-plastic rather than Thermo-set and the RNP FSSPMD should be referenced instead.
The determination of meeting the requirements of CF-A1 depends on the information in RNP-0153.
As presently written, RNP-0153 leads to questions of whether it is really applicable to RNP.
Revise and update RNP-0153. Perform a confirmation that the balance of information in RNP-0153 is valid for H.B. Robinson.
CF-A2-01          CHARACTERIZE the uncertainty       Dispositioned No characterization of the uncertainty associated  The basic events associated with hot short associated with the applied                      with the applied conditional circuit failure       probabilities have been assigned an error factor (NOT MET)         conditional failure probability                  probabilities was documented in EPM Report         when the combined cutset is created using the P2217-1021-01-01, Robinson Fire PRA                UNCERT code. Although important for assigned per CF-A1.
Quantification Calculation, as captured in EC      determining the statistical uncertainty of the PRA 90905.                                             cutsets, the criteria needed for the LAR are based on the mean values which are not significantly impacted by the uncertainty analysis.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
Based on the above disposition, SR CF-A2 is considered to be MET at CAT I/II/III.
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54002-CALC-01                                                                               Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                  Status                           Finding                                           Disposition CS-A1-01             IDENTIFY cables whose fire-        Dispositioned Attachment 7 of RNP-F/PSA-0066 (Equipment           All cables identified in Attachment 7 of induced failure could adversely                  Selection) contains a list of equipment which is to RNP-F/PSA-0066 have been routed and added (CAT I/II/III)       affect selected equipment and/or                 be credited in the Fire PRA, but which does not     to FSSPMD with the exception of components credited functions in the Fire PRA               have cables identified and routed. A spot check     that were not installed prior to RNP-F/PSA-0066
[Associated          plant response model.                            review of the cable database (RNP FSSPMD            being completed.
SR: CS-A10,                                                           R21_Read Only QA Record.mdb) reveals that CS-A2]                                                                several of the components listed in Attachment 7 are not included in the cable database. For example, Attachment 7 includes "Fire Tag" CHG-C-INDICATING-LIGHT corresponding to PRA BE JILCHGPCTF; however, CHG-C-INDICATINGLIGHT is not in the cable database, hence BE JILCHGPCTF will never be affected by fires. Others identified by spot check include 480V-52/11A, 480V-52/13B, EDGA-AMMETER, etc. Thus, there are some PRA components which are being credited in the PRA which do not have cable routing incorporated into the cable database; hence, they will never fail.
Review Attachment 7 of RNP-F/PSA-0066, identify components which are not in the cable database, and update the cable database to include all credited PRA components.
CS-A11-01            If assumed cable routing used in   Dispositioned The scope and extent of assumed cable routing       As noted in RNP Change Package RNP-0152 the Fire PRA, IDENTIFY the                       for non-credited components (i.e., components       (Attachment 19 of P2217-1021-01-03), cable (CAT I/II/III)       scope and extent, and PROVIDE                    assumed to fail for all fires) is documented in     toning was used to confirm cable routes. There a basis for the assumed cable                    Table 4 of RNP-F/PSA-0066. Although a basis is      were some instances where cable toning was not routing.                                         not provided for the assumed routing, it is a       possible within a specific compartment generally accepted practice to omit certain         (embedded cable, etc.). In these instances, the systems which do not perform a safety function      cable was assumed to be failed throughout the (but may back up a safety system) and would         entire compartment that it was known to traverse require immense amount of work to manually          through.
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54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                           Topic                 Status                            Finding                                            Disposition
[Associated                                                        trace the cables. The credited components with SR: CS-C3]                                                         assumed routing are contained in the FSSPMD,          As noted in RNP change package RNP-0205, however, there is no documented basis for the         (Attachment 19 of P2217-1021-01-03), the routing, and it is unclear if the scope and extent is assumed cable route data determines the cable-understood. The basis for assumed routing is not      to-fire zone correlation (which is sufficient for documented. Furthermore, to understand, and to        NSCA), but does not determine the cable-to-be able to evaluate the uncertainty associated        raceway-to-fire zone correlation (which is needed with assumed routing some form of                     for PRA). Based on this assumption, any ignition documentation should be assembled describing          source within a given fire zone will impact all the scope and extent of the assumed routing. For      cables with assumed cable routes in the ignition example, are there fire compartments with a           sources fire zone.
significant amount of assumed routing, are there high significance fire compartments, systems, trains or components dominated by cable failure of cables which have an assumed routing, etc.
As part of an overall Task 3 documentation package, describe the scope and extent of assumed routing used in the RNP Fire PRA.
CS-C1-01          DOCUMENT the cable selection       Dispositioned There is no notebook encompassing Task 3              In the process at RNP, Fire Protection/NSCA and location methodology applied                 (Cable Selection) making review and update            develops and maintains the cable selection and (NOT MET)         in the Fire PRA in a manner that                 difficult. There are numerous change packages,        circuit analysis data. This data is then referenced facilitates Fire PRA applications,               and a database (FSSPMD) which is a repository         as inputs to the Component Selection and upgrades, and peer review.                       for the cable routing information; however, there     Quantification FPRA calculations. This process is no document explaining what tasks were             and associated results are easily reviewable, has performed, which procedures or guidelines were        been peer reviewed multiple times for our other employed, and in which document the analysis is       sites and found to be acceptable. There is no contained.                                            requirement to have a separate PRA notebook.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
With no change being made, RNP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be Revision 0A                                                                                                                                       Page 63 of 79

54002-CALC-01                                                                            Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                           Topic                 Status                            Finding                                           Disposition sufficiently resolved for SR CS-C1 to be assessed as CAT I/II/III is MET.
ES-A2-01          REVIEW power supply, interlock     Dispositioned Power supplies, support systems, interlock           A detailed review of the modeling of associated circuits, instrumentation, and                   circuits and instrumentation are included in the     circuits for all components functional states in the (NOT MET)         support system dependencies                      internal events PRA model. For fire PRA specific     fire PRA CAFTA model was performed. The details of the review are provided in Attachment 2 and IDENTIFY additional                          logic, examples indicate that not all support
[Associated                                                                                                             of Hughes Associates Inc. (HAI) equipment whose fire-induced                     system dependencies were completely                  1RCS04042.414.031-002, and revisions to the SR: ES-B4,        failure, including spurious                      considered. The area of concern is the lack of       fire PRA CAFTA model, the RR file table BE, and CS-A2]            actuation, could adversely affect                modeling of instrumentation power dependency         Component Selection Database (CSDB) table any of the equipment identified                  (self-identified by the utility). For example, see   PRA_SSEL as a result of this detailed review are per SR ES-A1.                                    gate HRAPORV-455C "PORV FAILS TO CLOSE               listed in 1RCS04042.414.031-001, Attachments DUE TO HRA", one of its inputs is an AND gate        3, 4, and 5, respectively. Revision 4 of the RNP fire PRA calculation for component selection for "INDICATIONS THAT PORV IS OPEN FAIL."
(RNP-F/PSA-0066) incorporates all of the For the indicators modeled here, if there is a loss  changes required to the fire PRA CAFTA model of power, the indicators will scale low, potentially as a result of the detailed review of the preventing the operators from taking necessary       associated circuits.
actions. For the indication discussed in the example above, a loss of power will cause the        Review credited components to ensure that indicators to scale low, potentially preventing the  power supplies, support systems, interlock circuits, and instrumentation dependencies are operators from taking necessary actions, and adequately captured.
thus could adversely impact fire risk.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
With the above described changes incorporated, this SR is considered to be MET at CAT I/II/III.
ES-A3-01          INCLUDE equipment whose fire-      Dispositioned The "RCP Seal No. 1 Leak-Off" and "SI test line"     RNP-0148, Multiple Spurious Operation Expert induced failure, not including                   are both ISLOCA paths that were screened from        Panel Report, did consider spurious isolation of (NOT MET)         spurious operation, contributes to               the Internal Events analysis; however these          the RCP Seal No. 1 leakoff valves. As noted in or causes                                        screened paths are not explicitly dispositioned in the report, 125V DC power will be removed within (a) fire-induced initiating events               the fire PRA component selection calculation.
treated in the Fire Safe Shutdown                These items were discussed with the host utility
                  / Appendix R analysis                            that provided the following input on the two Revision 0A                                                                                                                                      Page 64 of 79

54002-CALC-01                                                                               Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                  Status                            Finding                                            Disposition (b) Internal Events PRA initiators               ISLOCA paths mentioned above: The seal leak-         ten minutes as noted in RNP Dedicated as identified using the IE                       off is normally open and seal leak-off is part of     Shutdown Procedure (DSP-002).
requirements in Part 2 (including                the normal cooling path. Failure in either direction any gradations across capability                                                                       As noted in Attachment 3 of the RNP Component is non-consequential" and "The test line has 3+
categories in that standard) as                                                                        Selection Calculation, RCP Seal No. 1 Leak-Off modified per 4-2.5, or                           locked closed manual valves." Based on this feedback, it is judged that this is a documentation   is not consequential. Leakoff is not required to (c) unique fire-induced initiating events not addressed or                          issue with no impact on the analysis. This finding    maintain adequate seal cooling in the PRA otherwise screened from the                      is related to the consideration of previously         model. Furthermore, seal leakoff is not modeled above two analyses if SR IE-C4                   screened ISLOCA pathways from the Internal            in the Fire PRA.
in Part 2 cannot be met.                         Events PRA for inclusion in the Fire PRA.
Include, in the component selection calculation, a RESOLUTION OF CAPABILITY CATEGORY discussion of screened paths and why or why not CLASSIFICATION:
they contain components that warrant inclusion in the fire PRA.                                         Based on the above disposition, this SR is considered to be MET at CAT I/II/III.
FQ-B1-01             PERFORM the quantification in      Dispositioned In review of the fire PRA documentation there no      RNP-F/PSA-0077 documented proof of (CAT I/II/III)       accordance with HLR-QU-B and                     convergence study was performed in support of         convergence for CDF and LERF for the internal its SRs in Part 2                                the selection of truncation level for quantification. events PRA. As the fire PRA model is largely and                                                                                                    dependent on the internal events PRA, it can be DEVELOP a defined basis to                                                                             concluded that there is convergence for CDF and support the claim of non-                                                                              LERF in the fire PRA applicability of any of the requirements under HLR-QU-B in                                                                         The model was quantified based on a CDF Part 2.                                                                                                quantification of 1E-12. This is about seven orders of magnitude below the final CDF. This is generally accepted as a good bounding truncation level FQ-F1-01             DOCUMENT the CDF and LERF          Dispositioned The contents of the elements of applicable SRs        The RNP Fire PRA was developed using the (CAT I/II/III)       analyses in accordance with                      of Part 2 were addressed in the FQ and                Internal Events PRA. The Internal Events PRA is HLR-QU-F and HLR-LE-G and                        associated documents; however, no explicit            aligned with RG 1.200 and was peer reviewed.
their SRs in Part 2 with the                     connections were established in the documents         Therefore, the back-references associated with following clarifications:                        to associate with the "back-references"               requirements LE-G2, LE-G4 and LE-G5 are (a) SRs QU-F2 and QU-F3 of                       requirements LE-G2, LE-G4 and LE-G5.                  considered to be met.
Part 2 are to be met including Revision 0A                                                                                                                                          Page 65 of 79

54002-CALC-01                                                                                Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                   Status                             Finding                                       Disposition identification of which fire scenarios and which physical analysis units (consistent with the level of resolution of the Fire PRA such as fire area or fire compartment) are significant contributors (as defined in Part 1);
(b) SR QU-F4 of Part 2 is to be met consistent with 4-2.13 (c) SRs LE-G2 (uncertainty discussion) and LE-G4 of Part 2 are to be met consistent with 4-2.13, and DEVELOP a defined basis to support the claim of non-applicability of any of the requirements under these sections in Part 2.
FSS-B1-01            DEFINE and JUSTIFY the              Dispositioned The conditions and timing that lead to Main        Section 5.9.2 of P2217-1021-01-03 provides a (CAT I/II/III)       conditions that are assumed to                    Control Room (MCR) abandonment due to              discussion of Main Control Room abandonment cause MCR abandonment and/or                      environmental conditions and reliance on ex-       due to environment conditions.
[Associated          reliance on ex-control room                       control room actions are described in main SR: FSS-A6,          operator actions including remote                 control room analysis report (Report               The probability associated with main control room FSS-B2]              and/or alternate shutdown                         0004-0042-412-002).                                abandonment due to a loss of habitability has actions.                                                                                             been incorporated into the quantification of the Abandonment due to equipment damage and to         Fire PRA.
loss of habitability is based on the guidelines provided in Section 11.5.2.11 of NUREG/CR-6850. The analysis method described in NUREG/CR-6850 Appendix L is used to assess scenarios in the MCB.
This SR is considered met. A new Finding (FSS-B1-01) has been assigned to this SR because all conditions requiring reliance on Revision 0A                                                                                                                                      Page 66 of 79

54002-CALC-01                                                                                Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                            Topic                   Status                            Finding                                           Disposition remote/alternate shutdown are not identified.
Fires in the MCR or other location in the plant that may lead to loss of control room functions such that use of remote/alternate shutdown capability is required are not characterized and evaluated. Such scenarios for MCR habitability are identified and analyzed.
FSS-C2-01          CHARACTERIZE ignition source         Dispositioned Section 5.6.2 of the Quantification Calculation       Section 5.6.2 of the Fire PRA Quantification intensity using a realistic time-                  (EPM Report No. P2217-1021-01-01, Rev. 1,             Calculation discusses the growth profile used.
(CAT II/III)       dependent fire growth profile (i.e.,               dated February 2013 included in EC 90905) and         This can also be seen in the RNP_EVAL_Rev2 a time-dependent heat release                      in particular the third bullet under Figure 7 on      spreadsheet used during the quantification rate) for significant contributors                 Page 51 of 113 indicates that "The HGL                process. This spreadsheet includes a cable tray as appropriate to the ignition                     threshold assumes a constant HRR up to HGL            propagation model (HGL_Time worksheet).
source.                                            formation." The basis for using the time versus HGL based on a constant HRR and then                  Furthermore, Calculation NED-M/MECH-1009 compare it to a curve based on a variable HRR         also provides a time to damage based on fire has not been provided. At the top of Page 52 of       growth.
113 in the Quantification Calculation it states that, "The net effect of these uncertainties is generally a conservative time to HGL. For ignition sources that are high risk, more detailed fire modeling may be pursued on a case by case basis."
The intent of the Standard is that conservative =
Category I, and that realistic = Category II/III, therefore this SR is evaluated at Category I. An example of a more realistic analysis is Hughes Report Number: 0004-0042-000-001 for Fire Compartment 20.
Finding FSS-C2-01 has not been resolved. For each scenario, the ignition source intensity is characterized using a time-dependent heat release rate, consistent with the Category II/III requirement for this SR; however, the total heat Revision 0A                                                                                                                                         Page 67 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                         Topic                  Status                            Finding                                         Disposition release required to cause hot gas layer formation is based on a fire that is initiated at full peak intensity, consistent with the Category I requirement for this SR. Suggestion FSS-C2-02 has been prepared to suggest further justification and validation of the methodology used to determine the total heat release required to cause hot gas layer formation.
FSS-C4-01       If a severity factor is credited in Dispositioned The severity factors, as described in Section      Section 5.5.4.5 of the Quantification Calculation the analysis, ENSURE that                         5.5.4.5 of the Quantification Calculation (EPM     (EPM Report No. P2217-1021-01-03) has been (CAT II)                                                          Report No. P2217-1021-01-01, dated                 updated to provide more detail as to why specific
: a. the severity factor remains                 February 2013 included in EC 90905) are            severity factors are applied in the MakeSScen independent of other                        apparently evaluated using the calculation logic   module.
quantification factors                      in Attachment 20 of the Quantification Calculation. This is not explained in a clear and
: b. the severity factor reflects                concise way in Section 5.5.4.5, nor is any the fire event set used to                  reference to Attachment 20 found in that section.
estimate fire frequency                     The general reference to module "MakeSScen" and the fact that its contents are listed in
: c. the severity factor reflects Attachment 20 in the introductory paragraph of the conditions and Section 5.5.4 is not considered to be sufficient.
assumptions of the specific Further, the descriptions in Section 5.5.4.5 fire scenarios under (pages 41-42 of 113) of how the Severity Factor analysis, and determination is done do not appear to exactly
: d. a technical basis supporting                match with the programming logic found on the severity factors                       pages 7-12 of Attachment 20. This needs to be determination is provided.                  clarified and clearly documented to support future use, update, and peer review of the Severity Factor calculations.
In addition, the use of generic fire modeling data and severity factors for different ignition source Bins from NUREG/CR-6850 without considering mode of exposure and position (i.e., not just Revision 0A                                                                                                                                    Page 68 of 79

54002-CALC-01                                                                              Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                              Topic                 Status                           Finding                                              Disposition distance) of the targets relative to the fire source may not fully constitute "explicit consideration" in quantifying the severity factor such that it reflects the conditions and assumptions of the specific fire scenario under analysis.
FSS-C7-01            If multiple suppression paths are  Dispositioned Credit is given for both automatic sprinkler          No credit has been applied for manual actuation credited, EVALUATE and                           suppression and manual firefighting, but mutual       or recovery of fixed suppression systems. The (NOT MET)            PROPERLY MODEL                                   dependency on the common water supply system          common water supply at RNP is sufficient to dependencies among the                           has not been evaluated or properly modeled.           provide water to both the automatic suppression
[Associated          credited paths including                                                                               system as well as the manual firefighting. The SR: FSS-D6]          dependencies associated with                                                                           reliability/unavailability of the fire pump and recovery of a failed fire                                                                              associated sprinkler has already been accounted suppression system, if such                                                                            for in the non-suppression probability.
recovery is credited.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
Based on the above disposition, SR FSS-C7 is considered to be MET at CAT I/II/III.
FSS-D2-01            USE fire models that have          Dispositioned This finding recommends that the HGL                  Attachment J of the LAR provides a discussion sufficient capability to model the               calculation (RNP-M/MECH-1826) be subjected to         on the software used during the development of (CAT I/II/III)       conditions of interest and only                  validation and verification in order to establish its the RNP HGL Calculation (RNP-M/MECH-1826).
within known limits of                           technical basis and known limits of applicability.
[Associated          applicability.
SR: FSS-D6]
FSS-D7-01            In crediting fire detection and    Dispositioned There is a failure to meet the Category I             Currently, system performance is monitored and (NOT MET)            suppression systems, USE                         requirement of having systems installed and           maintained at a high level as part of the System generic estimates of total system                maintained in accordance with applicable codes        Health Reporting and System Notebook unavailability provided that                     and standards. The Main Turbine Lube Oil              processes. Outlier behavior with respect to (a) the credited system is                       Deluge system must be replaced to account for         system availability would be evident to the installed and maintained in                      system deficiencies identified in NCR-425437          system engineer and plant management through accordance with applicable                       where a simultaneous actuation of the Turbine         the health data (available for the previous codes and standards                              Lube Oil suppression system, along with the Revision 0A                                                                                                                                             Page 69 of 79

54002-CALC-01                                                                            Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                          Topic                  Status                           Finding                                           Disposition (b) the credited system is in a                   mezzanine and ground level sprinkler systems,       12 months), which indicates overall Excellent fully operable state during plant                 could place a higher system demand on the           (Green) performance.
operation, and                                    water supply than can be provided by a single fire (c) the system has not                            pump.                                               Furthermore, during the periods when key fire experienced outlier behavior                                                                          protection systems are unavailable due to testing relative to system unavailability.                This was not identified, nor is a comparison        and maintenance, compensatory actions are provided in the Fire PRA of all installed detection taken such that the risk associated with the and suppression systems vs. the corresponding       system being unavailable does not increase.
Code Compliance calculation.
RESOLUTION OF CAPABILITY CATEGORY Finding FSS-D7-01 has not been resolved.            CLASSIFICATION:
Because evidence is not provided to support that    Using plant-specific information to quantify total credited detection/suppressions systems are         unavailability factors is a CAT III requirement and installed and maintained in accordance with         was not done.
applicable codes and standards. System health report for period Q2-2013 for systems               With no change being made, RNP considers the 6185/6181/6175/6195/6205/6180 notes that age,       risk results from the Fire PRA to be creditable for obsolescence and replacement part procurement       the NFPA 805 application and this finding to be is an issue. This system health report also notes   sufficiently resolved for SR FSS-D7 to be that "There are LTAMs budgeted for 2014 and         assessed as CAT II is MET.
2015 which study and replace the detection, CO2, and Halon Systems." This report suggests that some of the fire protection systems at RNP may be experiencing outlier behavior relative to system unavailability and may not be in a fully operable state during plant operation.
Consequently, this SR is still considered to be not met.
FSS-D7-03       In crediting fire detection and     Dispositioned Evidence needs to be provided to support that       Currently, system performance is monitored and (NOT MET)       suppression systems, USE                          credited detection/suppressions systems are         maintained at a high level as part of the System generic estimates of total system                 installed and maintained in accordance with         Health Reporting and System Notebook unavailability provided that (a)                                                                      processes. Outlier behavior with respect to applicable codes and standards. System health the credited system is installed                                                                      system availability would be evident to the and maintained in accordance                      report for period Q2-2013 for systems               system engineer and plant management through with applicable codes and                         6185/6181/6175/6195/6205/6180 notes that age,       the health data (available for the previous 12 standards, and (b) the credited                   obsolescence and replacement part procurement       months), which indicates overall Excellent system is in a fully operable state               is an issue for multiple fire protection systems.   (Green) performance.
during plant operation.                           This system health report also notes that "There Revision 0A                                                                                                                                     Page 70 of 79

54002-CALC-01                                                                          Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                 Status                        Finding                                         Disposition are LTAMs budgeted for 2014 and 2015 which        Furthermore, during the periods when key fire study and replace the detection, CO2, and Halon   protection systems are unavailable due to testing Systems." This report suggests that some of the   and maintenance, compensatory actions are taken such that the risk associated with the fire protection systems at RNP may be system being unavailable does not increase.
experiencing outlier behavior relative to system unavailability and may not be in a fully operable RESOLUTION OF CAPABILITY CATEGORY state during plant operation.                     CLASSIFICATION:
Using plant-specific information to quantify total unavailability factors is a CAT III requirement and was not done.
With no change being made, RNP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-D7 to be assessed as CAT II is MET.
FSS-E1-01           For any fire modeling parameters   Open       Section 4.3 of Calculation No. P2217-2100-00,     Section 4.3 of Calculation P2217-2100-01-01 will (CAT I/II/II)       not covered by the requirements               Fire Scenario Data, RNP-F/PSA-0079, dated         be updated appropriately at a later date. This is a of HLR-FSS-C or HLR-FSS-D,                    January 2013 contains information about fire      documentation issue that will not impact fire USE plant-specific parameter modeling parameters that were used. However,      scenario development or quantification.
estimates for fire modeling if available, or use generic                     Section 4.4 through 4.7 should be completed information modified as                       because they are missing information about other discussed in SR FSS-E2; USE                   relevant fire modeling parameters.
generic information for the remaining parameter estimates.                Add relevant information to the report.
Finding FSS-E1-01 has not been resolved.
Sections 4.3 through 4.7 still make reference to databases for the parameters used in the fire modeling. These parameters should be added to the report.
FSS-E3-01           PROVIDE a mean value of, and          Open    No statistical representation of uncertainty      RNP used the HRRs and applied them using the (CAT I)             statistical representation of, the            intervals (e.g., NUREG/CR-6850 Table E-1 or       guidance found in NUREG/CR-6850. As uncertainty intervals for the Revision 0A                                                                                                                                   Page 71 of 79

54002-CALC-01                                                                          Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                          Topic                 Status                           Finding                                          Disposition parameters used for modeling                    G-1 for HRR, Tables E-2 through E-9 for severity  NUREG/CR-6850 is the consensus methodology, the significant fire scenarios.                 factor) is documented for the mean values of      a detailed uncertainty analysis on these parameter estimates used for fire modeling the    parameters is not needed and does not add to significant fire scenarios.                       the credibility of the results. The majority of applied values are based on the 98th and 75th percentile fires from NUREG/CR-6850, and the ZOIs are applied conservatively. It is not believed that reducing these values would allow the use of reduced impacts for the applications being pursued.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
Although no change has yet been made that would improve the Capability Category assessments, RNP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because documenting the statistical representation of uncertainty intervals will not change the quantified risk metrics.
FSS-E4-01         PROVIDE a characterization of     Dispositioned There is some assumed cable routing applied       As noted in RNP Change Package RNP-0152 (NOT MET)         the uncertainties associated with               directly in FSSPMD. These routings are identified (Attachment 19 of P2217-1021-01-03), cable cases where cable routing has                   in RNP-0152. Components with unknown routing      toning was used to confirm cable routes. There
[Associated       been assumed based on SRs                                                                         were some instances where cable toning was not are otherwise assumed to fail in the PRA model.
SR: FSS-A3]       CS-A10 and/or CS-A11.                                                                             possible within a specific compartment However, there is no discussion that              (embedded cable, etc.). In these instances, the characterizes the uncertainties associated with   cable was assumed to be failed throughout the cases where cable routing has been assumed. A     entire compartment that it was known to traverse more detailed characterization of the             through.
uncertainties associated with cases where cable routing has been assumed is needed for SR         As noted in RNP change package RNP-0205 (Attachment 19 of P2217-1021-01-03), the FSS-E4 to be met.
assumed cable route data determines the cable-to-fire zone correlation (which is sufficient for NSCA), but does not determine the cable-to_raceway-to-fire zone correlation (which Revision 0A                                                                                                                                    Page 72 of 79

54002-CALC-01                                                                               Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                              Topic                  Status                           Finding                                          Disposition Provide a detailed uncertainty characterization    is needed for PRA). Based on this assumption, discussion in appropriate reports related to       any ignition source within a given fire zone will circuits with assumed routing.                     impact all cables with assumed cable routes in the ignition sources fire zone.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
Based on the above disposition, SR FSS-EF-04 is considered to be MET at CAT I/II/III.
FSS-F3-02            In crediting fire detection and     Dispositioned Section 2.4 of the Structural Steel report         The unavailability of the Turbine Lube Oil Deluge (CAT I/II/III)       suppression systems, USE                          (P2217-2300-01-03) states that "It is assumed      system has been updated to 0.05 in Rev. 3 of generic estimates of total system                 that an unavailability value of 0.01 will be       P2217-1021-01-03 (Fire PRA Quantification
[Associated          unavailability provided that (a) bounding and conservative for the deluge           Calculation). This change was made based on SR: FSS-D7]          the credited system is installed and maintained in accordance                      sprinkler system." What is the basis for this? Has engineering judgment.
with applicable codes and                         this value been confirmed against the plant-standards, and (b) the credited                   specific experience for availability of the system is in a fully operable state               detection/suppression systems?
during plant operation.
FSS-G1-01            APPLY all the supporting            Dispositioned Fire modeling performed in support of the Multi-   A detailed multi-compartment analysis was (NOT MET)            requirements listed in SRs FSS-                   Compartment Analysis (MCA) failed to consider      performed and is documented in C1 through FSS-C8 for fire                        the acute effect of hot gas flow through openings  RNP-F/PSA-0089. The Fire PRA Quantification
[Associated modeling of single physical                       and ducts on local potential targets. In addition, Calculation only updated results based on new SR: FSS-G2, FSS-G3]              analysis units to the modeling of                 the consideration and evaluation of additional     CCDPs and a new HGL Calculation. The open multi-compartment fire scenarios.                 aspects of multi-compartment fire scenarios was    Turbine Building was discussed in the HGL not documented and/or done, i.e., the fact that    Calculation (RNP-M/MECH-1826). The detailed the Turbine Building has no exterior walls was     review of hot gas flow through openings has not not identified and the impact of this evaluated. been performed. The impact of hot gas flow Treatment of MCA in only 7 pages of the            through openings and ducts on local targets is Quantification Calculation does not appear to be   expected to be minimal.
adequate. Furthermore, in their evaluation of compartments subject to a hot gas layer, RNP       RESOLUTION OF CAPABILITY CATEGORY used a criterion of 1E-07 for total ignition       CLASSIFICATION:
Based on the above disposition, SR FSS-G1-01 frequency. Based on this, RNP excluded all but is considered to be MET at CAT I/II/III.
two compartments. A review of the table that was Revision 0A                                                                                                                                        Page 73 of 79

54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                        Topic                 Status                            Finding                                           Disposition subject to the review showed there were two additional compartments that met the criterion for inclusion. This is an error rate of 100%.
FSS-G6-02       QUANTIFY the risk                 Dispositioned Table B-5 of this report defines lists fourteen (14) Section 5.10 of P2217-1021-01-03 provides a (CAT I)         contribution of any selected                    unscreened multi-compartment fire scenarios.         discussion of the multi-compartment scenarios multi-compartment fire scenarios                Attachment E of the same report offers risk          quantified in the Fire PRA. There are five in a manner consistent with the associated with these scenario. The                  scenarios that do not screen currently based on FQ requirements.
Quantification Report (P2217-1021-01-02)             an updated HGL Calculation Section 5.10 states that "Based on updated Hot       (RNP-M/MECH-1826, Rev. 0). These scenarios Gas Layer frequencies, all of the fire               have been quantified in a manner consistent with compartments previously analyzed in the Multi-       the FQ requirements.
Compartment Analysis (Reference 3.14), have now screen." Reference 3.14 is                       RESOLUTION OF CAPABILITY CATEGORY RNP-F/PSA-0089 Rev 0 and it does not support         CLASSIFICATION:
Based on the above disposition, SR FSS-G6 is the statement that all MC fire is screened.
considered to be Met at CAT II/III.
Finding FSS-G6-01 has been resolved. Hughes Calculation No. 1RCS04042.414.031, Revision 0 (Multi-Compartment Analysis for the Fire Probabilistic Risk Assessment at Robinson Nuclear Plant), provides a detailed multi-compartment analysis. This analysis has been included in Section 5.10 of the RNP Fire PRA Quantification Calculation. Based on updated Hot Gas Layer data, the scenarios identified in the Hughes calculation screen from further analysis.
FSS-H2-01       DOCUMENT a basis for target       Dispositioned The fire modeling in Hughes Report No.               Plant modifications in addition to intumastic cable (CAT I)         damage mechanisms and                           0004-0042-000-001 for Fire Compartment 20            coating are incorporated into the Fire PRA thresholds used in the analysis,                credits fire coating. Category II of this SR         Calculation (EPM Report No.
including references for any                                                                         P2217-1021-01-03). These plant modifications requires documentation of the references for any plant-specific or target-specific                                                                    are discussed in Attachment S of the LAR. The performance criteria applied in                 plant-specific or target-specific performance        ten minute time until cable damage (per the analysis.                                   criteria applied in the analysis, and a basis for Revision 0A                                                                                                                                    Page 74 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                         Topic                   Status                           Finding                                          Disposition target damage mechanisms and thresholds used       NUREG/CR-6850) is achieved via plant in the analysis, which has not been provided.      modifications discussed in Attachment S.
Finding FSS-H2-01 has not been resolved.
Plant-specific documentation should be provided    RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
for the performance criteria used to evaluate Based on the above disposition, SR FSS-H2 is damage to coated cables and cable trays with       considered to be Met at CAT II/III.
solid bottoms.
HRA-C1-01         For each selected fire scenario,    Dispositioned The fire-specific operator actions are evaluated   All fire response actions have been updated such QUANTIFY the HEPs for all                         using only the Cause-Based Decision Tree           that they are quantified using the CBDTM and (CAT II)          HFEs and ACOUNT FOR                               method (CBDTM). The HRAs from the Internal         HCR/ORE combination method similar to Internal relevant fire-related effects using               Events methods used both CBDTM and                 Events operator actions. P2217-1022-01-03
[Associated       detailed analyses for significant                 HCR/ORE to address the cognitive risk of the       documents the Fire HRA.
with FQ-C1]       HFEs and conservative estimates                   operator actions. The HCR/ORE method is (e.g., screening values) for non-                 generally the dominant risk value for actions with significant HFEs, in accordance                   short system time windows or very long median with the SRs for HLR-HR-G in                      response times. Events 0FIREOMA01 and Part 2 set forth under at least                   0FIREOMA02 are events with short time windows Capability Category II, with the                  where the HCR/ORE method would be the following clarification:                          dominant cognitive risk. The cognitive risk is underestimated by an order of magnitude.
: a. Attention is to be given to how the fire situation alters                The HRA needs to be revised to better address any previous assessments in                  the cognitive risk portion for each HEP.
non-fire analyses as to the influencing factors and the timing considerations covered in SRs HR-G3, HR-G4, and HR-G5 in Part 2 and Revision 0A                                                                                                                                   Page 75 of 79

54002-CALC-01                                                                                   Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                    Status                              Finding                                          Disposition
: b. Develop a defined basis to support the claim of non-applicability of any of the requirements under HLR-HR-G in Part 2.
IGN-A7-01            USE a plant-wide consistent          Dispositioned In general the SR was met. There are areas such      Transient fire scenarios have been postulated in methodology based on                               as the transient ignition sources related            each fire compartment as noted in Hughes (CAT I/II/III)       parameters that are expected to                    identification are still required, some of these is  Calculation No. 1RCS04042.414.031-002.
influence the likelihood of ignition               already self-identified as part of the "lead review" to apportion high-level ignition                   activities as indicated in the -0067 notebook. For frequencies (e.g., plant-wide                      instance, fire zone 22 and others are identified in values) to estimate physical                       the "lead reviews" as additional effort is needed.
analysis unit or ignition source level frequencies.                                 Address the transient ignition sources for application fire zones. Providing a complete and through ignition sources identification is important.
IGN-A9-01            POSTULATE the possibility of         Dispositioned RNP did postulate transient combustibles for all     Transient fire ignition frequency has been transient combustible fires for all                physical analysis units except for one, FC490.       assigned to FC490. Transient Influence Factors (CAT I/II/III)       physical analysis units regardless                 This physical analysis unit is the deepwell pump     of Low(1), Low(1), and Low(1) replaced factors of the administrative restrictions.                D enclosure. RNP needs to provide justification      No(0), No(0), and No(0) for maintenance, for why transient combustibles are not postulated    occupancy, and storage, since entry to FC490, for FC490.                                           which requires using a crane to remove the concrete enclosure, is not precluded.
PP-B6 (01)           ENSURE                               Dispositioned Underground cabling was not addressed in             The cable routing database system FSSPMD RNP-F/PSA-0067. The RNP project team noted a         was reviewed determining which manholes had (CAT I/II/III)       a. that collectively, the defined                comment was included in this calculation that a      cables within the scope of the Fire PRA physical analysis units                       Yard fire zone should be included in the next      associated with them. The manholes where encompass all locations                       revision of the report.                              cables were identified have been included as plant partitioning elements (i.e., Physical Analysis Revision 0A                                                                                                                                            Page 76 of 79

54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                         Topic                  Status                           Finding                                         Disposition within the global analysis                                                                   Units) in the Fire PRAs. The following manholes boundary                                                                                     have been included as plant partitioning and                                                                                              elements: MH M-34, MH M-35, MH-1, and MH-34. The cables that were identified as routed
: b. that defined physical                                                                        through MH-35, MH-36 and MH have been analysis units do not overlap                                                                included in the Intake Structure Fire Compartment, FC290. The cable loading for the manholes have been assessed following the same approach documented in the RNP combustible loading calc. The approach consists of multiplying the factor of 5,515 BTU/ln-ft to each linear ft of cable. Under this approach, a total of 496350 BTUs have been estimated for each manhole. This assumes nine 10 long exposed cables per manhole. Nine cables is the average number of cables identified in the manholes.
PRM-B11-01        MODEL all operator actions and    Dispositioned Main control room abandonment is discussed in    Fire response and Main Control Room operator influences in                          Section 7.1.4 of Calculation P2217-1022-01-01,   Abandonment HFEs have not been incorporated (NOT MET)         accordance with the HRA                         "Fire Human Reliability Analysis" and Table 7.1  in the combined Internal Events/Fire PRA. The element of this Standard.
lists the human failure events (HFEs) associated Main Control Room Abandonment HFEs from
[Associated                                                       with main control room abandonment. However,     Hughes Calculation No.
with HRA-B2                                                       these HFEs have not yet been incorporated into   RSC-CALKNX-2013-0301 has been incorporated and PRM-B2]                                                       the RNP fire PRA model. Without the HFEs for     into the recovery rule files used during the main control room abandonment, the RNP fire      quantification process as appropriate based on PRA is incomplete.                               the fire compartment being quantified.
Incorporate the main control room abandonment    RESOLUTION OF CAPABILITY CATEGORY HFEs into the fire PRA model.                    CLASSIFICATION:
Based on the abovementioned details, this SR is considered to be MET at CAT I/II/III.
Revision 0A                                                                                                                                  Page 77 of 79

54002-CALC-01                                                                                  Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                   Status                             Finding                                        Disposition PRM-B15-01           MODEL any new accident              Dispositioned Quantification of the RNP fire PRA model            Following additional refinements to the model, progressions beyond the onset of                  indicates that the fire-related LERF is about       LERF is more in line with typical results.
(CAT I/II/III)       core damage identified per                        90 percent of the fire-related CDF. This result is PRM-B13 to determine the                          exceptionally high and has been self-identified by fire-induced LERF in accordance                   the utility. The result is produced by MSO events with HLR-LE-A, HLR-LE-B,                          in the switchgear and cable spreading rooms that HLR-LE-C, and HLR-LE-D and                        cause a core damaging accident sequence and their SRs in Part 2 with the                      also result in the spurious opening of several following clarifications:                         containment isolation valves. Fire-related LERF (a) All the SRs under HLR-LE-A,                   for RNP is an unusually high proportion of the HLR-LE-B, HLR-LE-C, and HLR-                      fire-related CDF. Typically, fire-related LERF is LE-D in Part 2 are to be                          10 - 25 percent of the fire-related CDF. At RNP, addressed in the context of fire                  fire-related LERF is about 90 percent of the fire-scenarios including effects on                    related CDF.
system operability / functionality, operator actions, accident                        Investigate ways to remove conservatisms from progression, and possible                         the RNP fire PRA model, particularly for the containment failures accounting                   MSOs in the switchgear and cable spreading for fire damage to equipment and                  rooms.
associated cabling.
(b) LE-C2 and LE-C6 in Part 2 are to be met in a manner consistent with 4-2.10.
(c) LE-C6 in Part 2 is to be met in a manner consistent with PRM-B9 above.
(d) LE-C8 in Part 2 is to be met in a manner consistent with PRM-B6 above.
and DEVELOP a defined basis to support the claim of nonapplicability of any of these requirements in Part 2.
PRM-C1-02            DOCUMENT the Fire PRA plant         Dispositioned Documentation of development of the plant           The development of the plant response model is response model in a manner                        response model is spread over several               documented in Attachment 9 of RNP-/PSA-0066 (CAT I/II/III)       consistent with HLR-IE-D,                         calculations and documents. The primary             via a model change log. RNP-F/PSA-0066 also HLRAS-C, HLR-SC-C, HLR-SY-                        documents are RNP-F/PSA-0066 (component Revision 0A                                                                                                                                         Page 78 of 79

54002-CALC-01                                                                                 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                              Topic                  Status                             Finding                                         Disposition C, and HLR-DA-E and their SRs                     selection) and P2217-1021-01-01 (quantification). discusses assumptions made during the in Part 2 as well as 4-2.10 with                  Thus, key elements of the PRM development           development of the model.
the following clarifications:                     (e.g., assumptions that are made while (a) HLR-IE-D in Part 2 is to be                   developing the model) being dispersed and met in a manner consistent with                   difficult to assimilate and comprehend. The that required under HLR-IGN-B                     dispersion of pertinent information about of this Standard.                                 development of the PRM makes understanding of (b) Document any defined bases                    the development difficult for reviewers and for to support the claim of                           utility staff who will make future modifications to nonapplicability of any of the                    the RNP fire PRA model and documentation.
referenced requirements in Part 2 beyond that already covered by                    Develop a PRM-specific calculation or notebook the clarifications in this section.               which combines the pertinent portions of PRM documentation from existing documents or calculations.
UNC-A2               INCLUDE the treatment of            Dispositioned To comply with Section 2 applicable SRs intent      Section 7.0 of the RNP Fire PRA Quantification uncertainties, including their                    for UNC, a final documentation of importance        Calculation, provides an analysis of uncertainty (CAT I/II/III)       documentation, as called out in                   rankings considering various sensitivity studies    regarding the quantification of the RNP Fire PRA.
SRs PRM-A4, FQ-F1, IGN-A10,                       along with applicable sequences is suggested to     Importance rankings are also discussed in this IGN-B5, FSS-E3, FSS-E4, FSS-                      be included in the UNC discussion in an             section. Additionally, Section 9.0 provides a H5, FSS-H9, and CF-A2 and that                    independent document.                               discussion of sensitivities that were evaluated.
required by performing Part 2 referenced requirements throughout this Standard.
Revision 0A                                                                                                                                          Page 79 of 79

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 7 Pages (including this cover page)
E-C AUGMENT PLAN TABLES FOR THE SECOND TEN-YEAR INTERVAL

H. B. Robinson Steam Electric Plant, Unit No. 2 IWE/IWL Second Ten-Year Interval Inservice Inspection Plan 11.0 Attachment E-C AUGMENT Plan Tables This Attachment contains the E-C AUGMENT Plan Tables for the Second Ten-Year Interval RNP-PM-007                                                     Revision 5                Page 9 of 12

H. B. ROBINSON UNIT 2                    CATEGORY - E-C AUGMENT                                 SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary       Scope     Drawing Item No     Comp ID       Method    Component Description                         Period Class       System                  Code Case/Relief Request                    1   2 3                         Comments E-C         579800        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 240-Y  GENERAL - PANEL ADJACENT TO 240-X(L) & 240-Z(R)                  1128 MC          8010          DETAILED  RR-01 E-C         584500        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 248-BB GENERAL - PANEL ADJACENT TO 248-AA(L) & 248-CC(R)                1128 MC          8010          DETAILED  RR-01 E-C         590600        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 256-SS GENERAL - PANEL ADJACENT TO 256-RR(L) & 256-TT(R)                1128 MC          8010          DETAILED  RR-01 E-C         563700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 264-GG GENERAL - PANEL ADJACENT TO 264-FF(L) & 264-HH(R)                1128 MC          8010          DETAILED  RR-01 E-C         597600        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 268-DD GENERAL - PANEL ADJACENT TO 268-CC(L) & 268-EE(R)                1128 MC          8010          DETAILED  RR-01 E-C         597700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 268-EE GENERAL - PANEL ADJACENT TO 268-DD(L) & 268-FF(R)                1128 MC          8010          DETAILED  RR-01 E-C         597900        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 268-GG GENERAL - PANEL ADJACENT TO 268-FF(L) & 268-HH(R)                1128 MC          8010          DETAILED  RR-01 E-C         601000        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 272-FF GENERAL - PANEL ADJACENT TO 272-EE(L) & 272-GG(R)                1128 MC          8010          DETAILED  RR-01 E-C         601100        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 272-GG GENERAL - PANEL ADJACENT TO 272-FF(L) & 272-HH(R)                1128 MC          8010          DETAILED  RR-01 E-C         602300        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 272-TT GENERAL - PANEL ADJACENT TO 272-SS(L) & 272-UU(R)                1128 MC          8010          DETAILED  RR-01 E-C         569800        IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-A  GENERAL - PANEL ADJACENT TO 280-CCC(L) & 280-B(R)                CALCULATION RNP-C/STRU-1128 MC          8010          DETAILED  RR-01 E-C         570200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-E  GENERAL - PANEL ADJACENT TO 280-D(L) & 280-F(R)
MC          8010          DETAILED  RR-01 RNP-PM-007                                                           Revision 5                                                      Page 1 of 5

H. B. ROBINSON UNIT 2                     CATEGORY - E-C AUGMENT                                 SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary        Scope     Drawing Item No     Comp ID        Method    Component Description                         Period Class       System                   Code Case/Relief Request                    1   2 3                          Comments E-C         570400         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-G   GENERAL - PANEL ADJACENT TO 280-F(L) & 280-H(R)
MC          8010           DETAILED  RR-01 E-C         570700         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-L   GENERAL - PANEL ADJACENT TO 280-K(L) & 280-M1(R)
MC          8010           DETAILED  RR-01 E-C         570800         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-M1  GENERAL - PANEL ADJACENT TO 280-L(L) & 280-M2(R)
MC          8010           DETAILED  RR-01 E-C         570900         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-M2  GENERAL - PANEL ADJACENT TO 280-M1(L) & 280-M3(R)
MC          8010           DETAILED  RR-01 E-C         631200         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 280-P   GENERAL - PANEL ADJACENT TO 280-O(L) & 280-Q(R)                  1128 MC          8010           DETAILED  RR-01 E-C         633400         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-LL  GENERAL - PANEL ADJACENT TO 280-KK(L) & 280-MM(R)
MC          8010           DETAILED  RR-01 E-C         633500         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-MM  GENERAL - PANEL ADJACENT TO 280-LL(L) & 280-NN(R)
MC          8010           DETAILED  RR-01 E-C         572800         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      BULGES AND PITTING EVALUATED UNDER EC 72699, RNP E4.11       L-WALL 280-SS  GENERAL - PANEL ADJACENT TO 280-RR(L) & 280-TT(R)                CALCULATION RNP-C/STRU-1128 AND RNP-C/STRU-1130 MC          8010           DETAILED  RR-01 E-C         572900         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      BULGES AND PITTING EVALUATED UNDER EC 72699, RNP E4.11       L-WALL 280-TT  GENERAL - PANEL ADJACENT TO 280-SS(L) & 280-UU(R)                CALCULATION RNP-C/STRU-1128 AND RNP-C/STRU-1130 MC          8010           DETAILED  RR-01 E-C         573000         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-UU  GENERAL - PANEL ADJACENT TO 280-TT(L) & 280-VV(R)                CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         573100         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-VV  GENERAL - PANEL ADJACENT TO 280-UU(L) & 280-WW(R)                CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         573400         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-AAA GENERAL - PANEL ADJACENT TO 280-ZZ(L) & 280-BBB(R)               CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 RNP-PM-007                                                            Revision 5                                                      Page 2 of 5

H. B. ROBINSON UNIT 2                     CATEGORY - E-C AUGMENT                                 SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary        Scope     Drawing Item No     Comp ID        Method    Component Description                         Period Class       System                   Code Case/Relief Request                    1   2 3                         Comments E-C         573500         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-BBB GENERAL - PANEL ADJACENT TO 280-AAA(L) & 280-CCC(R)              CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         571200         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 284-B   GENERAL - PANEL ADJACENT TO 284-A(L) & 284-C(R)                  CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         571800         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 284-H   GENERAL - PANEL ADJACENT TO 284-G(L) & 284-I(R)                  1128 MC          8010           DETAILED  RR-01 E-C         571900         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 284-I   GENERAL - PANEL ADJACENT TO 284-H(L) & 284-J(R)                  1128 MC          8010           DETAILED  RR-01 E-C         575300         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 284-UU  GENERAL - PANEL ADJACENT TO 284-TT(L) & 284-VV(R)                1128 MC          8010           DETAILED  RR-01 E-C         637000         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 284-VV  GENERAL - PANEL ADJACENT TO 284-UU(L) & 284-WW(R)                1128 MC          8010           DETAILED  RR-01 E-C         575500         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 284-BBB GENERAL - PANEL ADJACENT TO 284-AAA(L) & 284-CCC(R)              CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         573700         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 284-CCC GENERAL - PANEL ADJACENT TO 284-BBB(L) & 284-A(R)                CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         638100         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-G   GENERAL - PANEL ADJACENT TO 288-F(L) & 288-H(R)                  1128 MC          8010           DETAILED  RR-01 E-C         638300         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-I   GENERAL - PANEL ADJACENT TO 288-H(L) & 288-J(R)                  1128 MC          8010           DETAILED  RR-01 E-C         639300         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-P   GENERAL - PANEL ADJACENT TO 288-O(L) & 288-Q(R)                  1128 MC          8010           DETAILED  RR-01 E-C         642300         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-TT  GENERAL - PANEL ADJACENT TO 288-SS(L) & 288-UU(R)                1128 MC          8010           DETAILED  RR-01 RNP-PM-007                                                            Revision 5                                                     Page 3 of 5

H. B. ROBINSON UNIT 2                    CATEGORY - E-C AUGMENT                                SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary       Scope     Drawing Item No     Comp ID       Method    Component Description                         Period Class       System                  Code Case/Relief Request                    1   2 3                        Comments E-C         642400        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-UU GENERAL - PANEL ADJACENT TO 288-TT(L) & 288-VV(R)                1128 MC          8010          DETAILED  RR-01 E-C         642500        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-VV GENERAL - PANEL ADJACENT TO 288-UU(L) & 288-WW(R)                1128 MC          8010          DETAILED  RR-01 E-C         644100        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 292-I  GENERAL - PANEL ADJACENT TO 292-H(L) & 292-J(R)                  1128 MC          8010          DETAILED  RR-01 E-C         648000        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 292-UU GENERAL - PANEL ADJACENT TO 292-TT(L) & 292-VV(R)                1128 MC          8010          DETAILED  RR-01 E-C         648100        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 292-VV GENERAL - PANEL ADJACENT TO 292-UU(L) & 292-WW(R)                1128 MC          8010          DETAILED  RR-01 E-C         648200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 292-WW GENERAL - PANEL ADJACENT TO 292-VV(L) & 292-XX(R)                1128 MC          8010          DETAILED  RR-01 E-C         649500        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 296-G  GENERAL - PANEL ADJACENT TO 296-F(L) & 296-H(R)                  1128 MC          8010          DETAILED  RR-01 E-C         649600        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 296-H  GENERAL - PANEL ADJACENT TO 296-G(L) & 296-I(R)                  1128 MC          8010          DETAILED  RR-01 E-C         649700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 296-I  GENERAL - PANEL ADJACENT TO 296-H(L) & 296-J(R)                  1128 MC          8010          DETAILED  RR-01 E-C         650700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 296-P  GENERAL - PANEL ADJACENT TO 296-O(L) & 296-Q(R)                  1128 MC          8010          DETAILED  RR-01 E-C         655400        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 300-H  GENERAL - PANEL ADJACENT TO 300-G(L) & 300-I(R)                  1128 MC          8010          DETAILED  RR-01 E-C         659700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 300-VV GENERAL - PANEL ADJACENT TO 300-UU(L) & 300-WW(R)                1128 MC          8010          DETAILED  RR-01 RNP-PM-007                                                           Revision 5                                                    Page 4 of 5

H. B. ROBINSON UNIT 2                    CATEGORY - E-C AUGMENT                                 SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary       Scope     Drawing Item No     Comp ID       Method    Component Description                         Period Class       System                  Code Case/Relief Request                    1   2 3                         Comments E-C         661300        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 304-I  GENERAL - PANEL ADJACENT TO 304-H(L) & 304-J(R)                  1128 MC          8010          DETAILED  RR-01 E-C         665200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130 E4.11       L-WALL 304-TT GENERAL - PANEL ADJACENT TO 304-SS(L) & 304-UU(R)
MC          8010          DETAILED  RR-01 E-C         655300        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 304-UU GENERAL - PANEL ADJACENT TO 304-TT(L) & 304-VV(R)                1128 AND BULGE EVALUATED IN RNP CALCULATION RNP-MC          8010          DETAILED  RR-01                                                  C/STRU-1130 E-C         655400        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130 E4.11       L-WALL 304-VV GENERAL - PANEL ADJACENT TO 304-UU(L) & 304-WW(R)
MC          8010          DETAILED  RR-01 E-C         666900        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-H  GENERAL - PANEL ADJACENT TO 308-G(L) & 308-I(R)                  1128 MC          8010          DETAILED  RR-01 E-C         668200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-R  GENERAL - PANEL ADJACENT TO 308-Q(L) & 308-S(R)                  1128 MC          8010          DETAILED  RR-01 E-C         668900        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-Y  GENERAL - PANEL ADJACENT TO 308-X(L) & 308-Z(R)                  1128 MC          8010          DETAILED  RR-01 E-C         671000        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-TT GENERAL - PANEL ADJACENT TO 308-SS(L) & 308-UU(R)                1128 AND BULGE EVALUATED IN RNP CALCULATION RNP-MC          8010          DETAILED  RR-01                                                  C/STRU-1130 E-C         575600        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-UU GENERAL - PANEL ADJACENT TO 308-TT(L) & 308-VV(R)                1128 MC          8010          DETAILED  RR-01 E-C         671100        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-VV GENERAL - PANEL ADJACENT TO 308-UU(L) & 308-WW(R)                1128 MC          8010          DETAILED  RR-01 E-C         671200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-WW GENERAL - PANEL ADJACENT TO 308-VV(L) & 308-XX(R)                1128 MC          8010          DETAILED  RR-01 RNP-PM-007                                                           Revision 5                                                      Page 5 of 5

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 2 Pages (including this cover page)
CONTAINMENT LINER DETAIL AT MOISTURE BARRIER

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==REFERENCES:==
: 1) G-190343
: 2) G-190353
: 3) G-190358
: 4) G-190359 fl      .~ ~                          - - - FLOOR  SLAB             5) RELIEF REQUEST IWE/IWL-02 FIGURE 1 EL 226'-0" -   --
                                                 .       })   ~~~~~~~D.Oi~::X:X:X:Z:~ X"       TOP OF MAT LINER BASE MAT - - -
0    ,,                EC 65306
                                                                                         - ~" TO ~" LINER TRANSITION REV IMTE              DESCRIPTION PROF'ESSIONAL ENGINEER:
OUAl.ITY lMl: SAFETY RELATED PROGRESS ENERGY                   ~~     rogress ENGINEERING SECTION                 ~J Ener Pl.ANT: ROBINSON PLANT - UNIT 2              SCAl.E: NONE TITLE:            EXAMINATION BOUNDARY OF CONTAINMENT LINER (DETAIL AT MOISTURE BARRIER)
FIGURE 16 F'llE:

R. Michael Glover
  /_~ DUKE                                                                          H. B. Robinson Steam Electric Plant Unit 2 ENERGY'&deg;                                                                       Site Vice President Duke Energy Progress 3581 West Entrance Road Hartsville, SC 29550 0 . 843 857 1704 F: 843 857 1319 Mike.Glawr@*.lke-ener!1f.can 10 CFR 50.90 Serial: R.Ne-RA/15-0090 NUV I 9 Z015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261 I RENEWED LICENSE NO. DPR-23 Proposed Amendment to Technical Specification 5.5.16 for the Adoption of Option B of 1 O CFR 50, Appendix J for Type B and Type C Testing and the Permanent Change in 1 OCFR 50, Appendix J, Integrated Leak Rate Test Interval and Type C Leak Rate Testing Frequency

==Dear Sir/Madam:==

In accordance with the provisions of 10 CFR 50.90 Duke Energy Progress, Inc. is submitting a request for an amendment to the technical specifications (TS) for H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2). The proposed change is a request to revise TS 5.5.16, Containment Leakage Rate Testing Program, TS 3.6.1, Containment, and TS 3.6.2, Containment Air Lock, for HBRSEP2, to allow the following:
Increase in the existing integrated leak rate test (ILRT) program test interval from 10 years to 15 years, Adopt 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, for the performance-based testing of Types B and C tested components in accordance with the guidance of Technical Specification Task Force (TSTF)-52, Implement 10 CFR 50, Appendix J, Option B (Reference 11 of the Enclosure),
Allow an extension to the 120-month frequency currently permitted by Option B for Type B leakage rate testing, Allow an extension from the 60-month frequency currently permitted by Option B to a 75-month frequency for Type C leakage rate testing,

U. S. Nuclear Regulatory Commission Serial: RNP-RA/15-0090 Page 2 The proposed change would also adopt a more conservative grace interval of 9 months, for Type Band Type C tests in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, revision 3-A.
The Enclosure provides a description and basis of the proposed change, a No Significant Hazards Consideration Determination, and an Environmental Analysis. Attachment 1 provides the existing TS pages marked up to show the proposed changes. Attachment 2 provides revised (clean) TS pages that reflect the proposed change. Attachment 3 provides the existing TS bases pages marked up to show the proposed changes, and are provided for information only. Attachment 4 provides an evaluation of risk significance of permanent ILRT extension. provides E-C AUGMENT plan tables for the second ten-year interval. Attachment 6 provides the containment liner detail at moisture barrier.
HBRSEP2 requests approval of the proposed license amendment by November 30, 2016, with the amendment being implemented within 120 days of issuance.
In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated South Carolina Official.
Please address any comments or questions regarding this matter to Mr. Richard Hightower, Manager - Nuclear Regulatory Affairs at (843) 857-1329.
I declare under penalty of perjury that the foregoing is true and correct. Executed on N0\1 ,~..:.-   ,,  I 2015.
Sincerely, R. Michael Glover Site Vice President RMG/jmw Enclosure Attachments
: 1. Proposed Technical Specification Changes
: 2. Revised Technical Specification Pages
: 3. Proposed Technical Specifications Bases Changes
: 4. Evaluation of Risk Significance of Permanent ILRT Extension
: 5. E-C AUGMENT Plan Tables for the Fifth Ten-Year Interval
: 6. Containment Liner Detail at Moisture Barrier cc: Region Administrator, NRC, Region II Ms. Martha C. Barillas, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP2 Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)

Enclosure Page 1 of 60 Evaluation of the Proposed Change

==Subject:==
Proposed Amendment to Technical Specification 5.5.16 for the Adoption of Option B of 10 CFR 50, Appendix J for Type B and Type C Testing and the Permanent Change in 10 CFR 50, Appendix J, Integrated Leakage Rate Test Interval and Type C Leak Rate Testing Frequency 1.0   

==SUMMARY==
DESCRIPTION 2.0   DETAILED DESCRIPTION

==3.0   TECHNICAL EVALUATION==

==4.0   REGULATORY EVALUATION==

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

==5.0   ENVIRONMENTAL CONSIDERATION==

==6.0   REFERENCES==

Enclosure Page 2 of 60 1.0      

==SUMMARY==
DESCRIPTION Pursuant to 10 CFR 50.90, Duke Energy Progress requests an amendment to the H. B.
Robinson Steam Electric Plant Unit No. 2 (HBRSEP2) Renewed Facility Operating License (DPR-23) by incorporating the attached proposed change into the HBRSEP2 Technical Specifications (TS). Specifically, the proposed change is a request to revise TS 5.5.16, "Containment Leakage Rate Testing Program," TS 3.6.1, Containment, and TS 3.6.2, Containment Air Lock, for HBRSEP2, to allow the following:
* Increase in the existing Type A integrated leakage rate test (ILRT) program test interval from 10 years to 15 years in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A and the conditions and limitations specified in NEI 94-01, Revision 2-A.
* Adopt 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, for the performance-based testing of Types B and C tested components in accordance with the guidance of Technical Specification Task Force (TSTF)-52, Implement 10 CFR 50, Appendix J, Option B (Reference 11).
* Adopt an extension of the containment isolation valve leakage testing (Type C) frequency from the 60 months currently permitted by 10 CFR 50, Appendix J, Option B, to a 75-month frequency for Type C leakage rate testing of selected components, in accordance with NEI 94-01, Revision 3-A.
* Adopt the use of ANSI/ANS 56.8-2002, Containment System Leakage Testing Requirements.
* Adopt a more conservative grace interval of 9 months, for Type A, Type B and Type C tests in accordance with Nuclear Energy Institute (NEI) Topical Report NEI 94-01, Revision 3-A.
The proposed change to the TS contained herein would revise HBRSEP2 TS 5.5.16, by replacing the reference to Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, (Reference 1) and 10 CFR 50, Appendix J, Option A with a reference to NEI topical report NEI 94-01, Revision 3-A (Reference 2), dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A (Reference 3), dated October 2008, as the implementation documents used by HBRSEP2 to implement the performance-based leakage testing program in accordance with 10 CFR 50, Appendix J, Option B. This license amendment request (LAR) also proposes an administrative change to TS 5.5.16 by deleting the information regarding the performance of the next Type A test no later than April 9, 2007, as this has already occurred.
The proposed change to the TS contained herein would revise HBRSEP2 TS SRs 3.6.1.1 and 3.6.2.1, by replacing the reference to 10 CFR 50, Appendix J, Option A with a reference to the Containment Leakage Rate Testing (CLRT) program and incorporate the changes recommended by TSTF-52, Revision 3, as applicable to HBRSEP2.
The associated TS Bases for SR 3.0.2, SR 3.6.1.1 and SR 3.6.1.2 are also being revised to reflect the proposed change removing references to 10 CFR 50, Appendix J, Option A and

Enclosure Page 3 of 60 incorporate the bases changes recommended by TSTF-52 as applicable to HBRSEP2.
2.0     DETAILED DESCRIPTION 2.1     Current Containment Leakage Rate Testing Program This program provides controls for implementation of the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guidelines contained in RG 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception:
: a.      NEI 94 1995, Section 9.3.2: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than April 9, 2007.
Types B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix J, Option A.
2.2     TS Change Description The following changes are requested as part of the implementation of 10 CFR 50, Appendix J, Option B, as described in TSTF 52, Revision 3, Implement 10 CFR 50, Appendix J, Option B, dated March 2000.
Surveillance Requirement (SR) 3.6.1.1, Containment, currently states:
SURVEILLANCE                                 FREQUENCY SR 3.6.1.1     Perform required Type B and C leakage rate           --------NOTE-------
testing except for containment air lock testing, in  SR 3.0.2 is not accordance with 10 CFR 50, Appendix J, Option        applicable A, as modified by approved exemptions.               ------------------------
The leakage rate acceptance criterion is  1.0 La. In accordance However, during the first unit startup following     with 10 CFR 50, testing performed in accordance with 10 CFR 50,      Appendix J, Appendix J, Option A, as modified by approved        Option A, as exemptions, the leakage rate acceptance criterion    modified by is < 0.6 La for the Type B and Type C tests.         approved exemptions TS SR 3.6.1.1 is being revised for consistency with TSTF-52 and the adoption of the guidelines contained in NEI Topical Report, NEI 94-01, Revision 3-A for Type B and Type C local leakage rate testing (LLRT). The proposed change will revise TS SR 3.6.1.1 to state:
SURVEILLANCE                                 FREQUENCY SR 3.6.1.1     Perform required visual examinations and Type B      --------NOTE-------
and C leakage rate testing except for containment    SR 3.0.2 is not air lock testing, in accordance with 10 CFR 50,      applicable

Enclosure Page 4 of 60 Appendix J, Option A, as modified by approved                 ------------------------
exemptions.
The leakage rate acceptance criterion is  1.0 La.            In accordance However, during the first unit startup following              with 10 CFR 50, testing performed in accordance with 10 CFR 50,               Appendix J, Appendix J, Option A, as modified by approved                 Option A, as exemptions, the leakage rate acceptance criterion             modified by is < 0.6 La for the Type B and Type C tests. the              approved Containment Leakage Rate Testing Program.                     exemptions the Containment Leakage Rate Testing Program SR 3.6.1.3, currently states:
SURVEILLANCE                                         FREQUENCY SR 3.6.1.3     Perform required visual examinations and Type A               In accordance leakage rate testing, in accordance with the                  with the Containment Leakage Rate testing Program                      Containment Leakage Rate Testing Program The proposed change will delete TS SR 3.6.1.3 in its entirety as it is now incorporated into SR 3.6.1.1:
SURVEILLANCE                                         FREQUENCY SR 3.6.1.3     Perform required visual examinations and Type A               In accordance leakage rate testing, in accordance with the                  with the Containment Leakage Rate testing Program                      Containment Leakage Rate Testing Program TS 3.6.2, Containment Air Lock, Required Action C.1 is being revised to remove an extra revision bar character. This is an administrative change. Please refer to the markup in .
SR 3.6.2.1, Containment Air Lock, currently states:
SURVEILLANCE                                         FREQUENCY SR 3.6.2.1     ---------------------------NOTES-----------------------------
: 1.         An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage

Enclosure Page 5 of 60 test.
: 2.         Results shall be evaluated against acceptance criteria of SR 3.6.1.1, in accordance with 10 CFR 50, Appendix J,                    --------NOTE-------
Option A, as modified by approved                         SR 3.0.2 is not exemptions.                                               applicable Perform required air lock leakage rate testing in                   In accordance accordance with 10 CFR 50, Appendix J, Option                       with 10 CFR 50, A, as modified by approved exemptions.                              Appendix J, Option A, as modified by approved exemptions The TS SR 3.6.2.1 is being revised for consistency with TSTF-52 and the adoption of the guidelines contained in NEI Topical Report, NEI 94-01, Revision 3-A for Containment Air Lock LLRT. The proposed change will revise TS SR 3.6.2.1 to state, as follows:
SURVEILLANCE                                               FREQUENCY SR 3.6.2.1     ---------------------------NOTES-----------------------------
: 1.         An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
: 2.         Results shall be evaluated against acceptance criteria of applicable to SR 3.6.1.1, in accordance with 10 CFR 50,                    --------NOTE-------
Appendix J, Option A, as modified by                      SR 3.0.2 is not approved exemptions.                                      applicable Perform required air lock leakage rate testing in                   In accordance accordance with 10 CFR 50, Appendix J, Option                       with 10 CFR 50, A,                                                                  Appendix J, as modified by approved exemptions. the                             Option A, as Containment Leakage Rate Testing Program.                           modified by approved exemptions the Containment Leakage Rate Testing Program.
HBRSEP2 TS 5.5.16, "Containment Leakage Rate Testing Program," currently states:

Enclosure Page 6 of 60 This program provides controls for implementation of the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception:
: a.      NEI 94 1995, Section 9.3.2: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than April 9, 2007.
Type B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix J, Option A.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42 psig, which exceeds the peak calculated containment internal pressure for the design basis loss of coolant accident.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of the containment air weight per day.
Leakage rate acceptance criteria are:
: a.      Containment leakage rate acceptance criteria is < 1.0 La. During the first unit 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 Type A tests.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
The proposed changes to HBRSEP2 TS 5.5.16, "Containment Leakage Rate Testing Program" will be (1) the administrative change to delete the performance of the next Type A test no later than April 9, 2007, that was previously approved by the NRC via Amendment No. 193, (2) to include the addition of the references to the guidelines contained in NEI Topical Report, NEI 94-01, Revision 3-A and the conditions and limitations contained in NEI Topical Report, NEI 94-01 Revision 2-A for Type A, Type B and Type C LLRT, and (3) incorporate the updates recommended by TSTF-52. The proposed change will revise TS 5.5.16, as follows, to state:
A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42 psig. The containment design pressure is 42 psig.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of the containment air weight per day.

Enclosure Page 7 of 60 Leakage rate acceptance criteria are:
: a.      Containment leakage rate acceptance criteria is < 1.0 La. During the first unit 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 Type A tests.
: b.      Air lock testing acceptance criteria are:
: 1. Overall air lock leakage rate is < 0.05 La when tested at > Pa.
: 2. For each door, leakage rate is < 0.01 La when pressurized to > 42 psig.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.
Mark-ups of TS 5.5.16; SR 3.6.1.1; SR 3.6.1.3; TS 3.6.2, Required Action C.1; and, SR 3.6.2.1 are provided in Attachment 1. The retyped TS pages are provided in Attachment 2.
A markup of TS Bases for SR 3.0.2, TS 3.6.1, SR 3.6.1.1, SR 3.6.1.3, B 3.6.2, and SR 3.6.2.1 are provided in Attachment 3 for informational purposes only.
This proposed change is requested to extend the performance of the next HBRSEP2 ILRT from the Spring 2017 refueling outage to a subsequent refueling outage no later than Spring 2021. contains the plant specific risk assessment conducted to support this proposed change. This risk assessment followed the guidelines of Nuclear Regulatory Commission (NRC) RG 1.174 (Reference 4) and NRC RG 1.200, Revision 2 (Reference 5). The risk assessment concluded that the increase in risk as a result of this proposed change is considered to be insignificant since it represents a very small change to the HBRSEP2 risk profile.
2.3      Deviations From TSTF-52, Revision 3 TS 1.1, Definitions TSTF-52 deletes La from the definitions in TS 1.1.
Deviation - There is no definition for La in TS 1.1 of the HBRSEP2 TS, so this change is unnecessary.

==3.0      TECHNICAL EVALUATION==

3.1      Justification for the Technical Specification Change

Enclosure Page 8 of 60 The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the containment, including systems and components that penetrate the containment, does not exceed the allowable leakage values specified in the TS. 10 CFR 50, Appendix J, also ensures that periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment and the systems and components penetrating primary containment. The limitation on containment leakage provides assurance that the containment would perform its design function following an accident up to and including the plant design basis accident. Appendix J identifies three types of required tests: (1) Type A tests, intended to measure the primary containment overall integrated leakage rate; (2) Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage limiting boundaries (other than valves) for primary containment penetrations; and (3) Type C tests, intended to measure containment isolation valve leakage rates. Types B and C tests identify the vast majority of potential containment leakage paths.
Type A tests identify the overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Types B and C testing.
In 1995, 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors," was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach.
Performance-based test intervals are based on consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50, Appendix J, refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.
Also, in 1995, RG 1.163 was issued. The RG endorsed NEI 94-01, Revision 0, (Reference 6) with certain modifications and additions. Option B, in concert with RG 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory ILRT performance history (i.e., two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A (ILRT) test from three tests in 10 years to one test in 10 years. This relaxation was based on an NRC risk assessment contained in NUREG-1493 (Reference 7), and Electric Power Research Institute (EPRI) TR-104285 (Reference 8), both of which showed that the risk increase associated with extending the ILRT surveillance interval was very small. In addition to the 10-year ILRT interval, provisions for extending the test interval an additional 15 months were considered in the establishment of the intervals allowed by RG 1.163 and NEI 94-01, but that this "should be used only in cases where refueling schedules have been changed to accommodate other factors."
In 2008, NEI 94-01, Revision 2-A, (Reference 3) was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the NRC Safety Evaluation Report (SER) on NEI 94-01. The NRC SER was included in the front matter of this NEI report. NEI 94-01, Revision 2-A, includes provisions for extending Type A ILRT intervals to up to fifteen years and incorporates the regulatory positions stated in Regulatory Guide 1.163 (September 1995). It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights.

Enclosure Page 9 of 60 Acceptability for referencing by licensees proposing to amend their TS is provided in Section 5.0 of the SER and states the following:
The NRC staff, therefore, finds that this guidance is acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.0 of this SE. In addition, in accordance with the NRC staff's resolution of the comments provided by NEI on the draft SE, the following changes will be made by NEI to the "-A" version of the TR. Therefore, consistent with the language in this final SE:
A. NEI TR 94-01, Revision 2, will be revised in the "-A" version of the report, as discussed in the last paragraph of Section 3.1.2.2, "Extending Type B & C Test Intervals," to the final SE.
B. EPRI Report No. 1009325, Revision 2, will be revised in the "-A" version of the report, to change the population dose acceptance guidelines and the CCFP guidelines. (As stated in Section 4.2 of the final SE Limitation and Condition #2).
In 2012, NEI 94-01, Revision 3-A, was issued. This document describes an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR 50, Appendix J, and includes provisions for extending Type A ILRT intervals to up to fifteen years. NEI 94-01 has been endorsed by RG 1.163 and NRC SERs dated June 25, 2008 (Reference 9) and June 8, 2012 (Reference 10) as an acceptable methodology for complying with the provisions of Option B to 10 CFR Part 50. The regulatory positions stated in RG 1.163 as modified by NRC SERs dated June 25, 2008, and June 8, 2012, are incorporated in this document. It delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights. Extensions of Type B and Type C test intervals are allowed based upon completion of two consecutive periodic as-found tests where the results of each test are within a licensees allowable administrative limits.
Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment airlocks) and up to a maximum of 75 months for Type C tests. If a licensee considers extended test intervals of greater than 60 months for Type B or Type C tested components, the review should include the additional considerations of as-found tests, schedule and review as described in NEI 94-01, Revision 3-A, Section 11.3.2.
Acceptability for referencing by licensees proposing to amend their TS is provided in Section 5.0 of the SER and states the following:
The NRC staff, therefore, finds that this guidance, as modified to include two limitations and conditions, is acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing. Any applicant may reference NEI TR 94-01, Revision 3, as modified by this SE and approved by the NRC, in a licensing action to satisfy the requirements of Option B to 10 CFR Part 50, Appendix J. The NRC staff is not required to repeat its review of the matters described in the TR conditioned upon the changes described in this SE (Sections 3 and 4) to be incorporated when the report appears as a reference which was complied with a request for relief, or other related licensing actions.

Enclosure Page 10 of 60 NEI 94-01, Revision 3-A, Section 10.1 concerning the use of grace in the deferral of Type B and Type C LLRTs past intervals of up to 120 months for the recommended surveillance frequency for Type B testing and up to 75 months for Type C testing, states:
Consistent with standard scheduling practices for Technical Specifications Required Surveillances, intervals of up to 120 months for the recommended surveillance frequency for Type B testing and up to 75 months for Type C testing given in this section may be extended by up to 25 percent of the test interval, not to exceed nine months.
Notes: For routine scheduling of tests at intervals over 60 months, refer to the additional requirements of Section 11.3.2.
Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. This provision (nine-month extension) does not apply to valves that are restricted and/or limited to 30-month intervals in Section 10.2 (such as BWR MSIVs) or to valves held to the base interval (30 months) due to unsatisfactory LLRT performance.
3.1.1   Current HBRSEP2 10 CFR 50, Appendix J Requirements Title 10 CFR Part 50, Appendix J, was revised, effective October 26, 1995, to allow licensees to choose containment leakage testing under either Option A, "Prescriptive Requirements," or Option B, "Performance Based Requirements." HBRSEP2 has implemented the requirements of 10 CFR Part 50, Appendix J, Option A for Types B and C testing and Option B for Type A testing. Current TS 5.5.16 requires the following:
This program provides controls for implementation of the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception:
: a.      NEI 94 1995, Section 9.3.2: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than April 9, 2007.
Type B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix J, Option A.
RG 1.163, Section C.1, states that licensees intending to comply with 10 CFR Part 50, Appendix J, Option B, should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01 (Reference 6) rather than using test intervals specified in American National Standards Institute (ANSI)/American Nuclear Society (ANS) 56.8-1994. NEI 94-01, Section 11.0, refers to Section 9, which states that Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per ten years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage was less than 1.0 La (where La is the maximum allowable leakage rate at design pressure). Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine performance shall be at least 24 months.

Enclosure Page 11 of 60 Adoption of the Option B performance based containment leakage rate testing program altered the frequency of measuring primary containment leakage in Type A tests but did not alter the basic method by which Appendix J leakage testing is performed. The test frequency is based on an evaluation of the "as found" leakage history to determine a frequency for leakage testing which provides assurance that leakage limits will not be exceeded. The allowed frequency for Type A testing as documented in NEI 94-01, is based, in part, upon a generic evaluation documented in NUREG-1493 (Reference 7). The evaluation documented in NUREG-1493 included a study of the dependence or reactor accident risks on containment leak tightness for differing types of containment types, including a reinforced, shallow domed concrete containment similar to HBRSEP2 containment structures. NUREG-1493 concluded in Section 10.1.2 that reducing the frequency of Type A tests (ILRT) from the original three tests per ten years to one test per twenty years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Types B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements. Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, NUREG-1493 concluded that increasing the interval between ILRTs is possible with minimal impact on public risk.
3.1.2   HBRSEP2 10 CFR 50, Appendix J, Option B Licensing History Amendment No. 169 - SER dated May 28, 1996 - ML020530639 (Reference 11)
The Commission issued Amendment No.169 to Facility Operating License No. DPR-23 for the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP2). This amendment changed the HBRSEP2 TS in response to the request dated January 31, 1996.
The amendment revised TS section 4.4 to allow the use of 10 CFR Part 50, Appendix J, Option B, Performance-Based Containment Leakage Rate Testing. A new TS section 6.12 described the containment leakage rate testing program that adopted 10 CFR Part 50, Appendix J, Option B for Type A tests; and 10 CFR Part 50, Appendix J, Option A, for Types B and C tests.
Amendment No. 193 - SER dated September 16, 2002 - ML022690765 (Reference 12)
The Commission issued Amendment No. 193 to Facility Operating License No. DPR-23 for HBRSEP2. This amendment consisted of changes to the TS in response to the application dated March 26, 2002, as supplemented by letters dated June 19, and August 8, 2002.
The amendment to the TS for HBRSEP2 modified TS SR 5.5.16, Containment Leakage Rate Testing Program, to require the performance of a Type A test within 12.1 years from the last Type A test, which was performed on April 9, 1992. This was a one-time extension to the 10-year performance-based Type A test interval based on an acceptably low level of risk as supported by a plant-specific risk assessment.
Amendment No. 199 - SER dated February 11, 2004 - ML040430023 (Reference 13)
The Commission issued Amendment No. 199 to Facility Operating License No. DPR-23 for HBRSEP2. This amendment changed the HBRSEP2 TS in response to the request dated June 11, 2003, as supplemented by letters dated August 20, and October 13, 2003.

Enclosure Page 12 of 60 The amendment allows the extension of the Appendix J, Type A, Containment Integrated Leak Rate Test, Option B, for HBRSEP2 from the scheduled May 2004 timeframe to no later than April 9, 2007.
Amendment No. 215 - SER dated June 15, 2007 - ML071070170 (Reference 14)
The Commission issued Amendment No. 215 to Renewed Facility Operating License No. DPR-23 for HBRSEP2. This amendment changed the TS in response to the application dated July 17, 2006. Specifically, the amendment revised the containment design pressure in SRs 3.6.8.1 and 3.6.8.5 concerning the Isolation Valve Seal Water System, and TS Section 5.5.16, Containment Leakage Rate Testing Program.
Amendment No. 220 - SER dated October 3, 2008 - ML0082210549 (Reference 15)
The Commission issued Amendment No. 220 to Renewed Facility Operating License No. DPR-23 for HBRSEP2 in response to the application dated November 29, 2007. The amendment consisted of changes to TS Section 3.6.8, "Isolation Valve Seal Water (IVSW) System."
The amendment revised SRs 3.6.8.2 and 3.6.8.6 related to IVSW tank volume and header flow rates. Specifically, the change clarified the wording of SR 3.6.8.2, and revised SR 3.6.8.6 to provide a total flow rate limit from all four headers in place of the individual header limits.
Relief Request IST-RR SER dated July 13, 2012 - ML12174A010 (Reference 16)
By letter to the NRC dated March 16, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12086A067), as supplemented by letter dated May 10, 2012, (ML12138A041), Carolina Power & Light Company doing business as Progress Energy Carolinas, Inc., submitted Relief Request-3 for the Inservice Testing (1ST) Program Plan for the fifth 10-year Interval for the H. B. Robinson Steam Electric Plant, Unit No.2 (HBRSEP).
HBRSEP requested approval to use an alternative test plan in lieu of certain IST requirements of the 2004 Edition through 2006 Addenda of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM) Code for the Category C check valves exercise tests or exams at HBRSEP. Specifically, pursuant to Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(a)(3)(i), HBRSEP requested to use proposed alternatives on the basis that the alternatives provided an acceptable level of quality and safety.
The NRC staff determined that the proposed alternative, described in IST-RR-3, provided reasonable assurance that valves IVSW-71, IVSW-72, IVSW-74 through IVSW-97, IVSW-100A, IVSW-100B, and IVSW-100C were operationally ready and provided an acceptable level of quality and safety. Accordingly, the NRC staff concluded that HBRSEP adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55(a)(3)(ii) and is in compliance with the ASME OM Code requirements. Therefore, the NRC staff authorized the proposed alternative in IST-RR-3 for the fifth IST interval at HBRSEP Unit No.2, which began on July 21, 2012, and is scheduled to end on February 18, 2022.
3.1.3    Containment Building Description General Description of the Containment (Reference UFSAR Section 3.8.1.1.1)

Enclosure Page 13 of 60 The reactor containment structure is a steel lined concrete shell in the form of a vertical right cylinder with a hemispherical dome and a flat base supported by means of piles. The containment structure is designed for an accident pressure based upon the pressure transients as shown in UFSAR Section 15.6. The containment structure is designed to contain radioactive material, which might be released from the core following a loss-of-coolant accident as described in UFSAR Section 6.2.1.
The structure consists of sidewalls measuring 126 ft. from the liner on the base to the springline of the dome and an inside diameter of 130 ft. The containment free volume is 1,950,000 ft3. The sidewalls of the cylinder and dome are 3 ft. 6 in. and 2 ft. 6 in. thick, respectively. The inside radius of the dome is equal to the inside radius of the cylinder (i.e., the discontinuity at the springline due to the change in thickness is on the outer surface). The base consists of a 10-foot thick structural concrete slab. The base liner is installed on top of the structural slab and covered with two feet of concrete.
The basic structural elements considered in the design of the containment structure are the piles, base slab, sidewalls, and dome acting as essentially one structure under all loading conditions. The bottom plates of the liner are laid loose on the foundation slab and are anchored only at the hang ways for the crane wall and primary shield. In the vertical walls and dome, the liner is anchored to the concrete shell by means of "KSM" shaped anchor studs fusion welded to the liner plate so that it forms an integral part of the entire composite structure under all loadings. The cylindrical portion of the liner is insulated. The dome of the containment is reinforced concrete. The cylinder walls are concrete-reinforced circumferentially and prestressed vertically. The base slab is reinforced concrete.
Containment Liner (Reference UFSAR Section 3.8.1.1.5)
The containment liner is designed to serve as a leakproof membrane and is not relied upon for the structural integrity of the containment except for resisting tangential shears in the dome. It is anchored to the concrete by means of "KSM" shaped steel studs. The liner is not anchored to the concrete base slab hence does not act compositely with it. It was laid loose on the base slab and the butt weld backing strips were set in grooves in the base slab. After welding, the distortions in the liner were considered too great and a neat cement grout was flowed beneath it to fill the voids. A bond breaker, form oil, was flowed first on the base slab to prevent the liner from acting compositely with the slab.
Stress conditions in the liner under all conditions of design have been analyzed to assure that the principle stresses do not exceed the yield or buckling stresses as provided in design stress criteria. Fatigue, accident, and operational loads are discussed in UFSAR Section 3.8.1.4.
Sample buckling calculations are shown indicating maximum loading conditions that prevail. A discussion on the liner anchors and transfer on their loads into the concrete has been submitted in the Containment Design Report.
The loading condition which produces maximum biaxial compression in the liner is that of winter operation combined with 1.0 times the hypothetical earthquake. Under this condition, the allowable buckling stress is not exceeded.
The steel liner and its welded seam joints are covered by carbon steel channels with pressurizing connections. These seam weld channels can be used to determine the leak-tightness of the liner seam welds.

Enclosure Page 14 of 60 The liner plate was protected against corrosion as follows: The entire inside surfaces of the dome and walls were sandblasted. The plate from elevation (el.) 228' to approximately el. 352' was coated with a zinc rich primer and modified alkyd topcoat (Keeler and long 6820/7230 system). The plate above el. 352' was coated with a zinc filed inorganic primer such as Carbozinc 11 and Phenoline 305 phenolic epoxy topcoat. The coated surfaces of the liner plate are protected by insulation and sheathing up to el. 367'10". All painted surfaces were inspected after erection and any damaged areas reprimed before finishing. The face of the liner plate in contact with the concrete has no primer or paint applied; the intimate contact with the concrete provides corrosion protection. Maintenance coatings are specified and applied equal to or better than the original coatings. (Reference UFSAR 3.8.1.6.1.5)
The only exception to the above described Liner Protection coating and inspection is in the areas of the field welds at the new Conax Cartridge Type Electrical Penetrations. The inside surface of the field weld is inaccessible after installation, and the coating and inspection described above cannot take place. These welds are radiographed at installation.
The containment liner is 1/2 in. thick from el. 226'-0" to el. 253'-0". The liner is 3/8 in. thick from el. 253'-0" to 352'-0" and 1/2 in. thick above 352'-0". The liner on top of the base mat at el. 226-0 is 1/4 in. thick. A partial diagram of the containment liner is provided in Attachment 6, Containment Liner Detail At Moisture Barrier.
Containment Penetrations Penetrations through the containment reinforced concrete pressure barrier for pipe, electrical conductors, ducts, and access hatches are typically of the double barrier type. In general, a penetration consists of a sleeve embedded in the concrete wall and welded to the containment liner. The weld to the liner is shrouded by a channel, which was used to demonstrate the integrity of the penetration-to-liner weld joint. The pipe, electrical conductor cartridge, or duct passes through the embedded sleeve and the ends of the resulting annulus are closed off, either by welded end plates, bolted flanges, or a combination of these. Provisions are made for differential expansion and misalignment between pipe or cartridge, and sleeve. Pressurizing connections are provided to demonstrate the integrity of the penetration assemblies.
(Reference UFSAR3.8.1.1.6)
An exception to this are electrical penetrations C1, C2, C3, C5, C9, C10, D9, E1, E5, E10 and F1. These penetrations have double pressure barrier protection in their header plate and therefore an endplate is required at one end only.
Electrical Penetrations (Reference UFSAR 3.8.1.1.6.1)
Cartridge type penetrations are used for all electrical conductors passing through the containment, with the exception of the penetration at sleeve numbers C9, C10, D9, E5 and E10 in the north cable vault, and sleeve numbers C1, C2, C3, C5, E1 and F1 in the south cable vault which are of the capsule type design. The penetration cartridge is a hollow cylinder closed on both ends, through which the conductors pass. This cartridge is provided with a pressure connection to allow continuous or intermittent pressurization of the penetration. The method used to seal the joint between the cartridge end plate and the conductor depends upon the type of cable involved.
In general, there are four types used:
: 1. Type 1 - High voltage power, 4160 volts

Enclosure Page 15 of 60
: 2. Type 2 - Power, control, and instrumentation, 600 volts and below
: 3. Type 3 - Thermocouple leads
: 4. Type 4 - Coaxial and triaxial cables Type 1 penetrations are rubber insulated copper rods. These insulated rods will pass through a leak tight gland fitting threaded into each end plate of the cartridge. Either alumina insulating bushings or fused glass seals may be used to provide the double barrier.
Type 2 penetrations are single or multiconductor mineral insulated cable with a metallic sheath.
This cable will pass through a leak tight gland threaded into each end of the penetration cartridge. The ends of the mineral insulated cable are potted with epoxy resin. Copper rod conductors with fused glass seals in the cartridge and plates is an alternate which may be used.
Type 3 penetrations are the same as Type 2 except the conductors will be thermocouple material. The sealing methods are the same as for the Type 2 penetrations.
Type 4 penetrations are used for coaxial and triaxial cables. In addition to the leak tight gland fittings in the cartridge end plate, a plug and receptacle connection provides a double barrier to leakage through the cable itself. An alternate method uses fused glass seals in the cartridge end plates and fused glass seals between the conductors of the coaxial or triaxial cable.
In the capsule penetration design, a single stainless plate is machined with the required quantity of feed-through ports which are interconnected by peripherally machined gun drills which creates a manifold system for pressure monitoring. Feed-throughs are assembled through the plate and sealed in place with a patented metal compression fitting assembly which creates seal zones at the front and backside of the plate, while allowing for a chamber to form between the seal zones to accommodate leakage monitoring.
The capsule series penetration is designed for a weldment interface to the containment nozzle.
The weldment interface is by a transition ring, factory welded to the penetration header plate, and field welded to the containment nozzle.
The penetration sleeves to accommodate the electrical penetration assembly cartridges are 10 in., Schedule 80 carbon steel pipe, except where otherwise noted. For the electrical penetrations C1, C2, C3, C5, C9, C10, D9, E1, E5, E10, and F1, the header plate and conductors are pressurized. There are 51 electrical penetrations.
Piping Penetrations (Reference UFSAR 3.8.1.1.6.2)
Double barrier piping penetrations are typically provided for piping passing through the containment. The pipe is centered in the embedded sleeve which is welded to the liner, except for small pipes where several pipes may pass through the same penetration sleeve.
The penetrations for the main steam, feedwater, blowdown, and sample lines are designed so that the penetration is stronger than the piping system and that the vapor barrier will not be breeched due to a hypothesized pipe rupture.
Typically, end plates are welded to the pipe at both ends of the sleeve. Several pipes may pass through the same embedded sleeve to minimize the number of penetrations required. In this case, each pipe is welded to both end plates. A connection to the penetration sleeve is provided to allow continuous or intermittent pressurization of the compartment formed between the piping

Enclosure Page 16 of 60 and the embedded sleeve. In the case of piping carrying hot fluid, the pipe is insulated. The RHR supply pipe is insulated to keep the concrete surrounding the embedded sleeve below 200&#xba;F. Typical hot and cold pipe penetrations are shown in UFSAR Figure 3.8.1-15. There are 46 containment penetrations sleeves for pipes. Pipes are anchored to the structural steel girders as close as possible to the inside of the wall or to the crane wall. Loads due to pipe ruptures within the containment or due to thermal stresses are not transferred to the liner.
An exception to this is the steam generator blowdown penetrations and two safety injection penetrations (RHR penetrations S-14 and S-15). The end plate is welded directly to the sleeve.
The sleeve is welded to the liner reinforcement plate. Piping loads are transmitted to the concrete wall, except for torsion loads, which are carried by the liner plate. However, the torsion loads are below the liner allowable stress.
Two piping penetrations are provided in the containment sump area.
Equipment and Personnel Access Hatches (Reference UFSAR 3.8.1.1.6.3)
An equipment hatch is provided which is fabricated from welded steel and furnished with a double-gasketed flange and bolted dished door. Equipment up to a diameter of approximately 18 ft. can be transferred into and out of containment via this hatch. The hatch barrel is embedded in the containment wall and welded to the liner and is a portion of the structural frame embedded in the wall. Provision is made to pressurize the space between the double gaskets of the door flanges and the weld seam channels at the liner joint, hatch flanges, and dished door. Pressure is relieved from the double gasket spaces prior to opening the door.
The personnel hatch is a double door, hydraulically-latched, welded steel assembly. It is attached to the structural frame embedded in the wall of which the frame barrel forms the central portion of the lock. A quick-acting type, equalizing valve connects the personnel hatch with the interior of the containment vessel for the purposes of equalizing pressure in the two systems when entering or leaving the containment.
The personnel hatch doors are interlocked to prevent both being opened simultaneously and to ensure that one door is completely closed before the opposite door can be opened. Indicating lights and annunciators situated in the control room indicate the door operational status.
Provision is made to permit bypassing the door interlocking system to allow doors to be left open during plant cold shutdown. Each door lock hinge is designed to be capable of independent three-dimensional adjustment to assist proper seating. An Emergency Lighting and Communication System operating from an external emergency supply is provided in the lock interior. Emergency access to either the inner door, from the containment interior; or to the outer door, from outside, is possible by the use of special door unlatching tools.
Fuel Transfer Penetration A fuel transfer penetration is provided for fuel movement between the refueling transfer canal in the reactor containment and the spent fuel pit. The penetration consists of a 20 in. stainless steel pipe installed inside a 24 in. pipe (Reference UFSAR Figure 3.8.1-16). The inner pipe acts as the transfer tube and is fitted with a double-gasketed blind flange in the refueling canal and a standard gate valve in the spent fuel pit. This arrangement prevents leakage through the transfer tube in the event of an accident. The outer pipe is welded to the containment liner and provision is made by use of a special seal ring for testing all welds essential to the integrity of

Enclosure Page 17 of 60 the penetration. Bellows expansion joints are provided on the pipes to compensate for any differential movement between the two pipes or other structures. (Reference UFSAR 3.8.1.1.6.4)
Containment Supply and Exhaust Purge Ducts The ventilation system purge ducts are each equipped with two quick-acting tight-sealing butterfly valves for isolation purposes. The valves are manually opened for containment purging, but are automatically actuated to the closed position upon a safety injection signal or high containment radiation level signal. (Reference UFSAR 3.8.1.1.6.5)
Containment Dome The dome is a hemispherical dome 65 ft. inside diameter and 2 ft. 6 in. thick reinforced concrete. The difference in cylinder and dome thickness is effected on the outside surface, the transition between thicknesses being accomplished 13 ft. above the springline of the dome at the anchor surface of the cylinder prestressing steel tendons. The inside of the dome is insulated from the springline to a point above the anchor surface of the cylinder prestressing steel tendons. The outer surface of the dome is covered with a membrane roof to provide weather protection. (Reference UFSAR 3.8.1.1.7)
Insulation Containment liner insulation consists of 44 in. x 84 in. x 1 1/4 in. thick, 4 lb/ft3 density crosslinked PVC foam and/or 2 lb/ft3 density Polyimide foam with an outer covering of 0.019 in.
thick stainless steel. Panels are erected with the 44 in. dimension vertical and the 84 in.
dimension horizontal. (Reference UFSAR 3.8.1.6.1.7)
Prestressing Steel Tendon Design The prestressing system chosen for post-tensioning the Robinson containment structure in the vertical direction, consists of 1 3/8 in. diameter high strength steel bars closely grouped into tendons consisting of six bars per tendon.
These tendons are placed within heavy wall 6 in. galvanized steel pipe sheaths.
Tendons are on the centerline of the wall and are spaced approximately every 3 ft around the periphery of the containment. This concept of grouping a number of high strength bars is not new to the industry, a seven bar group having been used in a dam in South America. It utilizes standard concepts and components proven adequate through experience in the United States, and more experience using the similar system in Europe.
The bottom anchorage is a steel plate with six threaded holes into which are screwed the steel bars. The thread used is a tapered, cut thread designed and proven to develop the minimum guaranteed ultimate tensile strength of the bar. Complete seating of the thread is necessary to develop the full tensile strength of the bar and this is easily accomplished and inspected since the thread is fully engaged when no threads are showing at the inner face of the plate.
Threaded holes in the bottom bearing plate have been sealed from the bottom by a steel plug with a binding compound between the plug and the bar. The entire bottom anchorage was designed and tested to show no permanent physical distortion at the minimum ultimate tensile strength of the tendon, which is a 25 percent greater load than the maximum load to which it will be subjected in the life of the structure.

Enclosure Page 18 of 60 Couplings consist of internally threaded sleeves into which the high strength steel bars are securely screwed. The same screw thread details are used as described for the bottom anchorage. The void space between the bars within the coupler body is filled with the same binding compound used to bind the threads to eliminate any possibility of corrosion.
The tendon coupling consists of a set of six individual bar couplers staggered in elevation. Each tendon has two couplers in its length, one at the construction joint in the cylinder wall at El. 250 ft. which is a field assembled coupling and a second half way up the wall between the construction joint and the top anchorage which is a shop assembled coupling. To assure the integrity of the coupler, the threads are coated with an epoxy compound, which binds the coupler sleeve to the bar and prevents the possibility of unthreading due to vibration during shipping or erection. Once tensioned, the friction within the coupler threads eliminates any tendency for unthreading.
The top anchorage consists of a steel plate bearing on the concrete with three of the bars anchored to this plate by means of Howlett Grip Nuts. A second and smaller plate bears on the top of these grip nuts and the remaining three bars are anchored to the top plate by means of Howlett Grip Nuts. The Howlett Grip Nut is a modified positive action wedge anchor, which has the advantages of a wedge anchor and the positive adjustment capability of a threaded anchor.
It does not require an exact predetermination of bar length along with all of the fine shimming required at the top anchorage with such a predetermination. This concept of stacking the top anchorage details allows a closer grouping of bars than is ordinarily required with a bar system.
The steel sheath surrounding each tendon is made of 6 in. Schedule 40 galvanized steel pipe with threaded and flanged connections. The sheath is connected to the bottom anchorage plate by means of a threaded coupling and provides protection of the tendon both during and after construction.
The corrosion protection scheme used during construction was a nitrogen atmosphere, a system widely used for corrosion protection. As an additional check, six extra removable bars were placed in six of the lower tendon assemblies. These bars were removed to monitor the effectiveness of the nitrogen atmosphere corrosion protection. All tendons were inspected visually from the top before coupling or grouting to check that corrosion had not occurred.
The tendons were initially tensioned in late March/early April 1970, and then re-tensioned in May 1970. The last exterior wall concrete lifts were placed in December 1969 and the first lifts were placed in November 1968. (Reference UFSAR 3.8.1.4.7)
Tendon Surveillance Duke Energy Progress (DEP) believes that there is sufficient evidence in the history of the prestressed concrete industry to justify the specifying of an uniaxially prestressed concrete containment vessel such as the HBRSEP2 containment with full confidence that it will perform within the criteria set in its design. Conservative values have been used in estimating qualities of materials, which affect the net prestressing force. (UFSAR 3.8.1.7.2)
As an example of the conservatism used, consider creep and shrinkage. Design values of 0.0003 in. shrinkage and 2.25 for coefficient of creep (creep strain/elastic strain) were specified.
It is expected that values of 0.0001 in. shrinkage and 1.7 for coefficient of creep are realistic values based on preliminary estimates using as a guide Hanson, T. C. and Mattock A. H.,

Enclosure Page 19 of 60 "Influence of Size and Shape of Member on the Shrinkage and Creep of Concrete," ACI Jour.,
Proceedings, Vol. 63, p. 267 (1966).
Such a conservative design can only result in higher precompression stresses in the concrete and higher tensile stresses in the tendons. This is of little interest, since even with these higher tensile stresses, the tendons will never reach the tensile stress imposed upon them with the initial prestressing operation.
There is no practical method of surveying the tendon stress and corrosion, creep and shrinkage of the concrete for a grouted tendon. Known conservative analytical procedures, in addition to successful experience application for grouted tendons, do not warrant a surveillance program.
However, two surveillance tendons similar to the service tendons and in a similar environment are provided. These may be uncovered at any time for surveillance of any corrosion.
The surveillance tendons consist of two short tendons similar to the service tendons. Each tendon consists of six 3/8 in. &#xf8; bars in 6 in. pipe sheath with anchor plates, prestressing hardware, and grout pipe identical except for length to the working tendons. They are embedded in a section of concrete approximating the same environment as that of the service tendons.
The program for inspection consisted of removing one tendon after 5 years and the other after 25 years.
The removed tendons were sent to a commercial laboratory qualified to perform material tests and analysis. The tendon bars were removed from the sheath and the grout removed. The visual inspection was performed to detect and record evidence of corrosion. Tensile tests were then performed on selected bars to develop stress-strain diagrams and determine the bars' ultimate tensile strengths. The results of these tests were compared with the original properties to determine any significant changes. DEP retains a qualified engineering firm to assess the results of these tests and make recommendations.
The first containment surveillance tendon sample was removed in March 1976. The second containment surveillance tendon sample was removed in April 1997. Based on the information presented in the reports, and on other available data on the Robinson containment system, the tendon surveillance program is judged to be satisfactory. The tests showed that surveillance tendon specimens tested exceeded the minimum-breaking load of 238,000 pounds given in the FSAR. It can be reasonably concluded that similar results would be obtained if bars from the actual containment tendons inservice were tested.
3.1.4    Isolation Valve Seal Water System (IVSW)
The IVSW System assures the effectiveness of certain containment isolation valves during any condition, which requires containment isolation, by providing a water seal at the valves. These valves are located in lines that are connected to the Reactor Coolant System (RCS), or that could be exposed to the containment atmosphere in the event of a loss of coolant accident (LOCA). The system provides a reliable means for injecting seal water between the seats and stem packing of the globe and double disc types of isolation valves, and into the piping between other closed isolation valves. The system provides assurance that, should an accident occur, the containment leak rate is no greater than that assumed in the accident analysis by providing seal water at a pressure  1.1 times Pa (46.2 psig). The system is designed to maintain this seal

Enclosure Page 20 of 60 for at least 30 days. The possibility of leakage from the containment or RCS past the first isolation point is thereby prevented by assuring that if leakage does exist, it will be from the IVSW System into containment.
The system includes one 175-gallon seal water tank capable of supplying the total requirements of the system. The IVSW tank's required volume is maintained and the tank is pressurized with nitrogen. The normal supply of makeup water to the IVSW tank is the Primary Water System. In the event Primary Water is not available, emergency makeup can be supplied from the Service Water System. The Plant Nitrogen System provides the normal supply of nitrogen to the IVSW tank. An automatic backup supply is provided from two dedicated high pressure nitrogen bottles.
The system is normally in a static condition with the seal water injection tank filled and pressurized. Indication of IVSW tank level and pressure along with corresponding low level and low pressure alarms are provided in the Control Room. The tank supplies pressurized water to four distribution headers. Header "A" requires manual operation and serves lines that are normally filled with fluid following a LOCA, and lines that must remain in service for a period of time following the accident. Headers "B", "C", and "D" are automatic headers that are pressurized through one or both of two redundant, fail open, air operated valves arranged in parallel. A loss of power will cause these valves to fail open. System operation is initiated by a Phase A containment isolation signal which accompanies any Safety Injection (SI) signal.
Safety Analyses The Design Basis Accident (DBA) that results in a release of radioactive material within containment is a LOCA. The analyses for the LOCA assumes the isolation of containment is completed and leakage from containment is at a rate equivalent to the design leakage rate. As part of the containment boundary, containment isolation valves function to support the leak tightness of containment. By maintaining this barrier, offsite dose calculations will be less than the limits of 10 CFR 100 or 10 CFR 50.67, as applicable, during a DBA.
The IVSW System actuates on a containment isolation signal and functions to assure the actual leakage is no greater than the design value. IVSW assures the effectiveness of certain isolation valves to limit containment leakage by pressurizing the affected containment penetration flow paths at a pressure  1.1 times Pa. IVSW is designed to maintain this seal for at least 30 days.
A single failure analysis shows the failure of any active component will not prevent fulfilling the design function of the system. By meeting these requirements, IVSW is considered a qualified seal system in accordance with 10 CFR 50, Appendix J.
The IVSW System satisfies Criterion 3 of the NRC Policy Statement.
Acceptance Review / Evaluation (Reference 28)
By letter dated April 23, 1979, the NRC completed the review and evaluation regarding the acceptability of the IVSW system. The results of this evaluation was as follows:
Paragraph III.C.3.b of Appendix J to 10 CFR 50 requires that the installed isolation valve seal water system fluid inventory be sufficient to assure the sealing function for at least 30 days at a pressure of 1.1 Pa. The IVSW system proposed by the licensee can provide seal water at a pressure equal or greater than 1.1 Pa and for a period greater than 30 days. We, therefore, conclude that the proposed IVSW system meets the requirements of Appendix J.

Enclosure Page 21 of 60 We have also reviewed the system design with respect to the requirements of an engineered safety feature, because the system will be used and relied upon during and following an accident. Based on the information in the licensee's submittal and the FSAR, we find that the proposed system including associated components, piping, and structures are designed to Class I seismic criteria. In addition, two separate, independent, seismically qualified sources of makeup water are provided for long term operation at a pressure greater than 1.1 Pa. A single failure analysis shows that the failure of any single active component will not prevent fulfilling the design function of the system. We, therefore, conclude that the proposed IVSW system meets the requirements of an engineered safety feature, and can be relied upon to fulfill its design function during and following an accident.
3.1.5   Containment Penetration Pressurization System (PPS)
The Containment PPS provides a means of testing pressure zones incorporated into the containment penetrations. It was originally designed to provide a means of continuously pressurizing the positive pressure zones in order to maintain these zones above the maximum containment post-accident pressure and to provide a means for continuous or intermittent monitoring of the leakage status of the containment penetrations.
Modification ESR/MOD 95-00888 removed the automatic continuous pressurization and monitoring features of this system. It is now only used during power operation to test the personnel hatch and during outages to test containment penetrations (LLRTs). The system is capable of providing continuous pressurization should the need arise.
3.2     Inspections 3.2.1   Primary Containment Coatings Condition Assessment Assessment of the protective coatings inside Primary Containment is conducted during each refueling outage. Protective coatings inside Primary Containment are assessed to identify and quantify coatings degradation and unqualified coatings. The condition assessment is consistent with ASTM D5163 and NUREG 1801, Chapter XI.S8, "Protective Coating Monitoring and Maintenance Program", and meets the requirements for protective coating monitoring and maintenance.
Coatings are considered to be acceptable, provided none of the following conditions are observed:
* Blistering is not greater than size No. 6 (Medium) as specified in ASTM D714
* Cracking greater than standard No. 6 as specified in ASTM D661 (Checking of any grade specified in ASTM D660 is acceptable and need not be recorded)
* Flaking greater than standard No. 6 as specified in ASTM D772
* Rusting equal to or greater than Grade 7 as specified in ASTM D610
* Rust staining, accumulated dirt or dirt containing iron compounds should not be confused with actual rusting of the steel substrate. Only rusting of the substrate under a coating need be considered.
* Insufficient adhesion, as determined by the Coating Program Manager (e.g., visual observation of loose coatings, qualitative or quantitative assessment, etc.). Adhesion

Enclosure Page 22 of 60 tests may be conducted to quantify the condition of coating adhesion per ASTM D4541 or ASTM D7234.
* Unqualified Coatings.
Mechanical damage is considered to be acceptable, provided none of the above noted conditions are observed in the surrounding coatings, and provided that any exposed substrate or underlying coating has been evaluated for acceptability.
Areas within the scope of this assessment, but not inspected, should be identified with justification provided. Justification should state reason the areas is considered inaccessible and assess potential for adverse impact on emergency core cooling system (ECCS) Sump performance from potential coating debris from the un-inspected areas.
Areas inspected under the ASME Section XI, Subsection IWE/IWL program do not require examination in accordance with the coatings assessment, although the results of the Service Level 1 coating inspections shall be included in the evaluation.
Recordable conditions shall be addressed in accordance with the Corrective Action Program.
Multiple recordable conditions found during any one outage assessment may be documented on the same NCR. The NCR should include consideration of the following:
* Past and projected impact on ECCS performance
* Extent of condition
* Maintenance Rule impact
* Corrective actions taken and/or planned If corrective action is taken to remediate a recordable condition during the refueling outage in which it was identified, a follow-up assessment of the area shall be performed.
A Condition Assessment Summary Report shall be prepared by the Program Manager, which includes the following:
* A brief description of assessment results and the general condition of Primary Containment coatings
* Apparent cause(s) of degradation and trend evaluation
* Recommendations for future repairs and surveillance
* Condition Assessment forms which include:
o Location, characterization, and disposition of recordable conditions o Corrective actions taken and recommended o Photographic documentation
* Total quantity of Unqualified Coatings (UC) and comparison to previous inspection results and the applicable Maintenance Rule and/or design limits for ECCS functionality
* List of NCRs initiated as a result of the condition assessment
* Identity and functional role (i.e.; QC, Coatings Program Manager, Coatings Service Level I Contractor, etc.) of those individuals performing the inspections.
Include the Service Level III - Safety Related coatings in the Condition Assessment Report as they relate to Primary Containment conditions.
3.2.2    Inservice Inspection Program for Containment - IWE/IWL On August 8, 1996, an amendment to 10 CFR 50.55a was published in the Federal Register (61 FR 41303) to incorporate by reference, the requirements of the ASME Code, Section XI,

Enclosure Page 23 of 60 Subsections IWE and IWL. The objective of this amendment was to specify requirements to ensure that the critical areas of a containment structure were routinely inspected to detect and correct defects that could compromise its structural integrity. The rationale of the NRC to issue this amendment was based on (1) the rate of occurrence of degradation of containments and (2) their determination that current licensee containment inspection programs were not adequate to detect degradation. The effective date of this amendment to 10 CFR 50.55a was September 9, 1996.
Until September 9, 1996, requirements for the inservice inspection of components classified as Class MC and CC were not mandated by Federal Regulation. For this reason, preservice and/or inservice inspection of the components classified as Class MC and CC at HBRSEP2 had not been performed in previous Inspection Intervals per ASME Section XI.
This program implements required License Renewal commitments.
IWE/IWL Program Description The program was developed to implement the requirements of ASME Section XI as modified by 10 CFR 50.55a. The examinations required by this program may be utilized to satisfy 10 CFR 50, Appendix J inspections and10 CFR 50.65 Maintenance Rule inspections, as deemed appropriate by plant organizations responsible for the implementation of those programs.
The components subject to ASME Section XI, Subsection IWE and IWL requirements are those that make up the containment structure, its leak tight barrier (including integral attachments) and those that contribute to its structural integrity. These components are listed in the Second Ten-Year Inservice IWE/IWL Inspection Plan.
Specifically included are Class MC pressure retaining components and their integral attachments, (including metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments), per IWE-1100; and Class CC reinforced concrete containments and unbonded posttensioning systems, per IWL-1100. Because Subsection IWL requires examination and testing of unbonded post-tensioning systems only, the grouted bonded post-tensioning system at HBRSEP2 is not subject to ASME Section XI rules.
The First Containment Inspection Interval at HBRSEP2 was effective from September 9, 1998 to September 8, 2008, for Subsection IWE and Subsection IWL activities. This time frame included the first and second five-year examinations required by Subsection IWL. As required by the NRC final rulemaking, this plan was developed to support the completion of expedited examinations by September 9, 2001. The applicable ASME Section XI Code utilized for the First Ten Year Interval was the 1992 Edition, 1992 Addenda.
The Second Containment Inspection Interval at HBRSEP2 is effective from September 9, 2008, to September 8, 2018, for Subsection IWE and Subsection IWL activities.
The applicable ASME Section XI Code Year and Addenda for the Second Ten-Year IWE/IWL Program is the 2001 Edition with the 2003 Addenda of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, Subsections IWA, IWE and IWL and Inspection Program B of IWE-2412 and IWL-2420.
The Class MC components subject to examination include the following:
* Pressure Retaining Metal Containment Liner

Enclosure Page 24 of 60
* Penetration Assemblies
* Airlock
* Equipment Hatch
* Moisture Barriers
* Pressure Retaining Bolting (Shown on the applicable Examination Boundary Drawings)
* Integral Attachments (Shown on the applicable Examination Boundary Drawings)
Per IWE-1220, the following components (or parts of components) are exempted from the examination requirements of IWE-2000. This section describes some of the design and access limitations. In addition, specific access component limitations are typically identified on the visual examination data sheets.
* Vessels, parts, and appurtenances that are outside the boundaries of the containment as defined in the Design Specification.
* Embedded or inaccessible portions of containment vessels, parts, and appurtenances that met the requirements of the Original Construction Code; components that meet this criteria include:
o Containment Vessel (CV) liner made inaccessible by the concrete floor slab at elevation 228 (see detail G-G on drawing G-190367) o Nelson studs welded to the CV liner embedded in the containment concrete o CV liner at the containment sumps embedded in concrete o Penetration assembly for the 14 ECCS containment sump piping (Penetrations S-X1 and S-X2) embedded in concrete (see drawings G-190267 and G-190359) o CV liner made inaccessible by weld test channels.
* The CV Liner considered inaccessible by the insulation and sheathing between elevations 228 and 367-10. Although not exempted from examination per IWE-1220, the CV Liner at these locations are considered inaccessible, unless a maintenance activity requires removal of the insulation and sheathing.
* Portions of containment vessels, parts, and appurtenances that becomes embedded or inaccessible as a result of vessel repair or replacement if the conditions of IWE-1232(a) and (b) and IWE-5220 are met.
As allowed by 10 CFR 50.55a(b)(2)(ix)(C), HBRSEP2 has elected to exclude Category E-B and Category E-F weld examinations from the Subsection IWE portion of the Inservice IWE/IWL Program.
IWE Examination Schedule The current period/interval schedule for IWE examinations is summarized in the Table below and scheduling of individual components per the requirements of Table IWE-2412-1 is contained in the Second Ten Year Interval Inservice IWE/IWL Inspection Plan.
Table 3.2.2-1, IWE PERIOD/INTERVAL SCHEDULE Interval               Period 1                Period 2              Period 3 1                 9/98 - 9/01            9/01 - 9/05            9/05 - 9/08 2                 9/08 - 9/11            9/11 - 9/15            9/15 - 9/18 3                 9/18 - 9/21            9/21 - 9/25            9/25 - 9/28 Inaccessible Class MC Areas

Enclosure Page 25 of 60 Inaccessible Class MC components subject to 10 CFR 50.55a requirements are those that meet the criteria of 10 CFR 50.55a(b)(2)(ix)(A), which requires the identification and evaluation of inaccessible Class MC areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
The following areas that meet this criterion have been identified and evaluated:
* Moisture barrier degradation (Behind six insulation panels)
* CV Liner at Elevation 228. (Behind six insulation panels)
* CV Liner from Elevation 232 to Elevation 367 (behind insulation)
* CV Liner behind weld channels (does not meet criteria of 10 CFR 50.55a(b)(2)(ix)(A).
* CV Liner below the floor slab.
The requirements of Subsection IWL for Class CC components are identified on the Examination Boundary Drawings and are listed in the Second Ten Year Interval Inservice IWE/IWL Inspection Plan.
The Class CC components subject to examination include the following:
* Dome roof
* Cylinder walls Per IWL-1220, the following components (or parts of components) are exempted from the examination requirements of IWL-2000. This section describes some of the design and access limitations. In addition, specific component access limitations are typically identified on the visual examination data sheets.
* Tendon end anchorages that are inaccessible, subject to the requirements of IWL-2521.1. This exemption is not applicable to HBRSEP2. The plant design includes a bonded post tensioning system. Per IWL-2220.2, examination requirements are applicable to unbonded Post-Tensioning systems.
* Portions of the concrete surface that are covered by liner, foundation material, or backfill, or are otherwise obstructed by adjacent structures, components, parts, or appurtenances. Components that meet this criteria include:
o Inside face of containment concrete made inaccessible by the liner o All surfaces of the vertical containment concrete slab made inaccessible by the liner and floor slab (inside containment) and by backfill (outside containment)
(see drawing G-190358).
o Outside face of containment concrete made inaccessible by the adjacent structures such as the auxiliary and fuel buildings.
Although normally exempt from the examination requirements of IWL-2000, when below-grade concrete is exposed by excavation for any reason, these IWL components are subject to examination consistent with the requirements for nonexempt IWL components in accordance with the respective License Renewal Commitments.
Degradation of below-grade IWL concrete attributed to aggressive groundwater will be used as a leading indicator for potential degradation to other below grade concrete structures in the scope of license renewal. Trending requirements for structures based on monitoring aggressive ground water are addressed in the respective License Renewal Commitments.
The piles supporting the containment concrete slab are outside the scope of Subsection IWL.

Enclosure Page 26 of 60 Implementation of the Subsection IWL portion of the Inservice IWE/IWL Program from a schedule standpoint will be in accordance with the Second Ten-Year Inservice IWE/IWL Inspection Plan IWL-2400, Inspection Schedule, which requires concrete to be examined at 1, 3-, and 5-year frequencies following the containment Structural Integrity Test (SIT) and every 5 years thereafter per IWL-2410. The 10-year and subsequent examinations shall commence not more than 1 year prior to the specified dates and shall be completed not more than 1 year after such dates. If plant-operating conditions are such that examination of portions of the concrete cannot be completed within this stated time interval, examination of those portions may be deferred until the next regularly scheduled plant outage.
IWL Examination Schedule The current schedule for IWL examinations is summarized in the Table below.
Table 3.2.2-2, IWL Schedule Inspection                               Year of Inspection 1st 5-Year Inspection                                  2000 nd 2 5-Year Inspection                                    2005 3rd 5-Year Inspection                                  2010 4th 5-Year Inspection                                  2015 5th 5-Year Inspection                                  2020 th 6 5-Year Inspection                                    2025 7th 5-Year Inspection                                  2030 The evaluation of nondestructive examinations of Subsection IWL components will be performed in accordance with Article IWL-3000.
The repair of Subsection IWL components will be performed in accordance with Article IWL-4000 (per IWA-4000).
Pressure testing of IWL components will be in accordance with Subsection IWL-5000.
Inaccessible Class CC Areas The components subject to examination per 10 CFR 50.55a are those that meet the criteria of 10 CFR 50.55a(b)(2)(viii)(E), which requires licensees to identify and evaluate the inaccessible Class CC areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas.
Currently, no areas have been identified that meet this criterion. In the event that such areas are identified, these areas will be added to the IWE/IWL Program scope of components and will be evaluated.
Relief Requests (RRs)
This Program contains requests for relief from certain Code requirements. The bases for these RRs and proposed alternative examinations or requirements are summarized within Appendix B, Relief Requests. More detailed descriptions, bases, and proposed alternative examinations

Enclosure Page 27 of 60 or requirements associated with each request for relief were submitted as a separate enclosure to the NRC.
The portions of the concrete surface that are covered by the liner, foundation material, or backfill or are otherwise obstructed by adjacent structures, components or appurtenances are exempt from the requirements of ASME Section XI, IWL-1220(b).
Examinations of components are scheduled in accordance with Inspection Program B (IWA, IWE, IWL and Tables IWE-2500-1 and IWL-2500-1). Examinations are scheduled, to the extent practical, based upon the preceding sequence of the First Ten-Year Interval IWE/IWL Inspection Plan.
IWE/IWL Augmented Examinations In accordance with the Second Ten-Year Interval IWE/IWL Program and Plan, IF areas are identified requiring Augmented Examinations (E-C) during the interval, THEN they will be listed in the E-C AUGMENT section of the Second Ten-Year Interval Plan.
Examinations identified in the Plan shall show the period in which the deficiency was identified as well as the successive examination period.
If an augmented examination is identified during the third period of an interval, the scheduling and re-examination shall occur during the first period of the successive interval.
The listing of augmented examinations is provided in Attachment 5, E-C AUGMENT Plan Tables for the Second Ten-Year Interval.
New Technologies to Perform Inspections of Inaccessible Class MC/CC Areas HBRSEP2 has not needed to implement any new technologies to perform inspections of any inaccessible areas at this time. However, Duke Energy Progress actively participates in various nuclear utility owners groups and ASME Code committees to maintain cognizance of ongoing developments within the nuclear industry. Industry operating experience is also continuously reviewed to determine its applicability to HBRSEP2. Adjustments to inspection plans and availability of new, commercially available technologies for the examination of the inaccessible areas of the containment would be explored and considered as part of these activities.
ASME Code Cases There are NO ASME Code Cases applicable to HBRSEP2.

Enclosure Page 28 of 60 Relief Requests Relief       Affected      Examination   Item   Examination         Alternative         Status Request     Component(s)       Category     No.        Area         Examinations Number IWE/IWL     Containment      E-A           E1.11  Insulated      General/Detailed      Granted RR-01       Liner                                 portions of    visual exam when      TAC No.
the            panels are removed    (MD8509)
Containment    for maintenance       Dated Liner          purposes and 1/3 of   4/30/2009 the lowest elevation panels per Period, 100% for the Interval IWE/IWL     Containment      E-A           E1.30  Containment    General/Detailed      Granted RR-02       Moisture Barrier                      moisture       visual exam when      TAC No.
barrier        panels are removed    (MD8509) for maintenance       Dated purposes and 1/3 of   4/30/2009 the lowest elevation panels per Period, 100% for the Interval A review of the data associated with the moisture barrier inspections indicated degradation of the moisture barrier in some locations, which required removal and reapplication of the moisture barrier in the First Ten-Year Interval. Liner inspections had revealed different grades of degradation of coatings, which required removal and reapplication of coatings in the First Ten-Year Interval. Because of the degraded conditions identified during the First Ten-Year Interval, the above listed relief requests will be discussed in detail in support of the continued acceptability of CV liner and moisture barriers.
The reliefs will be implemented during the HBRSEP2, Second Ten-Year IWE/IWL Inspection Interval (September 9, 2008, through September 8, 2018) for containment inspections required by ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Subsections IWE and IWL.
Relief Request No. IWE/IWL-RR-01, Visual Examination Of Insulated Containment Liner Code Requirements for Which Relief is Requested The ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Table IWE-2500-1, "Examination Categories," Examination Category E-A, Item Number E1.11, requires a general visual examination of 100% of the accessible surface areas of containment in accordance with acceptance standard IWE-3510, "Standards for Examination Category E-A, Containment Surfaces," paragraph IWE-3510.2, "Visual Examination of Coated and Noncoated Areas," for Class MC and metallic liners of Class CC components.
This relief was previously approved for the First Ten-Year IWE/IWL Inspection Interval under a Safety Evaluation Report (TAC No. MA4637) dated July 26, 1999.
Specific Relief Requested Relief is requested from performing general visual examinations in accordance with ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Table IWE-2500-1, Examination Category E-A, Item Number E 1.11 on the accessible surface areas of the containment liner,

Enclosure Page 29 of 60 which are insulated. The code requires 100% examination of the liner each examination period.
Proposed alternative examinations are provided below.
This request for relief is applicable to the insulated portion of the containment liner classified as Class MC and subject to the requirements of Table IWE-2500-1, Examination Category E-A.
Alternative Examination(s)
Currently, the IWE/IWL Program/Plan identifies 62 insulation panels at the interface between the concrete and the containment base mat (228-foot Elevation). Approximately one-third of the panels at the base mat interface will be removed and a general visual examination of the containment liner performed each examination period during the Second Ten-Year Interval. This will ensure that over the Second Ten-Year Interval, a 100% general visual examination of the liner at the base mat elevation will be performed. Attachment 6 provides details of that interface.
In addition, during the Second Ten-Year Interval, when an insulation panel at any elevation is removed for maintenance activities, a general visual examination of the liner beneath that panel will be performed.
Basis for Requesting Relief In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested for HBRSEP2, on the basis that the proposed alternative examinations/in conjunction with the examinations that have occurred during the First Ten-Year Interval, provide an acceptable level of quality and safety.
Table IWE-2500-1, Examination Category E-A does not address an insulated containment liner.
The containment liner at HBRSEP2, is partially insulated and covered by a stainless steel sheathing to provide for thermal protection of the liner during a design basis accident.
Justification for Granting Relief Relief is requested from the Code requirements for general visual examinations of the containment liner in areas that are insulated. Proposed alternative examinations provide an acceptable level of quality and safety.
The containment liner at HBRSEP2, is partially covered by insulation and stainless steel sheathing. The insulation and stainless steel sheathing form part of the defense-in-depth philosophy of the containment liner at HBRSEP2. The removal and reinstallation of the insulation sheathing panels has been determined to be time consuming and results in hardship and unusual difficulty.
During the First Ten-Year Interval, inspections were performed under 108 insulation panels that were removed. This included the planned removal of the entire inventory of 62 panels at the base mat elevation, for which both liner and moisture barrier inspections were performed. A total of 46 panels at higher elevations were removed for various reasons. Once removed, a liner inspection was performed.
A review of the data associated with the First Ten-Year Interval for the liner inspections indicated a degradation of the coating, which required coating removal and reapplication.
Subsequent ultrasonic and visual examination after coating removal revealed the minimum liner wall thickness was not violated and was acceptable to the procedural criteria. Liner coatings were reapplied and as-left examinations were performed prior to panel insulation and sheathing replacement.

Enclosure Page 30 of 60 The proposed alternative examinations provide an acceptable level of quality and safety.
Relief Request No. IWE/IWL-RR-02, Visual Examination of Moisture Barriers Code Requirements for Which Relief is Requested The ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Table IWE-2500-1, "Examination Categories," Examination Category E-A, Item Number E1.30, requires a visual examination of the containment moisture barrier, in accordance with Figure IWE-2500-1, "Examination Areas for Moisture Barriers," for Class MC, and metallic liners of Class CC, components.
The required method is a general visual examination of 100% of the moisture barrier materials intended to prevent intrusion of moisture against inaccessible areas of the pressure-retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces, which are not seal-welded. Containment moisture barrier materials include caulking, flashing, and other sealants used for this application. Deferral of the test to the end of the interval is not applicable due to the 100% per period requirement.
This relief was previously approved for the First Ten-Year IWE/IWL Inspection Interval under a Safety Evaluation Report (TAC No. MA4637) dated July 26, 1999.
Specific Relief Requested Relief is requested from performing general visual examinations, in accordance with ASME B&PV Code, 2001 Edition, 2003 Addenda, Section XI, Table IWE-2500-1, Examination Category E-A, on the containment moisture barriers. Proposed alternative examinations are provided below.
This request for relief is applicable to components classified as Class MC and subject to the requirements of Table IWE-2500-1, Examination Category E-A, at HBRSEP2.
Alternative Examination(s)
The IWE/IWL Program/Plan identifies 62 insulation panels at the interface between the concrete and the containment base mat (228-foot Elevation). Approximately one-third of the panels at the base mat interface will be removed and a general visual examination of the moisture barrier performed each examination period during the Second Ten-Year Interval. This will ensure that over the Second Ten-Year Interval, a 100% general visual examination of the moisture barrier will be performed.
Additionally, during the Second Ten-Year Interval, when an insulation panel on the 228-foot El.
is removed for maintenance activities, a general visual examination of the moisture barrier will be performed.
Basis for Requesting Relief In accordance with 10 CFR 50.5 5a(a)(3)(i), relief is requested for HBRSEP2, on the basis that the proposed alternative examinations, in conjunction with the examinations that have occurred during the First Ten-Year Interval, provide an acceptable level of quality and safety.
Table IWE-2500-1 Examination Category E-A does not address an insulated containment moisture barrier. The containment moisture barrier at HBRSEP2 is covered by stainless steel

Enclosure Page 31 of 60 sheathing and insulation to provide for thermal protection of the carbon steel liner during a design basis accident.
Justification for Granting Relief As shown in Figure IWE-2500-1, and noted in Table IWE-2500-1, moisture barrier materials are intended to prevent intrusion of moisture against inaccessible areas of the pressure-retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces which are not seal-welded. For HBRSEP2, the moisture barrier that meets this definition is the epoxy joint filler that interfaces with the concrete-to-containment liner interface at the 228-foot Elevation. Attachment 6 provides details of that interface.
The containment internal moisture barrier is covered with a layer of insulation and stainless steel sheathing. The removal and reinstallation of the insulation sheathing panels has been determined to result in hardship and unusual difficulty.
During the First Ten-Year Interval, 100% of the moisture barrier was inspected. A review of the data associated with the First Ten-Year Interval for the moisture barrier inspections indicated a degradation of the moisture barrier in some locations, which required removal and reapplication.
Visual examination of the liner after moisture barrier removal revealed that the minimum wall thickness of the liner behind the moisture barrier was not violated and was acceptable to the procedural criteria. Liner coatings and the moisture barrier were reapplied and as-left examinations were performed prior to panel insulation and sheathing replacement.
The proposed examination, which will ensure 100% moisture barrier inspection over the Second Ten-Year Interval, provides an acceptable level of quality and safety while not presenting an undue challenge to the moisture barrier insulation panels.
3.2.3   Results of recent IWE and IWL Examinations Recent IWE Examinations The results of recent IWE examinations is discussed in Section 3.4.1, IN 2010-12, Containment Liner Corrosion.
Recent IWL Examinations Examinations required by ASME B&PV Code, Section XI, Division 1, Subsection IWL (Inspection of concrete for the containment building) were performed the fall of 2013. The inspections were documented on "Visual examination of IWL (General)" Report Forms and had several newly reported adverse conditions. Per the requirements of ASME Section XI, a Registered Professional Engineer (RPE) with the ESG-0090N qualifications evaluated these conditions. The Items Identified have been screened by the RPE and are acceptable in that there are no operability concerns and the conditions do not affect the structural integrity or leak tightness of the HBRSEP2 containment.
VT-13-101, CONCRETE 0&deg; - 90&deg; DEGREES SURFACE - Reconfirmed previously reported condition from Report No. VT-10-001. No new conditions other than those previously identified and accepted.

Enclosure Page 32 of 60 VT-13-102, CONCRETE 90&deg; - 180&deg; DEGREES SURFACE - Between horizontal joints approximately 180 degrees, a horizontal rust material approximately 12" long, observed what appears to be a 1" board between forms 3rd and 4th from top 110 degrees, approximately 3" long partially covered by grout, and possibly an unpatched support hole at approximately 125 degrees four forms down from the top.
VT-13-103, CONCRETE 180&deg; - 270&deg; DEGREES SURFACE - Area at 230 degrees showed signs of further deterioration exposing underlying rebar.
VT-13-104, CONCRETE 270&deg; - 0&deg; DEGREES SURFACE - Concrete pop out with metal between 2nd and 3rd form from ground.
VT-13-107, CONCRETE DOME SURFACE - Reconfirmed previously reported condition from Report No. VT-10-007. No new conditions other than those previously identified and accepted.
R022 Disposition:
The recordable data from R020 has not changed. The separation at hooks #20 and 121 was minor with no active corrosion observed. The second hollow spot at hooks #29 and
       #30 was also at the dome/gutter interface. The concrete was not exposed and the depth of the hollow area appeared to be limited. Therefore, this 2nd hollow spot is also acceptable. The coatings have continued to degrade in limited areas on the dome. This potentially allows moisture to contact the concrete or exposes the concrete to the environment. In addition, another area of coating was missing adjacent to the dome ladder. Although the coating continues to degrade, corrosion or degradation of the concrete was not observed.

== Conclusion:==
No reinforcing steel was exposed and there was no evidence of active corrosion staining. These conditions do not affect the structural integrity of the containment structure and are therefore acceptable.
R024 Disposition The recordable data disposition from R022 has not changed except as noted below.
Coating at additional locations had deteriorated but the concrete at those locations did not identify any areas of concrete degradation. Although the amount of chipping on the concrete curb had increased, corrosion staining or rebar was not evident at these locations. This is acceptable. The chipping of the tie off hooks has no affect on the structural integrity of the containment structure. This is acceptable. It should be noted that several small areas of coating degradation on the normally inaccessible areas of the dome were evident. When viewed at a distance with binoculars, the exposed concrete in those areas did not exhibit any degradation.

== Conclusion:==
No reinforcing steel was exposed and there was no evidence of active corrosion staining. These conditions do not affect the structural Integrity of the containment structure and are therefore acceptable.
VT-13-109, CONCRETE CABLE VAULT ROOM NORTH SURFACE - Inspection revealed a pop out area approximately 3" x 7" with a depth of 5/8" and a crack-like Indication adjacent to both ends of the pop out extending 4" to 5" from pop out area.

Enclosure Page 33 of 60 VT-13-111, CONCRETE PURGE INLET ROOM SURFACE - Pop outs behind electrical equipment. Approximate size of pop out 5" x 3 1/4" x 1/2" deep and 2 1/2" x 2 1/4 x 14" deep.
VT-13-113, CONCRETE ROD CONTROL ROOM SURFACE - Reconfirmed previously reported condition from evaluation performed in 2007. No new conditions other than those previously identified and accepted.
R022 Disposition:
The recordable data from R020 has not changed. The 1 1/2 diameter hole did not appear to change from previous examination. The scaling documented in this examination was less than the recordable depth criteria with no active corrosion staining or rebar exposed.

== Conclusion:==
No reinforcing steel was exposed and there was no evidence of active corrosion staining. These conditions do not affect the structural integrity of the containment structure and are therefore acceptable.
R024 Disposition:
The recordable data disposition from R022 has not changed except as noted below. The horizontal portion of the foundation was included in this examination. The small voids meet the acceptance criteria.

== Conclusion:==
No reinforcing steel was exposed and there was no evidence of active corrosion staining. These conditions do not affect the structural integrity of the containment structure and are therefore acceptable.
The report forms above indicate either previously accepted items, original construction materials left from the initial concrete placement, unpatched support holes, minor rusting, and some pop outs. All are cosmetic in nature and will not affect the structural integrity or leak tightness of the containment. No action is required.
3.2.4    Integrated Leakage Rate Testing (ILRT) History Previous Type A tests confirmed that the HBRSEP2 reactor containment structure has leakage well under acceptance limits and represents minimal risk to increased leakage. Continued Type B and Type C testing for direct communication with containment atmosphere minimize this risk.
Also, the Inservice Inspection (IWE/IWL) program and maintenance rule monitoring provide confidence in containment integrity.
To date, seven operational Type A tests have been performed on HBRSEP2. There is considerable margin between these Type A test results and the TS 5.5.16 limit of 0.75 La, where La is equal to 0.1% by weight of the containment air per day at the peak accident pressure. These test results demonstrate that HBRSEP2 has a low leakage containment.
Table 3.2.4-1, ILRT Test Results Test Date              As Found Test Results             Results Adjusted to Pa
(% Weight per Day)                (% Weight per Day)
May 1974                       0.013 1                            0.013 2
Feb. 1978                      0.035                              0.049

Enclosure Page 34 of 60 Mar. 1982                      0.026 2                            0.037 Nov. 1984                      0.011 2                            0.016 April 1987                     0.041 2                            0.058 April 1992                   0.0602 1, 3                         0.0602 May 2007                     0.0244 1, 4                         0.0244 Note 1:         ILRT performed at Peak Containment Post LOCA pressure as identified in the plants TS in effect at the time of the test.
Note 2:         ILRT performed at 1/2 Pa and results are calculated at 1/2 Pa.
Note 3:         The test method used was the Absolute Method and the leakage rates were calculated using the Mass Point Analysis equations as described in ANSI/ANS 56.8-1987, "Containment System Leakage Testing Requirements.
Note 4:         The test method used was the Absolute Method and the leakage rates were calculated using the Mass Point Analysis equations as described in ANSI/ANS 56.8-1994, "Containment System Leakage Testing Requirements.
3.2.5   ILRT License Renewal Commitment To provide additional assurance of the tendons design capacity, testing at Integrated Leak Rate Test pressure, similar to the Structural Integrity Test performed in 1992, will be scheduled to coincide with Appendix J containment Integrated Leak Rate Testing conducted during the period of extended operation (required frequency in accordance with 10 CFR 50 Appendix J). The monitoring criteria for these tests will be limited to deformations and cracking associated with the vertical prestressed tendons, and will not include radial monitoring. Guidelines for performing the IWL examinations for these tests will include additional emphasis on looking for a pattern of horizontal cracks, and additional cracking in the discontinuity areas.
Procedures should include requirements for documentation and evaluation of test results showing structural integrity of the vertical prestressed tendons. Coordination with Civil Design group is required in order to ensure adequate controls are in place to complete the required test/inspections.
3.3     Containment Leakage Rate Testing Program, Type B and Type C Testing HBRSEP2 Types B and C testing program currently requires testing of electrical penetrations, airlocks, hatches, flanges, and containment isolation valves in accordance with 10 CFR Part 50, Appendix J, Option A. The results of the test program are used to demonstrate that proper maintenance and repairs are made on these components throughout their service life. The Types B and C testing program provides a means to protect the health and safety of plant personnel and the public by maintaining leakage from these components below appropriate limits. Per TS 5.5.16, the allowable maximum pathway total Types B and C leakage is 0.6 La where 0.6 La equals 91,490.78 sccm.
As discussed in NUREG-1493, Type B and Type C tests can identify the vast majority of all potential Containment leakage paths. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

Enclosure Page 35 of 60 As-Found Testing 10 CFR 50, Appendix J, Option A does not require As-found testing for Type B and Type C Penetrations.
Upon the implementation of the proposed amendments to HBRSEP2 Technical Specifications, As-Found LLRT testing will be required in accordance with the requirements of NEI 94-01 Revision 3-A, Section 10.2.1 for Type B Test Intervals, and Section 10.2.3 for Type C Test Interval.
Type B and C Acceptance Criteria 10 CFR 50, Appendix J, Option A, Acceptance Criterion.
The combined leakage rate for all penetrations and valves subject to Type B and C tests shall be less than 0.60 La. Leakage from containment isolation valves that are sealed with fluid from a seal system may be excluded when determining the combined leakage rate: Provided, That; (a) Such valves have been demonstrated to have fluid leakage rates that do not exceed those specified in the technical specifications or associated bases, and (b) The installed isolation valve seal-water system fluid inventory is sufficient to assure the sealing function for at least 30 days at a pressure of 1.10 Pa.
The IVSW system has been reviewed by the NRC and accepted as meeting the requirements of a seal system as defined in Appendix J. This review concluded that the IVSW system could be used in performing Type C tests. A description of the IVSW system is provided in Section 3.1.4.
Test results for valves that receive IVSW are not compared to the 0.6 La acceptance criteria of Appendix J Option A, Section III.C.3. Leakage from valves served by the IVSW system is excluded from this comparison. These test results are compared to the acceptance criteria of OST-933 and Technical Specification 3.6.8.6, which is based on a total allowable leakage value for the IVSW Tank. The allowable leakage is based on the UFSAR 6.8 criteria of 50-cc/hr./valve in.
Upon the implementation of the proposed amendments to HBRSEP2 Technical Specifications, the test results for valves that receive IVSW will continue to be exempt from the comparison to 0.6 La in accordance with NEI 94-01, Revision 3-A, Section 6.0, General Requirements, which provides the following relevant exemption:
An LLRT is not required for the following cases:
* Boundaries sealed with a qualified seal system Type B and Type C Test Results A review of the Type B and Type C test results from 2007 through 2015 for HBRSEP2 has shown an exceptional amount of margin between the actual As-Found (AF) and As-left (AL) outage summations and the regulatory requirements as described below:
* The As-Left leak rate average for HBRSEP2 shows an average of 15.3% of 0.6 La with a

Enclosure Page 36 of 60 high of 25.1% of 0.6 La.
Table 3.3.1-1 provides LLRT data summaries for HBRSEP2 since 2005 that encompasses the previous ILRT. This summary shows that there has been no failures that resulted in exceeding the TS 5.5.16 limit of 0.6 La (91,490.78 sccm) and demonstrates a history of successful tests.
The summations represent the high quality of maintenance of Type B and Type C tested components and the effective management of the Containment Leakage Rate Testing Program by the program owner.
Table 3.3-1, HBRSEP2 Types B and C LLRT As-Left Trend Summary RFO        RO-23       RO-24       RO-25     RO-26        RO-27        RO-28       RO-29 Fall       Spring       Fall   Summer        Spring         Fall       Spring 2005         2007       2008      2010         2012         2013         2015 As-Left      11985        18392      22972     16983        10436         9844         7122 sccm        sccm       sccm       sccm         sccm         sccm        sccm Fraction 0.131        0.201      0.251      0.186        0.114       0.108        0.079 of 0.6 La As shown in Table 3.3.1-1 above, the record keeping requirements for HBRSEP2 are different from other LARs requesting a permanent 15-year ILRT Interval with Containment Leakage Rate Testing Programs already on 10 CFR 50, Appendix J, Option B. 10 CFR 50, Appendix J, Option A and ANSI/ANS 56.8-1987 are not performance-based regulations and standards.
The recordkeeping requirements found in 10 CFR 50, Appendix J, Option A, Section V.B.2 are associated with the Type A test only and are as stated below:
For each periodic test, leakage test results from Type A, B, and C tests shall be included in the summary report. The summary report shall contain an analysis and interpretation of the Type A test results and a summary analysis of periodic Type B and Type C tests that were performed since the last type A test. Leakage test results from type A, B, and C tests that failed to meet the acceptance criteria of III.A.5(b), III.B.3, and III.C.3, respectively, shall be included in a separate accompanying summary report that includes an analysis and interpretation of the test data, the least squares fit analysis of the test data, the instrumentation error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria.
Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakage rate test measurements shall also be included.
The requirements regarding as-found and as-left, and minimum and maximum pathway leakage rates as normally reported in other LARs was not contained in ANSI/ANS 56.8-1987, hence they were also not reported in this LAR as which would have been reported in other LARs requesting a permanent 15-year ILRT Interval with Containment Leakage Rate Testing Programs already on 10 CFR 50 Appendix J, Option B. With the adoption of 10 CFR 50 Appendix J Option B, and NEI 94-01 Revision 3-A and the conditions and limitations of NEI 94-01 Revision 2-A as proposed in this LAR and, the recording/reporting of As-Found, As-Left, Minimum and Maximum pathway leakage rates as stated in ANSI/ANS 56.8-2002 will become a requirement of the HBRSEP2 Containment Leakage Rate Testing Program.
With the adoption of the proposed TS Amendment, the recordkeeping requirements will be changed to meet the following requirements:

Enclosure Page 37 of 60 10 CFR 50 Appendix J Option B, Section IV:
The results of the preoperational and periodic Type A, B, and C tests must be documented to show that performance criteria for leakage have been met. The comparison to previous results of the performance of the overall containment system and of individual components within it must be documented to show that the test intervals established for the containment system and components within it are adequate.
These records must be available for inspection at plant sites.
NEI 94-01 Revision 3-A, Section 12.1, 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 the report and a corrective action plan developed to restore the margin to an acceptable level.
Table 3.3-2 identifies the components that have not demonstrated acceptable performance during the previous two outages for HBRSEP2:
Table 3.3-2, HBRSEP2 Types B and C LLRT Program Implementation Review Penetration      Initial  Admin Limit   As-left   Cause    Corrective     Scheduled Number /        LLRT        SCCM       SCCM        of        Action       Interval Component        SCCM                             Failure 2013 RO-28 4            2800        2000        736     Note 1     Note 1      Appendix J Option A WD-1713 WD-1793 Primary System Vent Header 42           3000        3000         53     Note 2     Note 2      Appendix J Option A V12-12 V12-13 Containment Vacuum Relief 2015 RO-29 None

Enclosure Page 38 of 60 Note 1: While performing the initial LLRT on Penetration 4 (Primary System Vent Header) during RO-28, the leakage rate exceeded the Administrative Limit of 2000 sccm. The actual leak rate was 2800 sccm.
Carbon corrosion residue was discovered in check valve WD-1713 interior components after it had been dissembled. A light film of the carbon residue was present on the valves seat as well. The residue was likely to be from the carbon steel piping that runs downstream to the Nitrogen Supply Drain Tank Isolation Valve. The carbon corrosion residue was cleaned off the seat of the valve and the valves spring and disc were replaced.
Post maintenance LLRT Surveillance was performed with satisfactory results of 736 sccm.
The valve failed the same test in Refueling Outage-26. Based on its history of failures the past three refueling cycles, check valve WD-1713 will now be inspected every two refueling cycles in accordance with the corrective action program.
Note 2: While performing the initial LLRT on Penetration 42 (Containment Vacuum Relief) during RO-28, the leakage rate reached the Administrative Limit of 3000 sccm.
Containment vacuum relief isolation valves V12-12 and V12-13 were removed from the containment penetration piping to allow for routing communication cables into containment during the outage. The valves were removed and re-installed under work orders 2063378 (V12-12) and 2063365 (V12-13). The valves were not disassembled as part of this removal and re-installation.
After re-installation, on October 23, 2013, the valves were tested in accordance with EST-135-2, Local Leak Rate Test of Vacuum Relief Valves. The test pressure of 44 psig could not be obtained due to excessive leakage through V12-12. A visual inspection of the valve identified that the vane was not completely closed, leaving a gap to the valve seat-sealing ring.
The V12-12 seal ring was replaced and the travel stops were adjusted to ensure proper closure in accordance with maintenance procedure CM-M-GNRL-VLV-0001. Post maintenance LLRT Surveillance was performed with satisfactory results of 53 sccm.
3.4     NRC Information Notices (INs) 3.4.1   IN 2010-12, Containment Liner Corrosion This IN provides examples of containment liner degradation caused by corrosion. Concrete reactor containments are typically lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions. The reactor containment is required to be operable as specified in plant technical specifications to limit the leakage of fission product radioactivity from the containment to the environment. The regulations at 10 CFR 50.55a, Codes and Standards, require the use of Subsection IWE of ASME Section XI to perform inservice inspections of containment components. The required inservice inspections include periodic visual examinations and limited volumetric examinations using ultrasonic thickness measurements. The containment components include the steel containment liner and integral

Enclosure Page 39 of 60 attachments for the concrete containment, containment personnel airlock and equipment hatch, penetration sleeves, moisture barriers, and pressure-retaining bolting. The NRC also requires licensees to perform leak rate testing of the containment pressure-retaining components and isolation valves according to 10 CFR Part 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, as specified in plant technical specifications. This operating experience highlights the importance of good quality assurance, housekeeping and high quality construction practices during construction operations in accordance with 10 CFR Part 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
Corrosion to the containment liner is not a new industry issue. Programs and procedures are in-place to inspect the containment liner and would identify any areas subject to corrosion.
Containment Liner Corrosion Operating Experience Summary, Technical Letter Report Revision 1, dated August 2, 2011, Section 2.4, Previous Assessments and Containment Operating Experience provided the following descriptions of previously identified containment liner and moisture barrier degradation for HBRSEP2:
* Robinson Unit 2 and Beaver Valley Unit 1 had areas of bulging and spot corrosion of the liner plate and degradation of the liner coatings. Surface cracking of concrete and deterioration of earlier patched concrete was observed (Ashar and Bagchi, 1995).
* In April 1992, Robinson Unit 2 was observed to have discoloration of the vertical portion of the containment liner indicating possible corrosion of the liner at an insulation joint (Ashar and Bagchi, 1995). This was also addressed in Information Notice 97-19, Liner Plate Corrosion in Concrete Containments.
* In December 1996, H.B. Robinson Unit 2 was found to have degraded caulking and insulation sheathing panels during a containment walkdown. The vertical portion of the containment liner at Robinson is protected by a polyvinyl chloride insulation material and a metal sheathing material. The licensee determined that a portion of this insulation sheathing material was loose and that some of the caulking between the sheathing panels was deteriorated. After examination during subsequent refueling outages, they determined that the protective coating for the containment liner was degraded and that while some corrosion of the containment liner had occurred, the liner still met design requirements (Braverman, et al., 2000).
* During the First Ten-Year Interval, 100% of the moisture barrier was inspected. A review of the data associated with the First Ten-Year Interval for the moisture barrier inspections indicated a degradation of the moisture barrier in some locations, which required removal and reapplication. Visual examination of the liner after moisture barrier removal revealed that the minimum wall thickness of the liner behind the moisture barrier was not violated and was acceptable to the procedural criteria. Liner coatings and the moisture barrier were reapplied and as-left examinations were performed prior to panel insulation and sheathing replacement.
* During the First Ten-Year Interval, inspections were performed under 108 insulation panels that were removed. This included the planned removal of the entire inventory of 62 panels at the base mat elevation, for which both liner and moisture barrier inspections

Enclosure Page 40 of 60 were performed. A total of 46 panels at higher elevations were removed for various reasons. Once removed, a liner inspection was performed.
A review of the data associated with the First Ten-Year Interval for the liner inspections indicated a degradation of the coating, which required coating removal and reapplication.
Subsequent ultrasonic and visual examination after coating removal revealed the minimum liner wall thickness was not violated and was acceptable to the procedural criteria. Liner coatings were reapplied and as-left examinations were performed prior to panel insulation and sheathing replacement.
The results of recent Second Ten-Year Interval inspections are described as follows:
* IWE Inspections of the CV dome produced a visual examination report of degraded coatings of the dome liner. CV dome degraded coatings documented by the IWE Inspections during R028 confirm degraded coatings found during coatings assessment walkdowns of previous outages. A minor increase in the amount of degraded coatings was identified. Most all of the degraded areas observed in the IWE Inspection of R028 have been previously documented by the coatings program. All degraded coatings on the dome liner must be documented and Incorporated Into the coatings exempt log to monitor the containment exempt coatings margin, and all coatings from the IWE Inspection not already contained In the coatings exempt log will be added.
The coatings program will continue to monitor and trend the containment dome degraded coatings and will Incorporate IWE Inspection data into the coatings exempt log.
* IWE Inspections of the CV wall liner during RO-26, RO-27 and RO-28 produced visual examination reports of degraded coatings, excessive corrosion/pitting, broken pins, gouging, bulging, and abnormal wear/discoloration of the wall liner. All degraded coating were restored. All VT-3 indications were dispositioned by evaluation. No base metal repairs were required.
3.4.2  IN 2014-07, Degradation of Leak Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner This IN was issued to inform addressees of issues identified by the NRC staff concerning degradation of floor weld leak-chase channel systems of steel containment shell and concrete containment metallic liner that could affect leak-tightness and aging management of containment structures.
IN 2014-07 described the leak chase channel system as follows:
Consists of steel channel sections that are fillet welded continuously over the entire bottom shell or liner seam welds and subdivided into zones, each zone with a test connection. Each test connection consists of a small carbon or stainless steel tube (less than 1-inch (2.5 centimeters) diameter) that penetrates through the back of the channel and is seal-welded to the channel steel. The tube extends up through the concrete floor slab to a small steel access (junction) box embedded in the floor slab. The steel tube, which may be encased in a pipe, projects up through the bottom of the access box with a threaded coupling connection welded to the top of the tube, allowing for pressurization

Enclosure Page 41 of 60 of the leak-chase channel.
IN 2014-07 describes a recessed box with a cover plate at floor level that allows for water to pool inside the recessed box and cause degradation. The HBRSEP2 system is not the same as the cited systems since it uses manifolds 3.5 feet above the floor instead of a capped floor access box. The manifolds are vented to containment atmosphere, however they are sealed passages and not conducive to flow to transport moisture.
The general review of NRC IN 2014-07 did identify a gap in the IWE program as described below:
The piping/tubing runs that go through the concrete have not been previously included in the IWE inspection program to verify that there is no corrosion or a breach near the floor level which would allow moisture to enter in a manner like the plants described in the IN.
The PPS to the Weld Channels was abandoned in the mid 1970s and per the 10 CFR 50.55(b)(2)(ix)(A) requirement; the piping/tubing should be examined closely enough to say that there is not a breach.
To provide reasonable assurance that aging effects of the containment liner are being managed, the IWE program was revised to perform visual inspections of accessible tubing in the PPS system from manifold to floor of the CV.
3.4.3    IN 92-20, Inadequate Local Leak Rate Testing NRC IN 92-20 was issued to alert licensees to problems with local leak rate testing of two-ply stainless steel bellows used on piping penetrations at some plants. Specifically, local leak rate testing could not be relied upon to accurately measure the leakage rate that would occur under accident conditions since, during testing, the two plies in the bellows were in contact with each other, restricting the flow of the test medium to the crack locations. Any two-ply bellows of similar construction may be susceptible to this problem.
This is not applicable to HBRSEP2. There are both single and two-ply bellows assemblies, which are also Containment isolation barriers, installed at HBRSEP2. The difference is that the bellows assemblies are tested using the original Penetration Pressurization System (PPS).
ESR/MOD 95-00888 removed the automatic continuous pressurization and monitoring features of this system. It is now only used during power operation to test the personnel hatch and during outages to test containment penetrations (local leak rate tests) such as the penetration bellows assemblies.
The PPS test connection is located on the outboard penetration sleeve for each bellows assembly and not between the plies on the two-ply assemblies so there is no restriction in airflow between the inboard and outboard bellows assemblies.
The leakage rate performance of the bellows assemblies subject to Type B testing is shown in Table 3.4.3-1 below:

Enclosure Page 42 of 60 Table 3.4.3-1, HBRSEP2 Type B Bellows As-Left Trend Summary RO-23         RO-24       RO-25      RO-26       RO-27      RO-28       RO-29 Fall        Spring        Fall    Summer       Spring       Fall     Summer 2005          2007        2008       2010        2012       2013        2015 0        1355.46     3002.01    2723.21     2478.37      2054        1687 sccm         sccm        sccm        sccm        sccm       sccm       sccm 3.5      Supplemental Inspections In the Safety Evaluation Report for NEI 94-01 Revision 2-A, the NRC stated the following requirement for the performance of Supplemental Visual Inspections in SER Section 3.1.1.3, Adequacy of Pre-Test Inspections (Visual Examinations):
Subsections IWE and IWL (References 13 and 14) of the ASME Code, Section XI, as incorporated by reference in 10 CFR 50.55a, require general visual examinations two times within a 10-year interval for concrete components (Subsection IWL), and three times within a 10-year interval for steel components (Subsection IWE). To avoid duplication or deletion of examinations, licensees using NEI TR 94-01, Revision 2, have to develop a schedule for containment inspections that satisfy the provisions of Section 9.2.3.2 of this TR and ASME Code, Section XI, Subsection IWE and IWL requirements.
The performance of a supplemental inspection is a requirement of the HBRSEP2 Containment Integrated Leak Test surveillance procedure. A pretest visual inspection of accessible interior and exterior surfaces of Containment structures and components for evidence of deterioration is completed and documented in the surveillance. The purpose of this inspection is to identify evidence of structural deterioration that might affect either the primary containment structural integrity or leak lightness. Any irregularities such as cracking, peeling, delamination, corrosion, and structural deterioration shall be recorded and evaluated, or repaired as required.
This inspection requirement may be met in part OR its entirety by completing the containment building IWE and IWL inspections OR by walking down the Containment per the instructions contained in the Containment Integrated Leak Test surveillance procedure.
3.6      Limitations and Conditions 3.6.1    Limitations and Conditions Applicable to NEI 94-01, Revision 2-A The NRC staff found that the use of NEI TR 94-01, Revision 2, was acceptable for referencing by licensees proposing to amend their TS to permanently extend the ILRT surveillance interval to 15 years, provided the following conditions as listed in Table 3.6-1 were satisfied:
Table 3.6-1, NEI 94-01 Revision 2-A Limitations and Conditions

Enclosure Page 43 of 60 Limitation/Condition (From Section 4.0 of SE)                        HBRSEP2 Response For calculating the Type A leakage rate, the    HBRSEP2 will utilize the definition in licensee should use the definition in the NEI   NEI 94-01 Revision 2-A, Section 5.0.
TR 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002. (Refer to SE                Reference Table 3.2.3-1, ILRT Test Section 3.1.1.1.)                               Results, Notes (3) and (4)
The licensee submits a schedule of              Reference Tables 3.2.2-1, 3.2.2-2 and containment inspections to be performed prior   3.5-1 of this submittal.
to and between Type A tests. (Refer to SE Section 3.1.1.3.)
The licensee addresses the areas of the         Reference Section 3.2.2, Inaccessible containment structure potentially subjected to  Class MC Areas, of this submittal.
degradation. (Refer to SE Section 3.1.3.)
Reference Section 3.2.2, IWE/IWL augmented Examinations and Attachment 5, E-C AUGMENT Plan Tables for the Second Ten-Year Interval.
The licensee addresses any tests and            HBRSEP2 steam generator inspections performed following major           replacements were completed in 1984 modifications to the containment structure, as  using the equipment hatch.
applicable. (Refer to SE Section 3.1.4.)
There are no planned modifications for HBRSEP2 that will require a Type A test prior to the next scheduled Type A test proposed under this LAR.
There is no anticipated addition or removal of plant hardware within the containment building, which could affect its leak-tightness.
The normal Type A test interval should be less  HBRSEP2 will follow the requirements than 15 years. If a licensee has to utilize the of NEI 94-01 Revision 2-A, Section 9.1.
provision of Section 9.1 of NEI TR 94-01,       In accordance with the requirements of Revision 2, related to extending the ILRT       94-01 Revision 2-A, SER Section interval beyond 15 years, the licensee must     3.1.1.2, HBRSEP2 will also demonstrate demonstrate to the NRC staff that it is an      to the NRC staff that an unforeseen unforeseen emergent condition. (Refer to SE     emergent condition exists in the event Section 3.1.1.2.)                               an extension beyond the 15-year interval is required.

Enclosure Page 44 of 60 For plants licensed under 10 CFR Part 52,           Not applicable. HBRSEP2 was not applications requesting a permanent extension       licensed under 10 CFR Part 52.
of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.
3.6.2    Limitations and Conditions Applicable to NEI 94-01 Revision 3-A The NRC staff found that the guidance in NEI TR 94-01, Revision 3, was acceptable for referencing by licensees in the implementation for the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J. However, the NRC staff identified two conditions on the use of NEI TR 94-01, Revision 3 (Reference NEI 94-01 Revision 3-A, NRC SER 4.0, Limitations and Conditions):
Topical Report Condition 1 NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs be increased to 75 months with the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84-months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI TR 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR MSIVs), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months.
Response to Condition 1 Condition 1 presents three (3) separate issues that are required to be addressed as follows:
* ISSUE 1 - The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.
* ISSUE 2 - In addition, a corrective action plan shall be developed to restore the margin to an acceptable level.
* ISSUE 3 - Use of the allowed 9-month extension for eligible Type C valves is only authorized for non-routine emergent conditions.

Enclosure Page 45 of 60 Response to Condition 1, Issue 1 The post-outage report shall include the margin between the Type B and Type C Minimum Pathway Leak Rate (MNPLR) summation value, as adjusted to include the estimate of applicable Type C leakage understatement, and its regulatory limit of 0.60 La.
Response to Condition 1, Issue 2 When the potential leakage understatement adjusted Types B and C MNPLR total is greater than the HBRSEP2 administrative leakage summation limit of 0.50 La, but less than the regulatory limit of 0.6 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the HBRSEP2 administrative limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues so as to maintain an acceptable level of margin.
Response to Condition 1, Issue 3 HBRSEP2 will apply the 9-month grace period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.
Topical Report Condition 2 The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time. The containment leakage condition-monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending overall containment leakage potential, the as-found minimum pathway leakage rates for the just tested penetrations are summed with the as-left minimum pathway leakage rates for penetrations tested during the previous 1 or 2 or even 3 refueling outages. Type C tests involve valves, which in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable. Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for.
Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.
When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total, and must be included in a

Enclosure Page 46 of 60 licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.
Response to Condition 2 Condition 2 presents two (2) separate issues that are required to be addressed as follows:
* ISSUE 1 - Extending the LLRT intervals beyond 5 years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI TR 94-01, Revision 3, Section 12.1.
* ISSUE 2 - When routinely scheduling any LLRT valve interval beyond 60-months and up to 75-months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B and C total, and must be included in a licensee's post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.
Response to Condition 2, Issue 1 The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25%
in the LLRT periodicity. As such, HBRSEP2 will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the actual As-Left leak rate, which will increase the As-Left leakage total 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, if the MNPLR is greater than the HBRSEP2 administrative 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 HBRSEP2 administrative leakage limit. The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.
Response to Condition 2, Issue 2 If the potential leakage understatement adjusted leak rate MNPLR is less than the HBRSEP2 administrative leakage summation limit of 0.50 La, then the acceptability of the greater than a 60 month test interval up to the 75-month LLRT extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

Enclosure Page 47 of 60 In addition to Condition 1, Parts 1 and 2, which deal with the MNPLR Type B and C summation margin, NEI 94-01, Revision 3-A also has a margin related requirement as contained in Section 12.1, Report Requirements.
A post-outage report shall be prepared presenting results of the previous 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.
At HBRSEP2, in the event an adverse trend in the aforementioned potential leakage understatement adjusted Type B and C summation is identified, and then an analysis and determination of a corrective action plan shall be prepared to restore the trend and associated margin to an acceptable level. The corrective action plan shall focus on those components which have contributed the most to the adverse trend in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.
At HBRSEP2, an adverse trend is defined as three (3) consecutive increases in the final pre-RCS Mode Change Type B and C MNPLR leakage summation values, as adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.
3.7     Evaluation of Risk Impact 3.7.1   Methodology 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 HBRSEP2. 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 RG 1.200 as applied to ILRT interval extensions, and risk insights in support of a request for a plants licensing basis as outlined in 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 1009325, Revision 2-A of EPRI 1018243
[Reference 24].
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, is acceptable for referencing by licensees proposing to amend their TS to extend the ILRT surveillance interval to 15 years, subject to the

Enclosure Page 48 of 60 limitations and conditions noted in Section 4.0 of the SE. Table 3.7.1-1 addresses each of the four limitations and conditions for the use of EPRI 1009325, Revision 2.
Table 3.7.1-1, EPRI Report No.TR-1009325 Revision 2 Limitations and Conditions Limitation/Condition                          HBRSEP2 Response (From Section 4.2 of SE) 1.The licensee submits documentation          The technical adequacy of the HBRSEP2 indicating that the technical adequacy      PRA is consistent with the requirements of of their PRA is consistent with the         Regulatory Guide 1.200 as is relevant to requirements of RG 1.200 relevant to        this ILRT interval extension, as detailed in the ILRT extension.                         Attachment 4 of this submittal, PRA Risk Assessment for Extending ILRT Interval to 15 Years, Attachment 1.
: 2. The licensee submits documentation         EPRI Report No. 1009325, Revision 2-A, indicating that the estimated risk          incorporates these population dose and increase associated with permanently        CCFP acceptance guidelines, and these extending the ILRT surveillance             guidelines have been used for the interval to 15 years is small, and          HBRSEP2 plant specific assessments.
consistent with the clarification provided in Section 3.2.4.5 of this SE.
Specifically, a small increase in The increase in population dose is 0.020 population dose should be defined as person-rem/year.
an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive.
In addition, a small increase in CCFP      The increase in CCFP is 0.829%. The should be defined as a value               increase proved to be below 1.5 marginally greater than that accepted      percentage points and thus is considered in a previous one-time ILRT extension      to be small.
requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point.

Enclosure Page 49 of 60
: 3. The methodology in EPRI Report No.          EPRI Report No. 1009325, Revision 2-A, 1009325, Revision 2, is acceptable           incorporated the use of 100 La as the except for the calculation of the            average leak rate for the pre-existing increase in expected population dose         containment large leakage rate accident (per year of reactor operation). In order    case (accident case 3b), and this value to make the methodology acceptable,          has been used in the HBRSEP2 plant the average leak rate accident case          specific risk assessment.
(accident case 3b) used by the licensees shall be 100 La instead of 35 La
: 4. A licensee amendment request (LAR)          For HBRSEP2, containment over-pressure is required in instances where               is NOT relied upon for emergency core containment over-pressure is relied          cooling system (ECCS) performance.
upon for emergency core cooling system (ECCS) performance.
3.7.2    Summary of Internal Events PRA Quality Statement for Permanent 15-Year ILRT Extension Internal Events PRA Model The HBRSEP2 Internal Events PRA model RNP_12 is utilized to calculate CDF, LERF, and PDSs for the permanent 15-year ILRT extension. Any elements of the supporting requirements detailed in ASME/ANS RA-Sa-2009 that could be significantly affected by the application are required to meet Capability Category II requirements.
The Internal Events PRA provides an adequate base model for the development of the permanent 15-year ILRT extension. While there has been one instance of a change to the internal events model since the 2010 peer review that could be considered a model upgrade, this change is specifically applicable to the loss of component cooling water (CCW) initiating event and does not affect the permanent 15-year ILRT extension results.
To ensure that the current Internal Events PRA model remains an accurate reflection of the as-built, as-operated plant, the following configuration control activities are routinely performed:
* Design changes and procedure changes are reviewed for their impact on the Internal Events PRA model. PRA screening is required for all design and procedure changes.
* New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.
* Plant-specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated based upon reviews of plant program data, particularly data supporting the Maintenance Rule.

Enclosure Page 50 of 60 The HBRSEP2 Internal Events PRA is based on a detailed model of the plant developed from the Individual Plant Examination for Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities." The model is maintained and updated in accordance with HBRSEP2 procedures, and has been updated to meet the ASME PRA Standard and Revision 2 of RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities."
In accordance with RG 1.200, Rev. 2 and ASME/ANS RA-Sa-2009, a full scope Peer Review of the Internal Events model of the HBRSEP2 PRA was conducted in May 2010 by six PRA experts (Attachment U of Reference 29). The review provided findings and suggestions regarding the model and identified 25 supporting requirements within the internal events portion of the model that did not meet Capability Category II. These findings were either resolved by additional analysis and included in the quantitative results, or evaluated with their impact on the applicable risk evaluation. The dispositions for these findings are presented in Table 1, Internal Events PRA Peer Review - Facts and Observations. All findings, which had significant impact on this analysis, have been addressed. The ILRT application was determined to be an application requiring a Capability Category II PRA model per the RG 1.200 criteria, Revision 2.
This is based on the requirement for numerical results for CDF and LERF to determine the risk impact of the requested change and the fact that this change is risk-informed, not risk based. of this submittal, PRA Risk Assessment for Extending ILRT Interval to 15 Years, , Table 1 includes discussion of all findings from the industry peer review along with the assessment and evaluation of the finding that shows that they have either been addressed or have no material impact on the ILRT interval extension request.
Following disposition, all of the SRs evaluated Met Capability Category II or better.
Fire PRA Quality Statement for Permanent 15-Year ILRT Extension In accordance with RG 1.205 position 4.3: The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard supporting requirements important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable.
The HBRSEP2 Internal Events model is also updated to support the HBRSEP2 Fire PRA. The HBRSEP2 Combined Internal Events and Fire PRA was peer reviewed during the period of March 2013 (Attachment V of Reference 29). The peer review was conducted by a team of industry personnel (utility and vendor). The Westinghouse Owners Group performed the review and has documented the outcome via LTR-RAM-13-06, Fire PRA Peer Review of the H. B.
Robinson Nuclear Plant Fire Probabilistic Risk Assessment against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard. As noted in LTR-RAM-13-06, the HBRSEP2 Fire PRA was found to be consistent with the ASME/ANS Standard and is suitable for supporting risk-informed applications.
The peer review team noted a number of Facts and Observations (F&Os). As documented in LTR-RAM-13-06, 85% of the Supporting Requirements (SRs) were assessed at Capability Category II or higher. Approximately eighteen Finding level and nine Suggestion level F&Os were identified during the peer review conducted in March 2013. Duke Energy recognized that CDF and LERF were relatively high, as noted in LTR-RAM-13-06. Based on the CDF and LERF values at the time of the initial peer review, coupled with the number of findings associated with

Enclosure Page 51 of 60 the Fire Scenario Selection (FSS) Technical Element (18), Duke Energy decided to have a focused peer review.
The focused peer review was conducted during the period of July 2013 and evaluated the FSS Technical Element based on refinements to approved methodologies and updated documentation. The focused peer review was conducted by Frederick Mowrer (C P Fire, LLC) and Bijan Najafi (JENSEN HUGHES, Inc.) and is documented via JENSEN HUGHES Calculation No. 0004-0042-415-RPT-001, Robinson Nuclear Plant Fire PRA Focused Peer Review, Revision 0. As noted in LTR-RAM-13-06 and JENSEN HUGHES Calculation No. 0004-0042-415-RPT-001, the Fire PRA does apply the methodologies outlined in NUREG/CR-6850 correctly, is consistent with the ASME/ANS Standard and is applicable for supporting risk informed applications. Although several of the initial F&Os were resolved, seven new findings and three new suggestions were identified during the focused-scope peer review. of this submittal, PRA Risk Assessment for Extending ILRT Interval to 15 Years, , Table 2 documents the Finding level F&Os associated with both the initial and focused peer reviews. All but two findings have been dispositioned as meeting Capability Category II or better. FSS-E3-01, which was assessed as meeting Category I, involved providing a mean value of, and statistical representation of, the uncertainty intervals for the parameters used for modeling the significant fire scenarios. Although no change has yet been made that would improve the Capability Category assessment, HBRSEP2 considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because documenting the statistical representation of uncertainty intervals will not change the quantified risk metrics.
Supporting Requirement FSS-E1-01 has not been resolved, but this is a documentation issue that does not impact quantification. These two SRs not being resolved does not impact the ILRT extension analysis. The Fire PRA is adequate to support the ILRT extension.
3.7.3   Summary of Plant-Specific Risk Assessment Results The risk impact of permanently extending the Type A ILRT test frequency to once in fifteen years is as follows:
* RG.174 (Reference 4) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of Core Damage Frequency (CDF) 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 Large Early Release Frequency (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 9.36E-8/year using the EPRI guidance. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174.
* 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 0.020 person-rem/year. EPRI Report No. 1009325, Revision 2-A (Reference 20) 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

Enclosure Page 52 of 60 when compared to other severe accident risks is negligible.
* The increase in the conditional containment failure from the 3-in-10-year interval to 1-in-15-year interval is 0.545%. EPRI Report No. 1009325, Revision 2-A (Reference 20) states that increases in Conditional Containment Failure Probability (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 HBRSEP2 risk profile.
3.7.4   Previous Assessments The NRC in NUREG-1493 (Reference 7) has previously concluded that:
* Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.
* Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.
The findings for HBRSEP2 confirm these general findings on a plant-specific basis considering the severe accidents evaluated for HBRSEP2, the HBRSEP2 containment failure modes, and the local population surrounding HBRSEP2.
Details of the HBRSEP2 risk assessment are contained in Attachment 4 of this submittal.

==4.0     REGULATORY EVALUATION==

4.1     Applicable Regulatory Requirements/ Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. 10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR Part 50, "Leakage Rate Testing of Containment of Water Cooled Nuclear Power Plants." Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.
The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B and Type C testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage tests must be performed. Under the performance-based option of 10 CFR

Enclosure Page 53 of 60 Part 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The change to the Type A test frequency did not directly result in an increase in containment leakage. Similarly, the proposed change to the Type B and Type C test frequencies will not directly result in an increase in containment leakage.
EPRI TR-1009325, Revision 2, provided a risk impact assessment for optimized ILRT intervals up to 15 years, utilizing current industry performance data and risk informed guidance. NEI 94 01, Revision 3-A, Section 9.2.3.1 states that Type A ILRT intervals of up to 15 years are allowed by this guideline. The Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI Report 1018243 (Formerly TR-1009325, Revision 2) indicates that, in general, the risk impact associated with ILRT interval extensions for intervals up to 15 years is small.
However, plant-specific confirmatory analyses are required.
The NRC staff reviewed NEI TR 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2.
For NEI TR 94-01, Revision 2, the NRC staff determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J. This guidance includes provisions for extending Type A ILRT intervals to up to 15 years and incorporates the regulatory positions stated in RG 1.163. The NRC staff finds that the Type A testing methodology as described in ANSI/ ANS-56.8-2002, and the modified testing frequencies recommended by NEI TR 94- 01, Revision 2, serves to ensure continued leakage integrity of the containment structure. Type B and Type C testing ensures that individual penetrations are essentially leak tight.
In addition, aggregate Type B and Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths.
For EPRI Report No. 1009325, Revision 2, a risk-informed methodology using plant specific risk insights and industry ILRT performance data to revise ILRT surveillance frequencies, the NRC staff finds that the proposed methodology satisfies the key principles of risk-informed decision making applied to changes to TSs as delineated in RG 1.177 and RG 1.174. The NRC staff, therefore, found that this guidance was acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing, subject to the limitations and conditions noted in Section 4.2 of the Safety Evaluation Report (SER).
The NRC staff reviewed NEI TR 94-01, Revision 3, and determined that it described an acceptable approach for implementing the optional performance-based requirements of Option B to 10 CFR Part 50, Appendix J, as modified by the conditions and limitations summarized in Section 4.0 of the associated Safety Evaluation. This guidance included provisions for extending Type C LLRT intervals up to 75 months. Type C testing ensures that individual containment isolation valves are essentially leak tight. In addition, aggregate Type C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The NRC staff, therefore, found that this guidance, as modified to include two limitations and conditions, was acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing. Any applicant may reference NEI TR 94-01, Revision 3, as modified by the associated SER and approved by the NRC, and the conditions and limitations specified in NEI 94- 01, Revision 2-A, dated October 2008, in a licensing action to satisfy the requirements of Option B to 10 CFR Part 50, Appendix J.
4.2      Precedent

Enclosure Page 54 of 60 This LAR is similar in nature to the following license amendments previously authorized by the NRC to extend the Type A test frequency to 15 years and the Type C test frequency to 75 months:
* Surry Power Station, Units 1 and 2 (Reference 21)
* Donald C. Cook Nuclear Plant, Units 1 and 2 (Reference 22)
* Beaver Valley Power Station, Unit Nos. 1 and 2 (Reference 23)
* Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 (Reference 24)
* Peach Bottom Atomic Power Station, Units 2 and 3 (Reference 27)
This license amendment request is similar in nature to the following license amendment previously authorized by the NRC to adopt TSTF Technical Change Traveler 52, Revision 3, to implement Option B of 10 CFR 50 Appendix J:
* Oconee Nuclear Station, Units 1, 2, and 3 (Reference 25) 4.3     Significant Hazards Consideration Duke Energy Progress has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of amendment, as discussed below:
: 1.      Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment to the Technical Specifications (TS) involves the extension of the H. B. Robinson Steam Electric Plant Unit No. 2 (HBRSEP2) Type A containment test interval to 15 years, the extension of the Type B test intervals to 120 months for selected components, and the extension of the Type C test interval to 75 months for selected components. The current Type A test interval of 120 months (10 years) would be extended on a permanent basis to no longer than 15 years from the last Type A test.
The current Type B test interval of each reactor shutdown for refueling but in no case at intervals greater than 2 years would be extended on a performance basis to no longer than 120 months. The current Type C test interval of each reactor shutdown for refueling but in no case at intervals greater than 2 years would be extended on a performance basis to no longer than 75 months. Extensions of up to nine months (total maximum interval of 84 months for Type C tests) are permissible only for non-routine emergent conditions. The proposed extensions do not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. The containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident. The change in dose risk for changing the Type A test frequency from three-per-ten years to once-per-fifteen years, measured, as an increase

Enclosure Page 55 of 60 to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.020 person-rem/year. The Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2-A, states that a very small population dose is defined as an increase of  1.0 person-rem per year, or  1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended integrated leak rate test (ILRT) intervals. Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.
As documented in NUREG-1493, Type B and C tests have identified a very large percentage of containment leakage paths, and the percentage of containment leakage paths that are detected only by Type A testing is very small. The HBRSEP2 Type A test history supports this conclusion.
The integrity of the containment is subject to two types of failure mechanisms that can be categorized as: (1) activity based, and (2) time based. Activity based failure mechanisms are defined as degradation due to system and/or component modifications or maintenance. Local leak rate test requirements and administrative controls such as configuration management and procedural requirements for system restoration ensure that containment integrity is not degraded by plant modifications or maintenance activities. The design and construction requirements of the containment combined with the containment inspections performed in accordance with the American Society of Mechanical Engineers (ASME) Section XI, the Maintenance Rule, and TS requirements serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by a Type A test. Based on the above, the proposed extensions do not significantly increase the consequences of an accident previously evaluated.
The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.
Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment to the TS involves the extension of the HBRSEP2 Type A containment test interval to 15 years, the Type B test interval to 120 months for selected components and the extension of the Type C test interval to 75 months for selected components. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident do not involve any accident precursors or initiators. The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

Enclosure Page 56 of 60 The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
: 3.      Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The proposed amendment to TS 5.5.16 involves the extension of the HBRSEP2 Type A containment test interval to 15 years, the Type B test interval to 120 months for selected components and the extension of the Type C test interval to 75 months for selected components. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS Containment Leak Rate Testing Program exist to ensure that the degree of containment structural integrity and leak tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit specified by TS is maintained.
The proposed change involves only the extension of the interval between Type A containment leak rate tests, Type B tests and Type C tests for HBRSEP2. The proposed surveillance interval extension is bounded by the 15-year ILRT interval, the 120-month Type B interval and the 75-month Type C test interval currently authorized within NEI 94-01, Revision 3-A. Industry experience supports the conclusion that Types B and C testing detects a large percentage of containment leakage paths and that the percentage of containment leakage paths that are detected only by Type A testing is small. The containment inspections performed in accordance with ASME Section XI, TS and the Maintenance Rule serve to provide a high degree of assurance that the containment would not degrade in a manner that is detectable only by Type A testing.
The combination of these factors ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed change, since these are not affected by changes to the Type A, Type B and Type C test intervals.
The proposed amendment also deletes an exception previously granted to allow one-time extension of the ILRT test frequency for HBRSEP2. This exception was for an activity that has already taken place so the deletion is solely an administrative action that has no effect on any component and no impact on how the unit is operated.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, Duke Energy Progress concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c),
and, accordingly, a finding of no significant hazards consideration is justified.

Enclosure Page 57 of 60 4.4      Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.
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, describe an NRC-accepted approach for implementing the performance-based requirements of 10 CFR Part 50, Appendix J, Option B. It incorporated the regulatory positions stated in RG 1.163 and includes provisions for extending Type A intervals to 15 years, Type B test intervals to 120 months, and Type C test intervals to 75 months. NEI 94-01, Revision 3-A delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies. HBRSEP2 is adopting the guidance of NEI 94-01, Revision 3-A, and the conditions and limitations specified in NEI 94-01, Revision 2-A, for the HBRSEP2, 10 CFR Part 50, Appendix J Testing Program Plan.
Based on the previous ILRT tests conducted at HBRSEP2, it may be concluded that the permanent extension of the containment ILRT interval from 10 to 15 years represents minimal risk to increased leakage. The risk is minimized by continued Type B and Type C testing performed in accordance with Option B of 10 CFR Part 50, Appendix J and the overlapping inspection activities performed as part of the following HBRSEP2 inspection programs:
* Containment Inservice Inspection Program (IWE/IWL)
* Primary Containment Coatings Condition Assessment This experience is supplemented by risk analysis studies, including the HBRSEP2 risk analysis provided in Attachment 4. The findings of the risk assessment confirm the general findings of previous studies, on a plant-specific basis, that extending the ILRT interval from 10 to 15 years results in a very small change to the HBRSEP2 risk profiles.

==5.0      ENVIRONMENTAL CONSIDERATION==

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.
Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

==6.0      REFERENCES==

Enclosure Page 58 of 60
: 1. RG 1.163, Performance-Based Containment Leak-Test Program, September 1995.
: 2. NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 2012.
: 3. NEI 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, October 2008.
: 4. RG 1.174, Revision 2, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, May 2011.
: 5. RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009.
: 6. NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 1995.
: 7. NUREG-1493, Performance-Based Containment Leak-Test Program, January 1995.
: 8. EPRI TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, August 1994.
: 9. Letter from M. J. Maxin (NRC) to J. C. Butler (NEI), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 94-01, Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (TAC No.
MC9663), dated June 25, 2008.
: 10. Letter from S. Bahadur (NRC) to B. Bradley (NEI), Final Safety Evaluation of Nuclear Energy Institute (NEI) Report 94-01, Revision 3, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J (TAC No. ME2164), dated June 8, 2012.
: 11. Industry/TSTF Standard Technical Specification Change Traveler, Implement 10 CFR 50, Appendix J, Option B, TSTF-52 Revision 3 (ML040400371)
: 12. Letter from NRC (Ram Subbaratnam) to Carolina Power & Light (J. Moyer), H. B.
Robinson Steam Electric Plant Unit 2 - Issuance of Amendment [No. 193] - Technical Specification Change Regarding One-Time Extension of Containment Type A Test Interval, H. B. Robinson Steam Electric Plant, Unit No. 2 (TAC No. MB4658) dated September 16, 2002 (ML022690765)
: 13. Letter to Caroliina Power & Light (J. W. Moyer) from NRC (C. Patel) dated February 11, 2004. (ML040430023). H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of an Amendment Re: Containment Integrated Leak Rate Test (TAC NO. MB9662).
: 14. ML071070170, Letter to T. Walt from C. Patel (NRC) dated June 15, 2007. H. B.
Robinson Steam Electric Plant, Unit No. 2  Issuance of an Amendment for Technical

Enclosure Page 59 of 60 Specifications Changes Related to Containment Peak Pressure (TAC NO. MD2682)
: 15. ML0082210549, Letter to T. Walt from M. Vaaler (NRC) dated October 3, 2008. H. B.
Robinson Steam Electric Plant, Unit No. 2 -Issuance of Amendment Regarding Changes to the Technical Specifications Related to the Isolation Valve Seal Water System (TAC NO. MD7469).
: 16. ML12174A010, Letter to W. Gideon from D. Broaddus (NRC) dated July 13, 2012. H. B.
Robinson Steam Electric Plant, Unit No. 2 - Relief Request-3 for the Fifth 10-Year Interval Inservice Testing Program Plan (TAC NO. ME8260).
: 17. EPRI Report 1003102, Guideline on Nuclear Safety-Related Coatings, Revision 1 (formerly TR-109937).
: 18. Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals, Rev.
4, Developed for NEI by EPRI and Data Systems and Solutions, November 2001.
: 19. Letter from Calvert Cliffs Nuclear Power Plant (Mr. C. H. Cruse) to NRC (Document Control Desk), Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Docket No. 50-317, dated March 27, 2002. (ML020920100)
: 20. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, Revision 2-A of 1009325, EPRI, Palo Alto, CA: 2008.
: 21. ML14148A235, Letter to D. Heacock from S. Williams (NRC) dated July 3, 2014. Surry Power Station, Units 1 And 2- Issuance of Amendment Regarding the Containment Type A And Type C Leak Rate Tests.
: 22. ML15072A264, Letter to L. Weber from A. Dietrich (NRC) dated March 30, 2015. Donald C. Cook Nuclear Plant, Units 1 And 2 -Issuance of Amendments Re: Containment Leakage Rate Testing Program..
: 23. ML15078A058, Letter to E. Larson from T. Lamb (NRC) dated April 8, 2015. Beaver Valley Power Station, Unit Nos. 1 And 2 -Issuance of Amendment Re: License Amendment Request to Extend Containment Leakage Rate Test Frequency.
: 24. ML15154A661, Letter to G. Gellrich from A. Chereskin (NRC) dated July 16, 2015.
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 And 2 -Issuance of Amendments Re:
Extension Of Containment Leakage Rate Testing Frequency.
: 25. ML11186A906, Letter to P. Gillespie from J. Stang (NRC) dated July 28, 2011. Oconee Nuclear Station, Units 1, 2, and 3, Issuance of Amendments Regarding A Proposed Change To The Technical Specifications To Adopt Technical Specification Task Force (TSTF) Technical Change Traveler 52, Revision 3, To Implement Option B of Appendix J To Title 10 of the Code of Federal Regulations, Part 50 (TAC NOS. ME4557, ME4558, AND ME4559).

Enclosure Page 60 of 60
: 26. EPRI 1018243, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, Revision 2-A of 1009325.
: 27. ML15196A559, Letter to B. Hanson from R. Ennis (NRC) dated September 8, 2015.
Peach Bottom Atomic Power Station, Units 2 And 3 - Issuance of Amendments Re:
Extension of Type A and Type C Leak Rate Test Frequencies (TAC NOS. MF5172 AND MF5173).
: 28. Letter to J. Jones (CP&L) from A Schwencer (NRC) dated April 23, 1979. Safety Evaluation Report of the Isolation Valve Seal Water System, H. B. Robinson Unit 2 Docket No. 50-261.
: 29. Letter from W. R. Gideon (Duke Energy, H. B. Robinson) to U. S. Nuclear Regulatory Commission, Docket No. 50-261, ML13267A211, September 16, 2013. License Amendment Request (LAR) To Adopt NFPA 805 Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants (2001 Edition).

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 5 Pages (including this cover page)
PROPOSED TECHNICAL SPECIFICATION CHANGES

Containment 3.6.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                       FREQUENCY SR 3.6.1.1      Perform required Type B and C leakage rate testing       --------NOTE--------
except for containment air lock testing, in accordance   SR 3.0.2 is not Insert 1       with 10 CFR 50, Appendix J, Option A, as modified        applicable by approved exemptions.                                  ------------------------
Insert 2       The leakage rate acceptance criterion is  1.0 L a . In accordance with However, during the first unit startup following testing 10 CFR 50, performed in accordance with 10 CFR 50, Appendix         Appendix J, J, Option A, as modified by approved exemptions,         Option A, as              Insert 2 the leakage rate acceptance criterion is < 0.6 L a for   modified by the Type B and Type C tests.                             approved exemptions SR 3.6.1.2      Verify containment structural integrity in accordance    In accordance with the Containment Tendon Surveillance Program.        with the Containment Tendon Surveillance Program SR 3.6.1.3      Perform required visual examinations and Type A          In accordance with leakage rate testing, in accordance with the             the Containment Containment Leakage Rate Testing Program.                Leakage Rate Testing Program Insert 1:
visual examinations and Insert 2:
the Containment Leakage Rate Testing Program HBRSEP Unit No. 2                           3.6-2                         Amendment No. 176

Containment Air Lock 3.6.2 ACTIONS CONDITION             REQUIRED ACTION                    COMPLETION TIME B.  (continued)            B.3    ----------NOTE--------------
Air lock doors in high radiation areas may be verified locked closed by administrative means.
Verify an OPERABLE               Once per 31 days door is locked closed.
C. Containment air lock   C.1I   Initiate action to               Immediately inoperable for reasons        evaluate overall other than Condition A        containment leakage or B.                         rate per LCO 3.6.1.
Delete Typo            AND C.2    Verify a door is closed in       1 hour the air lock.
AND C.3    Restore air lock to              24 hours OPERABLE status.
D. Required Action and    D.1    Be in MODE 3.                    6 hours associated Completion Time not met.          AND D.2    Be in MODE 5.                    36 hours HBRSEP Unit No. 2                  3.6-5                              Amendment No. 176

Containment Air Lock 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                   FREQUENCY SR 3.6.2.1      ---------------------------NOTES---------------------------------
: 1.      An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
Insert 1
: 2.      Results shall be evaluated against acceptance criteria of SR 3.6.1.1, in accordance with 10                   ---------NOTE-------
CFR 50, Appendix J, Option A, as modified by                    SR 3.0.2 is not approved exemptions.                                            applicable Perform required air lock leakage rate testing in                       In accordance with accordance with 10 CFR 50, Appendix J, Option A,                        10 CFR 50, as modified by approved exemptions.                                     Appendix J, Insert 2                                                                                Option A, as             Insert 2 modified by approved exemptions.
SR 3.6.2.2      Verify only one door in the air lock can be                             24 months opened at a time.
Insert 1:
applicable to Insert 2:
the Containment Leakage Rate Testing Program.
HBRSEP Unit No. 2                                  3.6-6                                 Amendment No. 176

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16        Containment Leakage Rate Testing Program A program      This program provides controls for implementation of the leakage rate testing of shall          the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, establish      Option B, as modified by approved exemptions for Type A testing. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception:
Insert 1
: a. NEI 94 1995, Section 9.3.2: The first Type A test performed after the April 9, 1992, Type A test shall be performed no later than April 9, 2007.
Type B and C testing shall be implemented in the program in accordance with the requirements of 10 CFR 50, Appendix J, Option A.
The peak calculated containment internal pressure for the design basis loss of The            coolant accident, P a , is 42 psig, which exceeds the peak calculated containment containment    internal pressure for the design basis loss of coolant accident.
design pressure is    The maximum allowable containment leakage rate, L a , at P a , shall be 0.1% of the 42 psig.       containment air weight per day.
Leakage rate acceptance criteria are:
: a. 1. Containment leakage rate acceptance criteria is < 1.0 L a . During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 L a for the Type B and Type C tests, and
                              < 0.75 L a for Type A tests.
Insert 2 The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
Insert 3 5.5.17        Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be implemented to ensure that, with an OPERABLE Control Room Emergency Filtration System, CRE occupants can control the nuclear power unit safely following a radiological event, hazardous chemical release, or a smoke challenge. The program shall include the following elements:
Insert 1: NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.
Insert 2: 2. Air lock testing acceptance criteria are:
: a. Overall air lock leakage rate is < 0.05 La when tested at > Pa.
: b. For each door, leakage rate is < 0.01 La when pressurized to > 42 psig            (continued)
HBRSEP                                           5.0-22                       Amendment No. 219 Insert 3: Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 5 Pages (including this cover page)
REVISED TECHNICAL SPECIFICATION PAGES

Containment 3.6.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.6.1.1      Perform required visual examinations and leakage         In accordance with rate testing except for containment air lock testing, in the Containment accordance with the Containment Leakage Rate             Leakage Rate Testing Program.                                         Testing Program SR 3.6.1.2      Verify containment structural integrity in accordance    In accordance with the Containment Tendon Surveillance Program.        with the Containment Tendon Surveillance Program HBRSEP Unit No. 2                           3.6-2                             Amendment No.

Containment Air Lock 3.6.2 ACTIONS CONDITION            REQUIRED ACTION                    COMPLETION TIME B.  (continued)            B.3   ----------NOTE--------------
Air lock doors in high radiation areas may be verified locked closed by administrative means.
Verify an OPERABLE               Once per 31 days door is locked closed.
C. Containment air lock   C.1   Initiate action to               Immediately inoperable for reasons       evaluate overall other than Condition A       containment leakage or B.                        rate per LCO 3.6.1.
AND C.2   Verify a door is closed in       1 hour the air lock.
AND C.3   Restore air lock to              24 hours OPERABLE status.
D. Required Action and    D.1   Be in MODE 3.                    6 hours associated Completion Time not met.          AND D.2   Be in MODE 5.                    36 hours HBRSEP Unit No. 2                 3.6-5                                  Amendment No.

Containment Air Lock 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                   FREQUENCY SR 3.6.2.1      ---------------------------NOTES---------------------------------
: 1.      An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.
: 2.      Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.
Perform required air lock leakage rate testing in                       In accordance with accordance with the Containment Leakage Rate                            the Containment Testing Program.                                                        Leakage Rate Testing Program.
SR 3.6.2.2      Verify only one door in the air lock can be                             24 months opened at a time.
HBRSEP Unit No. 2                                  3.6-6                                     Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16       Containment Leakage Rate Testing Program A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in NEI 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A, dated July 2012, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008.
The peak calculated containment internal pressure for the design basis loss of coolant accident, Pa, is 42 psig. The containment design pressure is 42 psig.
The maximum allowable containment leakage rate, La, at Pa, shall be 0.1% of the containment air weight per day.
Leakage rate acceptance criteria are:
: 1. Containment leakage rate acceptance criteria is < 1.0 La. During the first unit 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 Type A tests.
: 2. Air lock testing acceptance criteria are:
: a. Overall air lock leakage rate is  0.05 La when tested at  Pa.
: b. For each door, leakage rate is  0.05 La when pressurized to  42 psig.
The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.
Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.
5.5.17       Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be implemented to ensure that, with an OPERABLE Control Room Emergency Filtration System, CRE occupants can control the nuclear power unit safely following a radiological event, hazardous chemical release, or a smoke challenge. The program shall include the following elements:
(continued)
HBRSEP                                         5.0-22                       Amendment No.

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 11 Pages (including this cover page)
PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES

SR Applicability B 3.0 BASES SR 3.0.1          Surveillances, including Surveillances invoked by Required (continued)
Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.
Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.
SR 3.0.2          SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that equires the periodic performance of the Required Action on a once per . . . interval.
SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).
The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25% extension of the interval specified in the Frequency does not apply.
These exceptions are stated in the individual Specifications. An example of where SR 3.0.2 does not apply is a Surveillance with a Frequency of in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions.
(continued)
HBRSEP Unit No. 2                         B 3.0-14                             Revision No. 52

An example of where SR 3.0.2 does not apply is in the Containment Leakage Rate Testing Program.                  SR Applicability This program establishes testing requirements and                        B 3.0 Frequencies in accordance with the requirements of BASES             regulations.
SR 3.0.2           The requirements of regulations take precedence over the TS.
(continued)       The TS cannot in and of themselves extend a test interval specified in the regulations.
Therefore, there is a Note in the Frequency stating, "SR 3.0.2 is not applicable."
As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ..." basis. The 25% extension applies to each performance after the initial performance. The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25% extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.
The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.
SR 3.0.3           SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.
This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.
The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, (continued)
HBRSEP Unit No. 2                         B 3.0-15                               Revision No. 52

Containment B 3.6.1 B 3.6 CONTAINMENT SYSTEMS B 3.6.1 Containment BASES BACKGROUND          The containment consists of the concrete reactor building, its steel liner, and the penetrations through this structure. The structure is designed to contain radioactive material that may be released from the reactor core design basis        following a Design Basis Accident (DBA). Additionally, this structure loss of             provides shielding from the fission products that may be present in the coolant             containment atmosphere following accident conditions.
accident (LOCA)              The containment is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow dome roof. The inside surface of the containment is lined with a stainless steel liner to ensure a high degree of leak tightness during operating and accident conditions.
The cylinder wall is prestressed with a post tensioning system in the vertical direction.
Design Basis Accident (DBA)
The concrete reactor building is required for structural integrity of the SR 3.6.1.1 leakage    containment under DBA conditions. The steel liner and its penetrations rate requirements     establish the leakage limiting boundary of the containment. Maintaining comply with 10 CFR    the containment OPERABLE limits the leakage of fission product 50, Appendix J,       radioactivity from the containment to the environment.
Option B (Ref. 1), as modified by approved  The isolation devices for the penetrations in the containment boundary exemptions.           are a part of the containment leak tight barrier. To maintain this leak tight barrier:
: a.       All penetrations required to be closed during accident conditions are either:
: 1.      capable of being closed by an OPERABLE automatic containment isolation system, or
: 2.      closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LCO 3.6.3, "Containment Isolation Valves";
(continued)
HBRSEP Unit No. 2                            B 3.6-1                              Revision No. 0

Containment B 3.6.1 BASES BACKGROUND        b.       The air lock is OPERABLE, except as provided in (continued)               LCO 3.6.2, "Containment Air Lock";
: c.       The equipment hatch is closed and sealed; and
: d.       The Isolation Valve Seal Water (IVSW) sytem is OPERABLE, except as provided in LCO 3.6.8.
APPLICABLE        The safety design basis for the containment is that the SAFETY ANALYSES containment must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.
The DBAs that result in a challenge to containment OPERABILITY from high pressures and temperatures are a loss of coolant accident (LOCA) and a steam line break (Ref. 2). In addition, release of significant fission product radioactivity within containment can occur from a LOCA. In the LOCA analyses, it is assumed that the containment is OPERABLE such that, for the LOCA, the release to the environment is controlled by the rate of containment leakage. The containment has an allowable leakage rate of 0.1% of containment air weight per day (Ref. 2). This leakage rate, used to evaluate offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Ref. 1), as L a : the maximum allowable
   , Option B containment leakage rate at the calculated peak containment internal pressure (P a ) resulting from the LOCA. At HBRSEP, Unit 2, P a is design basis specified as the containment design pressure of 42 psig, which exceeds the calculated peak LOCA containment pressure. The allowable leakage rate represented by L a forms the basis for the acceptance criteria imposed on all containment leakage rate testing. L a is assumed to be 0.1% per day in the safety analysis at P a = 42 psig (Ref. 2).
Satisfactory leakage rate test results are a requirement for the establishment of containment OPERABILITY.
The containment satisfies Criterion 3 of the NRC Policy Statement.
(continued)
HBRSEP Unit No. 2                          B 3.6-2                            Revision No. 34

Containment B 3.6.1 BASES (continued)
LCO               Containment OPERABILITY is maintained by limiting leakage to  1.0 L a ,
except prior to the first startup after performing a required 10 CFR 50, Containment      Appendix J, leakage test. At this time, the combined Type B and C Leakage Rate     leakage must be < 0.6 L a and the overall Type A leakage must be Testing          < 0.75 L a . applicable leakage limits must be met.
Program Compliance with this LCO will ensure a containment configuration, including the equipment hatch, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analysis.
Individual leakage rates specified for the containment air lock are not specifically part of the acceptance criteria of 10 CFR 50, Appendix J.
Therefore, leakage rates exceeding these individual limits only result in the containment being inoperable when the leakage results in exceeding overall          the acceptance criteria of Appendix J. 1.0 La APPLICABILITY     In MODES 1, 2, 3, and 4, a LOCA could cause a release of radioactive material into containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, containment is not required to be OPERABLE in MODE 5 to prevent leakage of radioactive material from containment. The requirements for containment during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations."
ACTIONS           A.1 In the event containment is inoperable, containment must be restored to OPERABLE status within 1 hour. The 1 hour Completion Time provides a period of time to correct the problem commensurate with the importance of maintaining containment OPERABLE during MODES 1, 2, 3, and 4. This time period also ensures that the probability of an accident (requiring containment OPERABILITY) occurring during periods when containment is inoperable is minimal.
(continued)
HBRSEP Unit No. 2                          B 3.6-3                               Revision No. 0

Containment B 3.6.1 BASES ACTIONS               B.1 and B.2 (continued)
If containment cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE          SR 3.6.1.1 REQUIREMENTS Maintaining the containment OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 1), Option Insert 1              A, as modified by approved exemptions. Air lock leakage is not acceptable if its contribution to overall Type B, and C leakage causes overall Type B and C leakage to exceed limits. As left leakage prior to the first startup after performing a required 10 CFR 50, Appendix J,         Insert 2 leakage test is required to be < 0.6 L a for combined Type B and C Insert 3              leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an overall leakage limit of  1.0 L a . At 1.0 L a the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by Appendix J, Insert 1              Option A. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.
Reviewer's Note: NEI 94-01 includes acceptance criteria for as-left and as-found Type A leakage rates and combined Type B and C leakage SR 3.6.1.2 rates, which may be reflected in the Bases.
This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program.
Insert 1: the Containment Leakage Rate testing Program.
Insert 2: Containment Leakage Rate testing Program Insert 3: , < 0.60 La for the Type B and Type C tests, and < 0.75 La for Type A tests.
(continued)
HBRSEP Unit No. 2                              B 3.6-4                             Revision No. 0

Containment B 3.6.1 BASES SURVEILLANCE      SR 3.6.1.3 REQUIREMENTS (continued)      Maintaining the containment OPERABLE requires compliance with the visual examinations and leakage rate test requirements of the Containment Leakage Rate Testing Program applicable to Type A leakage rate tests. Air lock leakage is not acceptable if its contribution to overall Type A leakage causes overall Type A leakage to exceed limits.
As left leakage after performing a required 10 CFR 50, Appendix J, leakage test is required to be < 0.75 L a for overall Type A leakage. At 1.0 L a the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by the Containment Leakage Rate Testing Program. This periodic testing requirement verifies that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.
REFERENCES        1.       10 CFR 50, Appendix J. , Option B
: 2.       UFSAR, Section 6.2.
HBRSEP Unit No. 2                         B 3.6-5                               Revision No. 0

Containment Air Lock B 3.6.2 BASES SAFETY ANALYSES leakage. The containment has an allowable leakage rate of 0.1% of (continued)      containment air weight per day at 42 psig (Ref. 2).
The containment air lock satisfies Criterion 3 of the NRC Policy Statement.
LCO               The containment air lock forms part of the containment pressure boundary. As part of containment, the air lock safety function is related to control of the containment leakage rate resulting from a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.
The air lock is required to be OPERABLE. For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door of an air lock to be opened at one time. This provision ensures that a gross breach of containment does not exist when containment is required to be OPERABLE. Closure of a single door in the air lock is sufficient to provide a leak tight barrier following postulated events. Nevertheless, both doors are kept closed when the air lock is not being used for normal entry into and exit from containment.
or APPLICABILITY     In MODES 1, 2, 3, and 4, a DBA could cause a release of radioactive material to containment. In MODES 5 and 6, the probability and consequences of these events are reduced due to the pressure and temperature limitations of these MODES. Therefore, the containment air locks are not required in MODE 5 to prevent leakage of radioactive material from containment. The requirements for the containment air locks during MODE 6 are addressed in LCO 3.9.3, "Containment Penetrations."
(continued)
HBRSEP Unit No. 2                          B 3.6-7                                 Revision No.34

Containment Air Lock B 3.6.2 BASES ACTIONS           C.1, C.2, and C.3 (continued) inoperable air lock to OPERABLE status, assuming that at least one door is maintained closed in the air lock.
D.1 and D.2 If the inoperable containment air lock cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE      SR 3.6.2.1 REQUIREMENTS Maintaining the containment air lock OPERABLE requires compliance with the leakage rate test requirements of 10 CFR 50, Appendix J (Ref. 1), Option A, as modified by approved exemptions. This SR reflects the the leakage rate testing requirements with regard to air lock leakage Containment       (Type B leakage tests). The periodic testing requirements verify that the Leakage Rate      air lock leakage does not exceed the allowed fraction of the overall Testing           containment leakage rate. The Frequency is required by Appendix J Program           (Ref. 1), Option A, as modified by approved exemptions. Thus, SR 3.0.2 (which allows Frequency extensions) does not apply.
The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria of SR 3.6.1.1. This ensures that air lock leakage is properly accounted for which is         in determining the overall containment leakage rate.
applicable to combined Type B and C (continued)
HBRSEP Unit No. 2                        B 3.6-11                              Revision No. 0

Containment Air Lock B 3.6.2 BASES SURVEILLANCE      SR 3.6.2.2 REQUIREMENTS (continued)      The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not inadvertently occur. Due to the purely mechanical nature of this interlock, and given that the interlock mechanism is not normally challenged when the containment air lock door is used for entry and exit (procedures require strict adherence to single door opening), this test is only required to be performed every 24 months. The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage, and the potential for loss of containment OPERABILITY if the surveillance were performed with the reactor at power. The 24 month Frequency for the interlock is justified based on generic operating experience. The 24 month Frequency is based on engineering judgment and is considered adequate given that the interlock is not challenged during the use of the interlock.
REFERENCES        1.      10 CFR 50, Appendix J. , Option B
: 2.      UFSAR, Paragraph 6.9.2.
HBRSEP Unit No. 2                          B 3.6-12                             Revision No. 0

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 80 Pages (including this cover page)
EVALUATION OF RISK SIGNIFICANCE OF PERMANENT ILRT EXTENSION

Robinson Nuclear Plant:
Evaluation of Risk Significance of Permanent ILRT Extension 54002-CALC-01 Prepared for:
Robinson Nuclear Plant Project

==Title:==
Permanent ILRT Extension Revision: 0A Name and Date Preparer: Justin Sattler Reviewer: Matthew Johnson Review Method                  Design Review  Alternate Calculation Approved by: Nicholas Lovelace Revision 0A                                                                Page 1 of 79

54002-CALC-01       Evaluation of Risk Significance of Permanent ILRT Extension REVISION RECORD 

==SUMMARY==

Revision                    Revision Summary 0A                          Initial Issue Revision 0A                                                         Page 2 of 79

54002-CALC-01                                   Evaluation of Risk Significance of Permanent ILRT Extension 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 .................................................................................................................20 5.2.4 Step 4 - Determine the Change in Risk in Terms of LERF...................................22 5.2.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability
               ..........................................................................................................................25 5.2.6 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage
               ..........................................................................................................................25 5.3 Sensitivities .................................................................................................................28 5.3.1 Potential Impact from External Events Contribution .............................................28 5.3.2 Potential Impact from Steel Liner Corrosion Likelihood ........................................30 5.3.3 Expert Elicitation Sensitivity .................................................................................32 5.3.4 Large Leak Probability Sensitivity Study ..............................................................33 6.0    RESULTS ......................................................................................................................34

==7.0    CONCLUSION==
S AND RECOMMENDATIONS ..............................................................35 A. Attachment 1 .................................................................................................................36 Revision 0A                                                                                                               Page 3 of 79

54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension 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 Robinson Nuclear Plant (RNP). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A [Reference 1], the methodology used in EPRI TR-104285 [Reference 2], the NEI Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals from November 2001 [Reference 3], the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide 1.200 as applied to ILRT interval extensions, and risk insights in support of a request for a 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 1009325, Revision 2-A of EPRI 1018243 [Reference 24].
2.0      SCOPE Revisions to 10CFR50, Appendix J (Option B) allow individual plants to extend the Integrated Leak Rate Test (ILRT) Type A surveillance testing frequency requirement from three in ten years to at least once in ten years. The revised Type A frequency is based on an acceptable performance history defined as two consecutive periodic Type A tests at least 24 months apart in which the calculated performance leakage rate was less than limiting containment leakage rate of 1La.
The basis for the current 10-year test interval is provided in Section 11.0 of NEI 94-01, Revision 0, and established in 1995 during development of the performance-based Option B to Appendix J. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, September 1995 [Reference 6], provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessment of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the 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 RNP.
NEI 94-01 Revision 2-A contains a Safety Evaluation Report that supports using EPRI Report No. 1009325 Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, for performing risk impact assessments in support of ILRT extensions [Reference 24].
The Guidance provided in Appendix H of EPRI Report No. 1009325 Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.
It should be noted that containment leak-tight integrity is also verified through periodic in-service Revision 0A                                                                               Page 4 of 79

54002-CALC-01                          Evaluation of Risk Significance of Permanent ILRT Extension inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for in-service inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment. The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, Type B local leak tests performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets are also not affected by the change to the Type A test frequency.
The acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanent extension of the Type A test interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. Since the Type A test does not impact CDF, the relevant criterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 10-6 per reactor year. RG 1.174 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the Conditional Containment Failure Probability (CCFP), which helps ensure the defense-in-depth philosophy is maintained, is also calculated.
Regarding CCFP, changes of up to 1.1% have been accepted by the NRC for the one-time requests for extension of ILRT intervals. In context, it is noted that a CCFP of 1/10 (10%) has been approved for application to evolutionary light water designs. Given these perspectives, a change in the CCFP of up to 1.5% is assumed to be small.
In 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.
For those plants that credit containment overpressure for the mitigation of design basis accidents, a brief description of whether overpressure is required should be included in this section. In addition, if overpressure is included in the assessment, other risk metrics such as CDF should be described and reported.
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==3.0     REFERENCES==

The following references were used in this calculation:
: 1. Revision 3-A to Industry Guideline for Implementing Performance-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: Surry Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 3, Revision 1, Part 1, October 1990.
: 8. Letter from R. J. Barrett (Entergy) to U. S. Nuclear Regulatory Commission, IPN-01-007, January 18, 2001.
: 9. United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB0178), April 17, 2001.
: 10. Impact of 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. Calculation No. RNP-F/PSA-0077, Revision 2, Robinson Nuclear Plant, RNP QUANTIFICATION CALCULATION, September 2012.
: 18. Calculation No. RNP-F/PSA-0048, Revision 2, H.B. Robinson Steam Electric Plant, PSA Model Section 9.0 - Source Term Development.
: 19. Robinson Nuclear Plant License Renewal Application, 2004.
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54002-CALC-01                         Evaluation of Risk Significance of Permanent ILRT Extension
: 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. Procedure EST-085, Revision 17, H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2, CONTAINMENT INTEGRATED LEAK RATE TEST.
: 28. Letter L-14 -121, ML14111A291, FENOC Evaluation of the Proposed Amendment, Beaver Valley Power Station, Unit Nos. 1 and 2, April 2014.
: 29. Technical Letter Report ML112070867, Containment Liner Corrosion Operating Experience Summary, Revision 1, August 2011.
: 30. Calculation No. RNP-RA/02-0028, Revision 0, H.B. Robinson Steam Electric Plant, One-Time Extension of Containment Type A Test Interval, March 2002.
: 31. Armstrong, J., Simplified Level 2 Modeling Guidelines: WOG PROJECT: PA-RMSC-0088, Westinghouse, WCAP-16341-P, November 2005.
: 32. Calculation No. RNP-F/PSA-0046, Revision 1, H.B. Robinson Steam Electric Plant, PSA Model Section 7.0 - Containment Safeguards and Plant Damage States.
: 33. Evacuation Time Estimates for the RNP Steam Electric Plant, Plume Exposure Pathway, Emergency Planning Zone, HMM Associates, June 1987.
: 34. Harrison, D., Generic Component Fragilities for the GE Advanced BWR Seismic Analysis, International Technology Corporation, September 1988.
: 35. IPEEE, H. B. Robinson Steam Electric Plant, Unit No. 2, Individual Plant Examination Submittal, August 1992.
: 36. Calculation No. RNP-F/PSA-0047, Revision 2, H.B. Robinson Steam Electric Plant, PSA Model Section 8.0 - Containment Response Assessment.
: 37. Calculation No. 4150-Seismic-005, Revision 0, AMEC Foster Wheeler Environment &
Infrastructure, Inc., Soil Hazard and Smoothed GMRS at Elevation 226 ft for the Robinson Plant Area.
: 38. Letter from W. R. Gideon (Duke Energy, H. B. Robinson) to U. S. Nuclear Regulatory Commission, Docket No. 50-261, ML13267A211, September 16, 2013.
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54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension
: 39. Calculation No. RNP-F/PSA-0054, Revision 1, H.B. Robinson Steam Electric Plant, Unit No. 2, PSA Appendix J - Containment Overpressure Capacity.
: 40. Calculation No. RNP-F/PSA-0050, Revision 0, H.B. Robinson Steam Electric Plant, Unit No. 2, Estimate of 50 Mile Population Dose from Design Basis Containment Leakage Following a Core Melt Accident.
: 41. 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.
: 42. The Nuclear Energy Institute - Seismic Risk Evaluations for Plants in the Central and Eastern United States, ML14083A596, March 2014.
4.0     ASSUMPTIONS AND LIMITATIONS The following assumptions were used in the calculation:
The technical adequacy of the RNP PRA is consistent with the requirements of Regulatory Guide 1.200 as is relevant to this ILRT interval extension, as detailed in Attachment 1.
The RNP Level 1 and Level 2 internal events PRA models provide representative results.
It is appropriate to use the RNP 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 analysis and the detailed Fire PRA (model RNP_12Fh_s1) are used for this sensitivity analysis. It is reasonable to assume that the impact from the ILRT extension (with respect to percent increases in population dose) will not substantially differ if detailed analysis of seismic and 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.
The representative containment leakage for Class 3a sequences is 10La based on the previously approved methodology performed for Indian Point Unit 3 [Reference 8, Reference 9].
The representative containment leakage for Class 3b sequences is 100La based on the guidance provided in EPRI Report No. 1009325, Revision 2-A (EPRI 1018243)
[Reference 24].
The Class 3b can be very conservatively categorized as LERF based on the previously approved methodology [Reference 8, Reference 9].
The impact on population doses from containment bypass scenarios is not altered by the proposed ILRT extension, but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes on the conclusions from this analysis will result from this separate categorization.
The reduction in ILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal.
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54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension 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 LLRT test intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that is used as the basis for the consequence analysis of the ILRT interval extension for RNP. 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 Revision 0A                                                                              Page 9 of 79

54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension 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:
Sthe 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.
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 Revision 0A                                                                             Page 10 of 79

54002-CALC-01                         Evaluation of Risk Significance of Permanent ILRT Extension
: 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:
Sthe 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 yearS 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 Surry. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. With the RNP Level 2 model end-states assigned to one of the NUREG/CR-4551 APBs, it is considered adequate to represent RNP. (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 methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms was factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder and dome and a concrete base-mat, each with a steel liner.
EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals [Reference 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 RNP 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.
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54002-CALC-01                              Evaluation of Risk Significance of Permanent ILRT Extension 5.1.2   Plant Specific Inputs The plant-specific information used to perform the RNP ILRT Extension Risk Assessment includes the following:
Level 1 Model results [Reference 17]
Level 2 Model results [Reference 17, Reference 18, Reference 19]
Release category definitions used in the Level 2 Model [Reference 18, Reference 19]
Dose within a 50-mile radius [Reference 19]
ILRT results to demonstrate adequacy of the administrative and hardware issues
[Reference 30]
Containment failure probability data [Reference 18, Reference 32, Reference 33]
RNP Model The Internal Events PRA Model that is used for RNP is characteristic of the as-built plant. The current Level 1 model (RNP PRA Model Version RNP_12) [Reference 17] is a linked fault tree model. The total CDF is 1.05E-5/year [Table 18 of Reference 17]. When the #PDS_TOP gate is quantified to use flag events to classify the cutsets into plant damage states, the calculated CDF is 1.72E-5/year [Table 14 of Reference 17]. Since this calculation uses the results from the PDS quantification that contains double counting of scenarios risk by classifying the same cutset in multiple PDSs and the CAFTA software uses the min-cut upper bound quantification approach, results of this analysis are conservative. Table 5-1 and Table 5-2 provides a summary of the Internal Events CDF and LERF results for RNP PRA Model Version RNP_12. The total LERF when quantified with PDSs is 1.67E-6/year.
The total Fire CDF is 1.80E-5/year, and the total Fire LERF is 3.76E-6/year (RNP Fire PRA Model Version RNP_12Fh_s1). The Seismic PRA results from the IPEEE Seismic Margins Analysis do not result in estimate of CDF [Reference 35]. Refer to Section 5.3.1 for further details on external events as they pertain to this analysis.
Table 5 Internal Events CDF (RNP PRA Model Version RNP_12)
Internal Events                  Frequency (per year)
ISLOCA                                          2.83E-08 LOCA                                            1.64E-06 SGTR                                            1.12E-07 Internal Flood                                  1.92E-06 TransientLOCA                                  3.38E-06 TransientLoss of DHR                           9.10E-06 ATWS                                            1.01E-06 Total Internal Events CDF                       1.72E-05 Total Internal Events CDF 1.71E-05 (Excluding ISLOCA & SGTR)
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54002-CALC-01                               Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Internal Events LERF (RNP PRA Model Version RNP_12)
Internal Events                        Frequency (per year)
ISLOCA                                                  1.60E-09 LOCA                                                    3.71E-08 SGTR                                                    1.05E-07 Internal Flood                                          3.55E-07 TransientLOCA                                          3.53E-07 TransientLoss of DHR                                   7.59E-07 ATWS                                                    5.55E-08 Total Internal Events LERF                              1.67E-06 Population Dose Calculations The population dose calculation was reported in the One-Time Extension of Containment Type A Test Interval and Estimate of 50 Mile Population Dose from Design Basis Containment Leakage Following a Core Melt Accident [References 30 and 40]. Table 5-3 presents dose exposures calculated from methodology described in Reference 1 and data from Reference 30.
Reference 30 provides the population dose (person-rem) for Classes 1, 2, 6, 7, and 8; Class 3a and 3b population dose values are calculated from the Class 1 population dose and represented as 10La and 100L a, respectively, as guidance in Reference 1 dictates.
Table 5 Population Dose Accident Class                             Description                               Release (person-rem) 1                                 Containment Remains Intact                               1.56E+03 2                                Containment Isolation Failures                            2.22E+06 3a                     Independent or Random Isolation Failures SMALL                     1.56E+041 3b                     Independent or Random Isolation Failures LARGE                     1.56E+052 Isolation Failure in which pre-existing leakage is not                  n/a 4
dependent on sequence progression. Type B test Failures Isolation Failure in which pre-existing leakage is not                  n/a 5
dependent on sequence progression. Type C test Failures 6                Isolation Failure that can be verified by IST/IS or surveillance          8.85E+06 7                       Containment Failure induced by severe accident                     2.62E+06 8                         Accidents in which containment is by-passed                      2.45E+07
: 1. 10
* La
: 2. 100
* La Release Category Definitions Table 5-4 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.
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54002-CALC-01                                   Evaluation of Risk Significance of Permanent ILRT Extension Table 5 EPRI Containment Failure Classification [Reference 2]
Class                                                      Description Containment remains intact including accident sequences that do not lead to containment failure in the 1       long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant.
Containment isolation failures (as reported in the Individual Plant Examinations) including those accidents 2
in which there is a failure to isolate the containment.
Independent (or random) isolation failures include those accidents in which the pre-existing isolation 3
failure to seal (i.e., provide a leak-tight containment) is not dependent on the sequence in progress.
Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 3 isolation 4
failures, but is applicable to sequences involving Type B tests and their potential failures. These are the Type B-tested components that have isolated, but exhibit excessive leakage.
Independent (or random) isolation failures including those accidents in which the pre-existing isolation 5       failure to seal is not dependent on the sequence in progress. This class is similar to Class 4 isolation failures, but is applicable to sequences involving Type C test and their potential failures.
Containment isolation failures including those leak paths covered in the plant test and maintenance 6
requirements or verified per in-service inspection and testing (ISI/IST) program.
Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J 7
testing requirements do not impact these accidents.
Accidents in which the containment is bypassed (either as an initial condition or induced by phenomena) 8 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 5-4, 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 delta LERF values for several plants may fall above the very small change guidelines of the NRC Regulatory Guide 1.174
[Reference 4]. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for 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 Revision 0A                                                                                              Page 14 of 79

54002-CALC-01                          Evaluation of Risk Significance of Permanent ILRT Extension 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 RNP, as detailed in Section 5.2, involves the following:
The Class 2 and Class 8 sequences and some Class 6 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. The subtracted Class 6 sequences involve large early containment failure. These sequences are already considered to contribute to LERF in the RNP 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 5-5.
The analysis performed examined RNP-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 Revision 0A                                                                                Page 15 of 79

54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension evaluation as part of the baseline risk profile. However, they are not affected by the ILRT frequency change.
Class 4 and 5 sequences are impacted by changes in Type B and C test intervals; therefore, changes in the Type A test interval do not impact these sequences.
Table 5 EPRI Accident Class Definitions Accident Classes                                             Description (Containment Release Type) 1                                              No Containment Failure 2                                     Large Isolation Failures (Failure to Close) 3a                                      Small Isolation Failures (Liner Breach) 3b                                      Large Isolation Failures (Liner Breach) 4                                Small Isolation Failures (Failure to Seal - Type B) 5                                Small Isolation Failures (Failure to Seal - Type C) 6                                Other Isolation Failures (e.g., Dependent Failures) 7                                 Failures Induced by Phenomena (Early and Late) 8                                        Bypass (Interfacing System LOCA)
CDF                           All CET End States (Including Very Low and No Release)
The steps taken to perform this risk assessment evaluation are as follows:
Step 1 - Quantify the baseline risk in terms of frequency per reactor year for each of the accident classes presented in Table 5-5.
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 5-5 were developed for RNP by first determining the frequencies for Classes 1, 2, 6, 7, and 8. Table 5-6 presents the grouping of each release category in EPRI Classes based on the associated description. Table 5-7 presents the release categories, frequency, and EPRI category for each sequence and the totals of each EPRI classification. Table 5-8 provides a summary of the accident sequence Revision 0A                                                                                         Page 16 of 79

54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension frequencies that can lead to radionuclide release to the public and have been derived consistent with the definitions of accident classes defined in EPRI TR-104285 [Reference 2], the NEI Interim Guidance [Reference 3], and guidance provided in EPRI Report No. 1009325, Revision 2-A [Reference 24]. Adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion of the steel liner per the methodology described in Section 5.2.6. 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
         =         = 0.0092 217 Multiplying the CDF by the probability of a Class 3a leak yields the Class 3a frequency contribution in accordance with guidance provided in Reference 24. As described in Section 5.1.3, additional consideration is made to not apply failure probabilities on those cases that are already LERF scenarios (i.e., release categories RC-2, RC-2B, RC-4, RC-4C, RC-5, and RC-5C
[Reference 18]; values for these release categories are shown in Table 5-6). Therefore, these LERF contributions from CDF are removed. Therefore, the frequency of a Class 3a failure is calculated by the following equation:
               =                    2     2        4    4       5    5
         =        1.72E 5.80E 3.82E 0 - 1.15E 0 - 5.88E-7 = 1.43E-7 In the database of 217 ILRTs, there are zero containment leakage events that could result in a large early release. Therefore, the Jeffreys non-informative prior is used to estimate a failure rate and is illustrated in the following equations:
5678 9: ;<=6 > + 1/2 Jeffreys Failure Probability =
5678 9: A >B> + 1 0 + 1/2 C =             = 0.0023 217 + 1 The frequency of a Class 3b failure is calculated by the following equation:
C =         C        2 2  4 4  5 5
             .D
         =        1.72E 5.80E 3.82E 0 - 1.15E 0 - 5.88E-7 = 3.56E-8 E
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 Revision 0A                                                                             Page 17 of 79

54002-CALC-01                                Evaluation of Risk Significance of Permanent ILRT Extension per year is initially determined from the EPRI Accident Class 1 frequency listed in Table 5-7 and then subtracting the EPRI Class 3a and 3b frequency (to preserve total CDF), calculated below:
            =                                              C Class 2 Sequences. This group consists of core damage accident progression bins with large containment isolation failures. The frequency per year for these sequences is obtained from the EPRI Accident Class 2 frequency listed in Table 5-7.
Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis, consistent with approved methodology.
Class 5 Sequences. This group consists of all core damage accident progression bins for which a containment isolation failure-to-seal of Type C test components. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis, consistent with approved methodology.
Class 6 Sequences. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs.
These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution. For RNP, this class is defined as the release categories RC-2, RC-2B, and RC-3. The frequency per year for these sequences is obtained from the EPRI Accident Class 6 frequency listed in Table 5-7.
Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (e.g., overpressure). For this analysis, the frequency is determined from the EPRI Accident Class 7 frequency listed in Table 5-7.
Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass or SGTR occurs. For this analysis, the frequency is determined from the EPRI Accident Class 8 frequency listed in Table 5-7.
LERF quantification is distributed into EPRI categories based on release categories. Table 5-6 shows this distribution.
Table 5 Release Category Frequencies Release                         Description of Outcome                       EPRI   Frequency (/yr)
Category                                                                   Category IC-1                              Containment intact                         1         1.41E-05 RC-1       Late containment failure caused by gradual overpressurization     7         0.00E+00 Late containment failure caused by gradual overpressurization RC-1A                                                                         7         0.00E+00 with revaporization Late containment failure caused by gradual overpressurization RC-1B                                                                         7         1.16E-06 without scrubbing Late containment failure caused by gradual overpressurization RC-1BA                                                                        7         4.66E-08 with revaporization, without scrubbing RC-2                       Large early containment failure                   6         5.80E-07 RC-2B              Large early containment failure without scrubbing          6         3.82E-07 Early containment isolation failure RC-3                                                                         6         1.84E-07 with a small leakage rate Early containment isolation failure RC-3B                                                                         7         2.31E-08 with a small leakage rate without scrubbing RC-4               Containment bypass with a small leakage rate              2         0.00E+00 RC-4C      Containment bypass with a small leakage rate without scrubbing     8         1.15E-07 Revision 0A                                                                                  Page 18 of 79

54002-CALC-01                                  Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Release Category Frequencies Release                           Description of Outcome                            EPRI      Frequency (/yr)
Category                                                                          Category Containment bypass with a large RC-5                                                                                8           0.00E+00 leakage rate Containment bypass with a large RC-5C                                                                                8           5.88E-07 leakage rate without scrubbing Contribution to EPRI Classification 6                                  1.15E-06 Contribution to EPRI Classification 7                                  1.23E-06 Contribution to EPRI Classification 8                                  7.03E-07 Table 5 Accident Class Frequencies Release Categories                      EPRI Category              Frequency (/yr)
IC-1                                Class 1                  1.41E-05 RC-4                                Class 2                  0.00E+00 RC-2, RC-2B, RC-3                           Class 6                  1.15E-06 RC-1, RC-1A, RC-1B, RC-1BA, RC-3B                    Class 7                  1.23E-06 RC-4C, RC-5, RC-5C                           Class 8                  7.03E-07 Total (CDF)                               N/A                    1.72E-05 Table 5 Baseline Risk Profile Class                                    Description                                  Frequency (/yr) 1                               No containment failure                                1.39E-052 2                      Large containment isolation failures                           0.00E+00 3a                      Small isolation failures (liner breach)                        1.43E-07 3b                      Large isolation failures (liner breach)                        3.56E-08 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)          1.15E-06 7          Severe accident phenomena induced failure (early and late)                 1.23E-06 8                                Containment bypass                                   7.03E-07 Total                                       1.72E-05
: 1. represents a probabilistically insignificant value.
: 2. The Class 3a and 3b frequencies are subtracted from Class 1 to preserve total CDF.
5.2.2    Step 2 - Develop Plant-Specific Person-Rem Dose (Population Dose)
Plant-specific release analyses were performed to estimate the person-rem doses to the population within a 50-mile radius from the plant [References 30 and 40]. The releases are based on RNP-specific dose calculations summarized on Table 5-3. Table 5-9 provides a correlation of RNP population dose to EPRI Accident Class. Table 5-10 provides population dose for each EPRI accident class.
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54002-CALC-01                                Evaluation of Risk Significance of Permanent ILRT Extension 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:
G     H =;>> 3; 9I6=;B<9J 9> = 10  1.56G+3 = 1.56G+4 G     H =;>> 38 9I6=;B<9J 9> = 100  1.56G+3 = 1.56G+5 Table 5 Mapping of Population Dose to EPRI Accident Class Release Category           EPRI Category            Frequency (/yr) Dose (person-rem)
IC-1                     Class 1                 1.41E-05         1.56E+03 RC-4                     Class 2                 0.00E+00         2.22E+06 RC-2, RC-2B, RC-3               Class 6                 1.15E-06         8.85E+06 RC-1, RC-1A, RC-Class 7                 1.23E-06         2.62E+06 1B, RC-1BA, RC-3B RC-4C, RC-5, RC-Class 8                 7.03E-07         2.45E+07 5C Table 5 Baseline Population Doses Class                                Description                                 Population Dose (person-rem) 1                             No containment failure                                     1.56E+03 2                     Large containment isolation failures                               2.22E+06 3a                     Small isolation failures (liner breach)                           1.56E+041 3b                     Large isolation failures (liner breach)                           1.56E+052 4                Small isolation failures - failure to seal (type B)                        N/A 5                Small isolation failures - failure to seal (type C)                        N/A 6      Containment isolation failures (dependent failure, personnel errors)              8.85E+06 7          Severe accident phenomena induced failure (early and late)                    2.62E+06 8                              Containment bypass                                        2.45E+07
: 1. 10*La
: 2. 100*La 5.2.3    Step 3 - Evaluate Risk Impact of Extending Type A Test Interval from 10 to 15 Years The next step is to evaluate the risk impact of extending the test interval from its current 10-year interval to a 15-year interval. To do this, an evaluation must first be made of the risk associated with the 10-year interval, since the base case applies to 3-year interval (i.e., a simplified representation of a 3-to-10 interval).
Risk Impact Due to 10-Year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and Class 3b sequences is impacted. The risk contribution is changed based on the NEI guidance as described in Section 5.1.3 by a factor of 10/3 compared to the base case values. The Class 3a and 3b frequencies are calculated as follows:
Revision 0A                                                                                         Page 20 of 79

54002-CALC-01                                   Evaluation of Risk Significance of Permanent ILRT Extension 10 2 K        LMN   =                         2         2     4      4       5     5 3 217 L
        =              1.55E-5 = 4.77E-7 10 . 5 K     C LMN    =                         2         2     4      4       5     5 3 218 L     .D
        =              1.55E-5 = 1.19E-7 E
The results of the calculation for a 10-year interval are presented in Table 5-11.
Table 5 Risk Profile for Once in 10 Year ILRT Class                 Description                 Frequency     Contribution     Population      Population
(/yr)         (%)        Dose (person-      Dose Rate rem)       (person-rem/yr) 1             No containment failure2            1.35E-05       78.62%          1.56E+03         2.11E-02 2     Large containment isolation failures       0.00E+00       0.00%           2.22E+06        0.00E+00 Small isolation failures (liner 3a                                                4.77E-07       2.77%           1.56E+04         7.44E-03 breach)
Large isolation failures (liner 3b                                                1.19E-07       0.69%           1.56E+05         1.85E-02 breach)
Small isolation failures - failure to 4                                                     1           1               1               1 seal (type B)
Small isolation failures - failure to 5                                                     1           1               1               1 seal (type C)
Containment isolation failures 6         (dependent failure, personnel          1.15E-06       6.66%           8.85E+06        1.01E+01 errors)
Severe accident phenomena 7                                                1.23E-06       7.16%           2.62E+06        3.22E+00 induced failure (early and late) 8              Containment bypass                7.03E-07       4.09%           2.45E+07        1.73E+01 Total                    1.72E-05                                       3.07E+01
: 1. represents a probabilistically insignificant value.
: 2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.
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:
15 2 K        DMN   =                         2         2     4      4       5     5 3 217
        =5            1.55E-5 = 7.16E-7 15 . 5 K     C DMN    =                         2         2     4      4       5     5 3 218
                  .D
        =5            1.55E-5 = 1.78E-7 E
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54002-CALC-01                                  Evaluation of Risk Significance of Permanent ILRT Extension The results of the calculation for a 15-year interval are presented in Table 5-12.
Table 5 Risk Profile for Once in 15 Year ILRT Class              Description               Frequency       Contribution   Population Dose       Population
(/yr)          (%)         (person-rem)          Dose Rate (person-rem/yr) 1          No containment failure2          1.32E-05         76.89%           1.56E+03            2.06E-02 Large containment isolation 2                                           0.00E+00         0.00%            2.22E+06           0.00E+00 failures Small isolation failures (liner 3a                                          7.16E-07         4.16%            1.56E+04            1.12E-02 breach)
Large isolation failures (liner 3b                                          1.78E-07         1.04%            1.56E+05            2.78E-02 breach)
Small isolation failures - failure 4                                                 1            1                1                  1 to seal (type B)
Small isolation failures - failure 5                                                 1            1                1                  1 to seal (type C)
Containment isolation failures 6       (dependent failure, personnel       1.15E-06         6.66%            8.85E+06           1.01E+01 errors)
Severe accident phenomena 7                                           1.23E-06         7.16%            2.62E+06           3.22E+00 induced failure (early and late) 8           Containment bypass              7.03E-07         4.09%            2.45E+07           1.73E+01 Total                 1.72E-05                                             3.07E+01
: 1. represents a probabilistically insignificant value.
: 2. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.
5.2.4    Step 4 - Determine the Change in Risk in Terms of LERF The risk increase associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from an intact containment could, in fact, result in a larger release due to the increase in probability of failure to detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.
Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 [Reference 4] defines very small changes in risk as resulting in increases of CDF less than 10-6/year and increases in LERF less than 10-7/year, and small changes in LERF as less than 10-6/year. Since containment overpressure is not required in support of ECCS performance to mitigate design basis accidents at RNP, the ILRT extension does not impact CDF. Therefore, the relevant risk-impact metric is LERF.
For RNP, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a 10-year test interval from Table 5-11, the Class 3b frequency is 1.19E-7/year; based on a 15-year test interval from Table 5-12, the Class 3b frequency is 1.78E-7/year. Thus, the increase in the overall probability of LERF due to Class 3b sequences that is due to increasing the ILRT test interval from 3 to 15 years is 1.42E-7/year. Similarly, the increase due to increasing the interval from 10 to 15 years is 5.94E-8/year. As can be seen, even with the conservatisms included in the evaluation (per Revision 0A                                                                                         Page 22 of 79

54002-CALC-01                              Evaluation of Risk Significance of Permanent ILRT Extension the EPRI methodology), the estimated change in LERF is below the threshold criteria for a small change when comparing the 15-year results to the current 10-year requirement, and slightly greater than the criteria when compared to the original 3-year requirement. Table 5-13 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)              3.56E-08            1.19E-07         1.78E-07 LERF (3 year baseline)                                 8.31E-08         1.42E-07 LERF (10 year baseline)                                                   5.94E-08 The increase in the overall probability of LERF due to Class 3b sequences being slightly greater than 10-7 is not unexpected. Since the target is exceeded, some refinement is necessary. One method to remove some conservatism is to examine the source term expected to be available for release during the accident sequence. The source term is greatly reduced if the debris expelled from the reactor remains covered with water. Therefore, if the accident sequence contains containment spray success, the source term is not considered to lead to a large early release. Excluding scenarios where containment spray is credited and therefore scrubbing the source term release results in a frequency reduction. Section 7.5 in Reference 32 describes the RNP plant damage states and the containment systems that are failed and successful in each end state. Containment safeguards event tree end states A, B, C, G, H, and I encapsulate scenarios that have successful containment spray during both the injection and recirculation phases of the accident. Since release classes RC-2, RC-2B, RC-4, RC-4C, RC-5, and RC-5C have already been subtracted from CDF when calculating Class 3 frequencies, frequency was not redundantly removed in these release categories for containment spray success. To refine the Class 3 calculations by removing the frequency of scenarios with containment spray success, the frequency of the scenarios in release categories IC-1, RC-1, RC-1A, RC-1B, RC-1BA, RC-3, and RC-3B and end states A, B, C, G, H, and I are summed and removed from the Class 3 frequencies. Table 5-14 describes each end state.
Table 5 Summary of RNP Containment Safeguards Event Tree End States End State    Containment Isolation      Containment Sprays      Containment Fans A         Successful isolation          Functioning          Functioning B        Small isolation failure        Functioning          Functioning C        Large isolation failure        Functioning          Functioning D         Successful isolation      Fail in recirculation    Functioning E        Small isolation failure    Fail in recirculation    Functioning F        Large isolation failure    Fail in recirculation    Functioning G         Successful isolation          Functioning             Failed H        Small isolation failure        Functioning             Failed I        Large isolation failure        Functioning             Failed J         Successful isolation      Fail in recirculation       Failed K        Small isolation failure    Fail in recirculation       Failed L        Large isolation failure    Fail in recirculation       Failed Revision 0A                                                                                     Page 23 of 79

54002-CALC-01                                 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Summary of RNP Containment Safeguards Event Tree End States End State    Containment Isolation          Containment Sprays       Containment Fans M          Successful isolation            Fail in injection         Functioning N         Small isolation failure          Fail in injection         Functioning O         Large isolation failure          Fail in injection         Functioning P          Successful isolation            Fail in injection           Failed Q         Small isolation failure          Fail in injection           Failed R         Large isolation failure          Fail in injection           Failed Table 5-15 shows adjusted release category frequencies after some conservatisms from containment spray success excluded.
Table 5 Adjusted Class Frequencies Class                                 Description                                          Frequency (/yr) 1                            No containment failure1                                       1.40E-05 2                      Large containment isolation failures                                0.00E+00 3a                     Small isolation failures (liner breach)                              9.41E-08 3b                     Large isolation failures (liner breach)                              2.34E-08 4                Small isolation failures - failure to seal (type B)                       0.00E+00 5                Small isolation failures - failure to seal (type C)                       0.00E+00 Containment isolation failures (dependent failure, personnel 6                                                                                          1.15E-06 errors) 7          Severe accident phenomena induced failure (early and late)                      1.23E-06 8                              Containment bypass                                          7.03E-07 Total (CDF)                                                                                      1.72E-05
: 1. The Class 1 frequency is reduced by the frequency of Class 3a and Class 3b in order to preserve total CDF.
Substituting these values into the previously defined equations and calculation method yields the final results displayed in Table 5-16.
Table 5 Impact on LERF due to Extended Type A Testing Intervals with Adjusted CDF ILRT Inspection Interval             3 Years (baseline)          10 Years         15 Years Class 3b (Type A LERF)                    2.34E-08              7.80E-08         1.17E-07 LERF (3 year baseline)                                         5.46E-08         9.36E-08 LERF (10 year baseline)                                                           3.91E-08 The adjusted containment spray inputs allow the values to be less than the 1E-7 LERF metric.
The delta LERF between the 3 years and the 15 years is 9.36E-8/yr. This value shows that the proposed extension meets the definition of a very small change in risk as defined in Regulatory Guide 1.174. Also, as explained in Section 5.2.1, since this calculation uses the results from the PDS quantification that contains double counting of risk, results are conservative.
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54002-CALC-01                             Evaluation of Risk Significance of Permanent ILRT Extension 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:
: J :
       =1 where f(ncf) is the frequency of those sequences that do not result in containment failure; this frequency is determined by summing the Class 1 and Class 3a results.
Since CCFP is only concerned with a containment failure and not whether the release is small or large, the Class 1 results without refinement must be used to calculate the CCFP. Table 5-17 shows the steps and results of this calculation. The difference in CCFP between the 3-year test interval and 15-year test interval is 0.829%.
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)             1.41E-05         1.40E-05          1.39E-05 f(ncf)/CDF                  0.819            0.814             0.810 CCFP                     0.181            0.186             0.190 CCFP (3 year baseline)                            0.483%            0.829%
CCFP (10 year baseline)                                              0.345%
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 0.829%. 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 base-mat, each with a steel liner.
The following approach is used to determine the change in likelihood, due to extending the ILRT, of detecting corrosion of the containment steel liner. This likelihood is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:
Differences between the containment base-mat and the containment cylinder and dome:
The historical steel liner flaw likelihood due to concealed corrosion The impact of aging The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw Revision 0A                                                                                 Page 25 of 79

54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension Assumptions Consistent with the Calvert Cliffs analysis, a half failure is assumed for base-mat concealed liner corrosion due to the lack of identified failures (See Table 5-18, 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 5-18, 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 5-18, Steps 2 and 3). Sensitivity studies are included that address doubling this rate every ten years and every two years.
In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere, given that a liner flaw exists, was estimated as 1.1% for the cylinder and dome, and 0.11% (10% of the cylinder failure probability) for the base-mat. These values were determined from an assessment of the probability versus containment pressure. For RNP, the ILRT maximum pressure is 42 psig [References 27] and ultimate pressure of 130 psig [Reference 39]. Probabilities of 1% for the cylinder and dome, and 0.1% for the base-mat are used in this analysis, and sensitivity studies are included in Section 5.3.2 (See Table 5-18, Step 4).
Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the base-mat region is considered to be less likely than the containment cylinder and dome region (See Table 5-18, Step 4).
Although the Calvert Cliffs analysis uses a 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10%, the RNP containment is insulated such that complete visual inspection is not possible.
Approximately 74% of the RNP containment is covered by insulation and is not readily visible. The remaining 26% which is visible is examined as required by ASME Section XI, Subsection IWE. Therefore, a 77% visual inspection failure likelihood is used (See Table 5-18, 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.
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54002-CALC-01                                 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Steel Liner Corrosion Base Case Step                 Description                      Containment Cylinder and             Containment Basemat Dome (85%)                             (15%)
Historical liner flaw likelihood           Events: 2                             Events: 0 Failure data: containment location         (Brunswick 2 and North Anna 2)        Assume a half failure specific                                   2 / (70 x 5.5) = 5.19E-03             0.5 / (70 x 5.5) = 1.30E-03 1     Success data: based on 70 steel-lined containments and 5.5 years since the 10CFR 50.55a requirements of periodic visual inspections of containment surfaces Year                   Failure rate   Year               Failure rate Aged adjusted liner flaw likelihood During the 15-year interval, assume        1                      2.05E-03       1                  5.13E-04 failure rate doubles every five years      average 5-10           5.19E-03       average 5-10       1.30E-03 2                                                15                     1.43E-02       15                 3.57E-03 (14.9% increase per year). The average for the 5th to 10th year set to the historical failure rate.            15 year average = 6.44E-03            15 year average = 1.61E-03 Increase in flaw likelihood between 3 and 15 years Uses aged adjusted                   0.73% (1 to 3 years)               0.18% (1 to 3 years) 3     liner flaw likelihood (Step 2),                    4.18% (1 to 10 years)              1.04% (1 to 10 years) assuming failure rate doubles every                9.66% (1 to 15 years)              2.41% (1 to 15 years) five years.
Likelihood of breach in containment 4                                                                  1%                                0.1%
given liner flaw 77%
Approximately 74% of the containment cylinder and dome liner is covered by insulation and is not readily visible. It is assumed the remaining approximately 26%
Visual inspection detection failure        which is visible would be visually                 100%
5 likelihood                                 examined. A 5% probability is         Cannot be visually inspected then assumed that visible failures would not be detected, and a 5%
probability was assumed that flaws could exist but would not be visible.
0.74 + (0.26 * (0.05 + 0.05))
0.0056% (3 years)                     0.000180% (3 years) 0.73% x 1% x 77%                      0.18% x 0.1% x 100%
Likelihood of non-detected                 0.0322% (10 years)                    0.00104% (10 years) 6     containment leakage (Steps 3 x 4 x
: 5)                                         4.18% x 1% x 77%                      1.04% x 0.1% x 100%
0.0744% (15 years)                    0.00241% (15 years) 9.66% x 1% x 77%                      2.41% 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 base-mat, as summarized below for RNP.
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54002-CALC-01                              Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Total Likelihood on Non-Detected Containment Leakage Due to Corrosion for RNP Description At 3 years: 0.00562% + 0.000180% = 0.00580%
At 10 years: 0.03279% + 0.00104% = 0.03323%
At 15 years: 0.07438% + 0.00241% = 0.07679%
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 a 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 NRG 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
[Reference 28]. 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.
Robinson is transitioning to NFPA 805 licensing basis for fire protection and submitted a License Amendment Request (LAR) on September 16, 2013 [Reference 38]. This transition included performing a Fire PRA and installing modifications to reduce the fire-induced CDF and LERF to those reported in the NFPA 805 LAR. It is anticipated that many, but not all, of the NFPA 805 modifications will be completed by the next scheduled ILRT. All the modifications will be installed prior to coming out of refueling outage 31 in 2018; the next scheduled ILRT is Revision 0A                                                                               Page 28 of 79

54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension February 2017. All of the modifications that have an appreciable impact on reducing overall fire CDF have been installed. The remaining modifications are of lower risk significance [Attachment S of Reference 38]. Compensatory measures have been implemented to reduce the fire risk until the modifications are implemented. These measures may be actions to reduce fire initiating event probabilities, actions to improve suppression probability, and/or actions to recover or protect systems that mitigate core damage and large early release accident sequences.
Therefore, the Fire PRA model is deemed applicable for this calculation.
The Fire PRA model RNP_12Fh_s1 was used to obtain the fire CDF and LERF values. To reduce conservatism in the model, the methodology of subtracting existing LERF from CDF is also applied to the Fire PRA model. The following shows the calculation for Class 3b:
0.5 C =         C          OG     =       1.80G-5  3.76G-6 = 3.26G-8 218 L                                 L     L.D C LMN =             C        OG      =              1.80G-5  3.76G-6 = 1.09E-7 E
D                                     L.D P      C DMN =              C       OG       =5            1.80G-5  3.76G-6 = 1.63E-7 E
The Seismic PRA results from the IPEEE Seismic Margins Analysis do not result in estimate of CDF [Reference 35]. The 2014 Seismic Reevaluations for operating reactor sites [Reference 42] state the conclusions reached in 2010 by GI-199 [Reference 41] remain valid for estimating Seismic CDF at plants in the Central and Eastern United States, which includes RNP. The most conservative Seismic CDF reported in Table D-1 of Reference 41 is 1.4E-5. Applying the internal event LERF/CDF ratio to the seismic CDF yields an estimated seismic LERF of 1.36E-6, as shown by the equation below.
LERFSeismic  CDFSeismic
* LERFIE / CDFIE = 1.4E-5
* 1.67E-6 / 1.72E-5 = 1.36E-6 Again subtracting LERF from CDF, the Class 3b frequency can be calculated by the following formulas:
L.D C =         C          OG     =   E 1.4E-5-1.36E-6 = 2.90E-8 L                               L    L.D C LMN =           C        OG      =           1.4E-5-1.36E-6 = 6.67E-8 E
D                               D    L.D C DMN =           C        OG      =           1.4E-5-1.36E-6 = 1.45E-7 E
The High Wind PRA results from the IPEEE [Section 5.3.6 of Reference 35] report a CDF of 2E-
: 6. Applying the internal event LERF/CDF ratio to the high wind CDF yields an estimated high wind LERF of 1.94E-7. Again subtracting LERF from CDF, the Class 3b frequency can be calculated by the following formulas:
L.D C =         C          OG     =      2.0E-6-1.94E-7 = 4.14E-9 E
L                               L    L.D C LMN =           C        OG      =           2.0E-6-1.94E-7 = 1.38E-8 E
D                               D    L.D C DMN =           C        OG      =           2.0E-6-1.94E-7 = 2.07E-8 E
The fire, seismic, and high wind contributions to Class 3b frequencies are then combined to obtain the total external event contribution to Class 3b frequencies. The change in LERF is Revision 0A                                                                                 Page 29 of 79

54002-CALC-01                                 Evaluation of Risk Significance of Permanent ILRT Extension calculated for the 1 in 10 year and 1 in 15 year cases and the change defined for the external events in Table 5-20.
Table 5 RNP 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            6.58E-08              2.19E-07               3.29E-07               2.63E-07 Internal Events           2.34E-08              7.80E-08               1.17E-07               9.36E-08 Combined               8.92E-08              2.97E-07               4.46E-07               3.57E-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 3.57E-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 + LERFHW + LERFclass3Bincrease
         = 1.67E-6/yr + 3.76E-6/yr + 1.36E-6/yr + 1.94/yr + 3.40E-7/yr = 7.33E-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.
5.3.2      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 5-23. Since CCFP is only concerned with a containment failure and not whether the release is small or large, the Class 1 results without containment spray refinement is used to calculate the CCFP. The Annual Dose Rate calculations are performed using the containment spray adjustments. 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               3b              3b             LERF            LERF             LERF Frequency         Frequency       Frequency         Increase        Increase         Increase (3-per-10        (1-per-10       (1-per-15      (3-per-10 to    (3-per-10 to     (1-per-10 to year ILRT)       year ILRT)      year ILRT)        1-per-10)       1-per-15)        1-per-15)
Internal Event 3B           2.34E-08         7.81E-08        1.17E-07         5.47E-08        9.38E-08         3.91E-08 Contribution Corrosion Likelihood         2.52E-08         1.11E-07        2.33E-07         8.63E-08        2.08E-07         1.21E-07 X 1000 Revision 0A                                                                                          Page 30 of 79

54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension Table 5 Steel Liner Corrosion Sensitivity Case: 3B Contribution 3b               3b               3b            LERF           LERF          LERF Frequency        Frequency       Frequency         Increase       Increase      Increase (3-per-10        (1-per-10        (1-per-15     (3-per-10 to   (3-per-10 to  (1-per-10 to year ILRT)       year ILRT)       year ILRT)       1-per-10)      1-per-15)     1-per-15)
Corrosion Likelihood    4.09E-08         4.12E-07         1.28E-06        3.71E-07       1.24E-06      8.64E-07 X 10000 Corrosion Likelihood    1.99E-07         3.42E-06         1.17E-05        3.22E-06       1.15E-05      8.29E-06 X 100000 Table 5 Steel Liner Corrosion Sensitivity: CCFP CCFP           CCFP          CCFP CCFP             CCFP             CCFP          Increase       Increase      Increase (3-per-10        (1-per-10        (1-per-15     (3-per-10 to   (3-per-10 to  (1-per-10 to year ILRT)       year ILRT)       year ILRT)       1-per-10)      1-per-15)     1-per-15)
Baseline 1.81E-01         1.86E-01         1.90E-01        4.83E-03       8.29E-03      3.45E-03 CCFP Corrosion Likelihood    1.81E-01         1.87E-01         1.90E-01        5.19E-03       8.91E-03      3.71E-03 X 1000 Corrosion Likelihood    1.83E-01         1.91E-01         1.97E-01        8.45E-03       1.45E-02      6.03E-03 X 10000 Corrosion Likelihood    1.97E-01         2.38E-01         2.67E-01        4.10E-02       7.03E-02      2.93E-02 X 100000 Table 5-23 -Steel Liner Corrosion Sensitivity: Does Rate Dose Rate      Dose Rate     Dose Rate Dose Rate      Dose Rate        Dose Rate        Increase       Increase      Increase (3-per-10      (1-per-10        (1-per-15     (3-per-10 to   (3-per-10 to  (1-per-10 to year ILRT)     year ILRT)       year ILRT)       1-per-10)      1-per-15)     1-per-15)
Dose Rate       3.65E-03       1.22E-02         1.83E-02        8.53E-03       1.46E-02      6.10E-03 Corrosion Likelihood X    3.93E-03       1.74E-02         3.63E-02        1.35E-02       3.24E-02      1.90E-02 1000 Corrosion Likelihood X    6.38E-03       6.43E-02         1.99E-01        5.79E-02       1.93E-01      1.35E-01 10000 Corrosion Likelihood X    3.10E-02       5.34E-01         1.83E+00        5.03E-01       1.80E+00     1.29E+00 100000 Revision 0A                                                                                Page 31 of 79

54002-CALC-01                             Evaluation of Risk Significance of Permanent ILRT Extension 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 5-24 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 RNP Summary of ILRT Extension Using Expert Elicitation Values (from Reference 24)
Leakage Size (La)     Jeffreys Non-Informative      Expert Elicitation Mean         Percent Reduction Prior              Probability of Occurrence 10                    2.70E-02                     3.88E-03                       86%
100                    2.70E-03                     9.86E-04                       64%
Taking the baseline analysis and using the values provided in Tables 5-12 and 5-13 for the expert elicitation sensitivity yields the results in Table 5-25.
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54002-CALC-01                                 Evaluation of Risk Significance of Permanent ILRT Extension Table 5 RNP Summary of ILRT Extension Using Expert Elicitation Values Accident                                                  ILRT Interval Class 3 per 10 Years                         1 per 10 Years             1 per 15 Years Base         Adjusted       Dose        Dose      Frequency       Dose      Frequency        Dose Frequency         Base       (person-       Rate                      Rate                       Rate Frequency        rem)      (person-                  (person-                   (person-rem/yr)                   rem/yr)                    rem/yr) 1        1.41E-05       1.40E-05      1.56E+03     2.19E-02    1.38E-05      2.16E-02     1.37E-05      2.14E-02 2        0.00E+00      0.00E+00       2.22E+06    0.00E+00     0.00E+00     0.00E+00      0.00E+00      0.00E+00 3a          N/A         6.40E-08      1.56E+04     9.98E-04    2.13E-07      3.33E-03     3.20E-07      4.99E-03 3b          N/A         1.63E-08      1.56E+05     2.54E-03    5.42E-08      8.45E-03     8.13E-08      1.27E-02 6        1.15E-06       1.15E-06      8.85E+06    1.01E+01     1.15E-06     1.01E+01      1.15E-06      1.01E+01 7        1.23E-06       1.23E-06      2.62E+06    3.22E+00     1.23E-06     3.22E+00      1.23E-06      3.22E+00 8        7.03E-07       7.03E-07      2.45E+07    1.73E+01     7.03E-07     1.73E+01      7.03E-07      1.73E+01 Totals     1.72E-05       1.72E-05      3.84E+07    3.07E+01     1.72E-05     3.07E+01      1.72E-05      3.07E+01 LERF (3 per 10                            N/A                                 3.79E-08                   6.50E-08 yrs base)
LERF (1 per 10                            N/A                                    N/A                     2.71E-08 yrs base)
CCFP                            18.01%                                  18.23%                     18.39%
The results illustrate how the expert elicitation reduces the overall change in LERF and the overall results are more favorable with regard to the change in risk.
5.3.4     Large Leak Probability Sensitivity Study The large leak probability is a vital portion of determining the Class 3b frequency. A sensitivity is done by using the WCAP method as an alternative to the EPRI method for calculating the large leak probability. Table 5-26 presents the large leak probabilities for the baseline test, 10 year test interval, and 15 year test interval. Table 5-27 was developed using the same process as to calculate Class 3b.
Table 5 RNP Large Leak Probabilities Using the WCAP Method Test Interval        WCAP Large Leak Probability           EPRI Accident Class 3b Frequency 3 per 10 years                    2.47E-4                                  3.46E-09 10 years                       7.41E-4                                  1.02E-08 15 years                       1.11E-3                                  1.50E-08 Using the same EPRI approach, but with an updated Class 3b frequency calculated from the WCAP large leak probability data, Table 5-27 contains the final results.
Table 5 Impact on LERF due to Extended Type A Testing Intervals with WCAP CDF ILRT Inspection Interval           3 Years (baseline)       10 Years         15 Years Class 3b (Type A LERF)                  3.46E-09           1.02E-08         1.50E-08 LERF (3 year baseline)                                    6.71E-09         1.16E-08 LERF (10 year baseline)                                                     4.85E-09 These results demonstrate that the EPRI methodology is conservative when used to calculate a large leak probability as compared to the WCAP method.
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54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension 6.0      RESULTS The results from this ILRT extension risk assessment for RNP are summarized in Table 6-1.
Table 6-1 -ILRT Extension Summary Class       Dose              Base Case                    Extend to              Extend to (person-           3 in 10 Years               1 in 10 Years          1 in 15 Years rem)
CDF/Year          Person-      CDF/Year      Person-  CDF/Year      Person-Rem/Year                   Rem/Year               Rem/Year 1       1.56E+03     1.40E-05         2.18E-02     1.37E-05      2.14E-02 1.35E-05     2.11E-02 2       2.22E+06    0.00E+00         0.00E+00      0.00E+00      0.00E+00 0.00E+00     0.00E+00 3a      1.56E+04     9.41E-08         1.47E-03     3.14E-07      4.89E-03 4.70E-07     7.34E-03 3b      1.56E+05     2.34E-08         3.65E-03     7.80E-08      1.22E-02 1.17E-07     1.83E-02 6       0.00E+00     1.15E-06        1.01E+01      1.15E-06      1.01E+01 1.15E-06     1.01E+01 7       0.00E+00     1.23E-06        3.22E+00      1.23E-06      3.22E+00 1.23E-06     3.22E+00 8       8.85E+06     7.03E-07        1.73E+01      7.03E-07      1.73E+01 7.03E-07     1.73E+01 Total                  1.72E-05        3.07E+01      1.72E-05      3.07E+01 1.72E-05     3.07E+01 ILRT Dose Rate from 3a and 3b From 3 N/A                       1.15E-02               1.97E-02 Total      Years Dose Rate    From 10 N/A                          N/A                 8.23E-03 Years From 3 N/A                       0.0376%                0.0644%
   %Dose       Years Rate      From 10 N/A                          N/A                 0.0268%
Years 3b Frequency (LERF)
From 3 N/A                       5.46E-08               9.36E-08 Years LERF From 10 N/A                          N/A                 3.90E-08 Years CCFP %
From 3 N/A                        0.318%                 0.545%
Years CCFP%
From 10 N/A                          N/A                  0.227%
Years Revision 0A                                                                               Page 34 of 79

54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension

==7.0      CONCLUSION==
S AND RECOMMENDATIONS Based on the results from Section 5.2 and the sensitivity calculations presented in Section 5.3, the following conclusions regarding the assessment of the plant risk are associated with extending the Type A ILRT test frequency to 15 years:
Regulatory Guide 1.174 [Reference 4] provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.174 defines very small changes in risk as resulting in increases of CDF less than 1.0E-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 9.36E-8/year using the EPRI guidance. As such, the estimated change in LERF is determined to be very small using the acceptance guidelines of Regulatory Guide 1.174 [Reference 4].
The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.020 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 from the 3 in 10 year interval to 1 in 15 year interval is 0.545%. 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 RNP risk profile.
Previous Assessments The NRC in NUREG-1493 [Reference 6] has previously concluded that:
Reducing the frequency of Type A tests (ILRTs) from 3 per 10 years to 1 per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.
Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the ILRT frequency beyond 1 in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, ILRTs also test integrity of the containment structure.
The findings for RNP confirm these general findings on a plant-specific basis considering the severe accidents evaluated for RNP, the RNP containment failure modes, and the local population surrounding RNP.
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54002-CALC-01                            Evaluation of Risk Significance of Permanent ILRT Extension A.       ATTACHMENT 1 Internal Events PRA Quality Statement for Permanent 15-Year ILRT Extension The RNP Internal Events PRA model RNP_12 is utilized to calculate CDF, LERF, and PDSs for the permanent 15-year ILRT extension. Any elements of the supporting requirements detailed in ASME/ANS RA-Sa-2009 that could be significantly affected by the application are required to meet Capability Category II requirements.
The Internal Events PRA provides an adequate base model for the development of the permanent 15-year ILRT extension. While there has been one instance of a change to the internal events model since the 2010 peer review that could be considered a model upgrade, this change is specifically applicable to the loss of CCW initiating event and does not affect the permanent 15-year ILRT extension results.
To ensure that the current Internal Events PRA model remains an accurate reflection of the as-built, as-operated plant, the following configuration control activities are routinely performed:
Design changes and procedure changes are reviewed for their impact on the Internal Events PRA model. PRA screening is required for all design and procedure changes.
New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.
Plant-specific initiating event frequencies, failure rates, and maintenance unavailabilities are updated based upon reviews of plant program data, particularly data supporting the Maintenance Rule.
The Robinson Internal Events PRA is based on a detailed model of the plant developed from the Individual Plant Examination for Generic Letter 88-20, "Individual Plant Examination for Severe Accident Vulnerabilities." The model is maintained and updated in accordance with Robinson procedures, and has been updated to meet the ASME PRA Standard and Revision 2 of Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities."
In accordance with RG 1.200, Rev. 2 and ASME/ANS RA-Sa-2009, a full scope Peer Review of the Internal Events model of the RNP PRA was conducted in May 2010 by six PRA experts (Attachment U of Reference 38). The review provided findings and suggestions regarding the model and identified 25 supporting requirements within the internal events portion of the model that did not meet Capability Category II. These findings were either resolved by additional analysis and included in the quantitative results, or evaluated with their impact on the applicable risk evaluation. The dispositions for these findings are presented in Table 1, Internal Events PRA Peer Review - Facts and Observations. All findings which had significant impact on this analysis have been addressed. The ILRT application was determined to be an application requiring a Capability Category II PRA model per the Regulatory Guide 1.200 criteria, Revision
: 2. This is based on the requirement for numerical results for CDF and LERF to determine the risk impact of the requested change and the fact that this change is risk-informed, not risk-based. Table 1 includes discussion of all findings from the industry peer review along with the assessment and evaluation of the finding that shows that they have either been addressed or have no material impact on the ILRT interval extension request.
Following disposition, all of the SRs evaluated Met Capability Category II or better.
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54002-CALC-01                           Evaluation of Risk Significance of Permanent ILRT Extension Fire PRA Quality Statement for Permanent 15-Year ILRT Extension In accordance with RG 1.205 position 4.3:
The licensee should submit the documentation described in Section 4.2 of Regulatory Guide 1.200 to address the baseline PRA and application-specific analyses. For PRA Standard supporting requirements important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable.
The Robinson Internal Events model is also updated to support the Robinson Fire PRA. The RNP Combined Internal Events and Fire PRA was peer reviewed during the period of March 2013 (Attachment V of Reference 38). The peer review was conducted by a team of industry personnel (utility and vendor). The Westinghouse Owners Group performed the review and has documented the outcome via LTR-RAM-13-06 Fire PRA Peer Review of the H. B. Robinson Nuclear Plant Fire Probabilistic Risk Assessment against the Fire PRA Standard Supporting Requirements from Section 4 of the ASME/ANS Standard. As noted in LTR-RAM-13-06, the RNP Fire PRA was found to be consistent with the ASME/ANS Standard and is suitable for supporting risk-informed applications.
The peer review team noted a number of Facts and Observations (F&Os). As documented in LTR-RAM-13-06, 85% of the Supporting Requirements (SRs) were assessed at Capability Category II or higher. Approximately eighteen Finding level and nine Suggestion level F&Os were identified during the peer review conducted in March 2013. Duke Energy recognized that CDF and LERF were relatively high, as noted in LTR-RAM-13-06. Based on the CDF and LERF values at the time of the initial peer review, coupled with the number of findings associated with the Fire Scenario Selection (FSS) Technical Element (18), Duke Energy decided to have a focused peer review.
The focused peer review was conducted during the period of July 2013 and evaluated the FSS Technical Element based on refinements to approved methodologies and updated documentation. The focused peer review was conducted by Frederick Mowrer (C P Fire, LLC) and Bijan Najafi (JENSEN HUGHES, Inc.) and is documented via JENSEN HUGHES Calculation No. 0004-0042-415-RPT-001, Robinson Nuclear Plant Fire PRA Focused Peer Review, Revision 0. As noted in LTR-RAM-13-06 and JENSEN HUGHES Calculation No. 0004-0042-415-RPT-001, the Fire PRA does apply the methodologies outlined in NUREG/CR-6850 correctly, is consistent with the ASME/ANS Standard and is applicable for supporting risk-informed applications. Although several of the initial F&Os were resolved, seven new findings and three new suggestions were identified during the focused-scope peer review.
Table 2 documents the Finding level F&Os associated with both the initial and focused peer reviews. All but two findings have been dispositioned as meeting Capability Category II or better. FSS-E3-01, which was assessed as meeting Category I, involved providing a mean value of, and statistical representation of, the uncertainty intervals for the parameters used for modeling the significant fire scenarios. Although no change has yet been made that would improve the Capability Category assessment, RNP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because documenting the statistical representation of uncertainty intervals will not change the quantified risk metrics. Supporting Requirement FSS-E1-01 has not been resolved, but this is a documentation issue that does not impact quantification. These two SRs not being resolved does not impact the ILRT extension analysis. The Fire PRA is adequate to support the ILRT extension.
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54002-CALC-01                                                                        Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic           Status                    Finding/Observation                                        Disposition ID)
AS-A5 (1)          Define the Accident  Dispositioned RNP-F/PSA-0043, Tables 4.2, 4.4, 4.5, 4.6, 4.7,     This is consistent with FRP-H.1 where Step 73a
[Cat I-III is MET] Sequence Model                     4.8, & 4.10 correlate safety function and           checks for SI pumps running and if not, the procedure success criteria to procedures and supporting       requires to go to Step 76 to verify pressurizer PORVs documents; The transient event tree (Figure         are all closed. If PORVs are not closed, the procedure 4.3) maps the progression of plant response to      directs to close associated PORV Block Valves.
an initiating event as is presented in the ESD. Therefore, description of sequence TBU in RNP-F/PSA-0043 section 4.2.3.5 is re-written to remove assumption.
Description of sequence TBU, in 0042, Section 4.2.3.5, "For sequences involving injection failure, it is assumed that the operators do not open the PORVs following the failure of safety injection. The RCS pressure increases until the pressurizer relief valves lift. The valves cycle to relieve pressure but successfully reclose. The potential for valve failure is addressed in the next sequence." Is this consistent with procedures and bounding for consequences?
AS-A5 (2)          Define the Accident  Dispositioned RNP-F/PSA-0043, Tables 4.2, 4.4, 4.5, 4.6, 4.7,     Transient-induced RCP seal failure results in transient
[Cat I-III is MET] Sequence Model                     4.8, & 4.10 correlate safety function and           similar to an S1 LOCA and is usually limiting in that an success criteria to procedures and supporting       S1 LOCA requires inventory makeup and secondary documents; The transient event tree (Figure         side heat removal. It is assumed that a small 4.3) maps the progression of plant response to      percentage of seal LOCA develops that would require an initiating event as is presented in the ESD. inventory makeup and heat removal using Safety Injection because of high leak rates, but the large breaks would occur after at least a 15 minute delay Describe how the range of seal LOCA break           resulting in reduction in transient severity. Section 4.4A flows, which can exceed small LOCA break flow       of RNP-F/PSA-0043 has been revised to explicitly were modeled.                                       include this assumption. Additionally, this assumption and uncertainty is captured in RNP-F/PSA-0074, RNP Uncertainty Analysis, for risk impact determination.
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54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic            Status                   Finding/Observation                                          Disposition ID)
AS-A5 (3)          Define the Accident    Dispositioned RNP-F/PSA-0043, Section 4.8, Consequential          An Induced SGTR model, based on the guidance in
[Cat I-III is MET] Sequence Model                       Pressure-induced SGTR not considered.               NUREG-1570, Risk Assessment of Severe "Consequential tube ruptures due to high            Accident-Induced Steam Generator Tube Rupture,
[Associated with                                        primary-to-secondary differential pressures         USNRC, March 1998, was developed and documented AS-B3]                                                  (e.g., due to secondary line breaks or ATWS         in RNP-F/PSA-0047. Sections 4.8 of RNP-F/PSA-0043 events) are not explicitly considered in the        and the Level 2 results have been updated to reflect model. Secondary line breaks can result in a        the induced SGTR model and results. More detailed primary-to-secondary differential pressure          modeling of Induced SGTR was added to the level 2 equivalent to RCS pressure."                        analysis as described in section 8.2.2 of RNP-F/PSA-0047, Rev. 2.
AS-B3 (1)          Identify               Dispositioned Section 7.5 of the system notebook addresses        A review of the systems and the impact of the initiating
[Cat I-III is NOT  phenomenological                     the requirements of SRs AS-B1 and AS-B3.            events on those systems were performed to ensure MET]               conditions created by                Based on a review of these section of several       that the impacts are captured in the model as the accident                         system notebooks, this reviewer did not identify    appropriate. A statement discussing the impact of the
[Associated with   progression                          any components or operator action that have         initiators on the system was added to section 7.1 of the AS-B1 and SY-                                           been considered to be impacted by any               system notebook for each system.
B14]                                                    initiating event (or accident sequences associated with the initiating event). Based on a   RESOLUTION OF CAPABILITY CATEGORY discussion with the Duke Energy's Lead PRA          CLASSIFICATION engineer, it is understood that the potential       Based on the disposition above, SR AS-B3 is impact of an IE on the systems credited to          considered to be MET at CAT I-III.
mitigate the consequences of the IE have been addressed. However, the discussions provided in the pertinent sections of the system notebooks are not adequate to provide a defensible and easily traceable documentation of the justification of 'no impact.' Given the relatively compact physical structure of the plant, housing of redundant components in the same location, and relatively high dependency on the operator action, a better documentation should be provided to support the 'no impact' assertion. For example, the potential impact of the SLB outside containment IE on the AFW function and the non-emergency buses should Revision 0A                                                                                                                                    Page 39 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic              Status                   Finding/Observation                                         Disposition ID) be included. Another potentially vulnerable system is the Fire Protection system, where the component of such a system is usually are not credited to work in harsh environment. The system notebook and the plant walk down documentation are thought to provide justification that this issue has been looked at.
Better documentation of the supporting analyses is needed to support the assertion of
                                                         'no impact'.
AS-C1 (1)          Correlation of system   Dispositioned RNP-F/PSA-0043 provides a reasonable               For each accident sequence events in
[Cat I-III is MET] top events and                        notebook with clear tables and discussions of      RNP-F/PSA-0043 success criteria tables, system top Accident Sequence                     work performed. However, it is difficult to        event gate names are added in the Supporting safety function events                correlate between system top events and            Documentation column.
accident sequence safety function events.
AS-C2 (1)          Correlation of Type B   Dispositioned RNP-F/PSA-0006 on Initiating Events identifies     RNP-F/PSA-0014 is added to Document Indexing
[Cat I-III is MET] & C HRA and                           transients and LOCAs covered in Accident           Table and Section 4.0, References, of Accident Sequence                     Sequence Analysis. Operator actions are            RNP-F/PSA-0006.
Analysis                              discussed in tables and narrative, but HRA Type B & C Notebook RNP-F/PSA-0014 is not on reference list. A tie to the HRA notebook is needed for clarity.
AS-C2 (2)          Accident sequence       Dispositioned RNP-F/PSA-0006 on Initiating Events identifies     The MTC values selected from WCAP-15831-P and
[Cat I-III is MET] development                           transients and LOCAs covered in Accident           listed in Table 4.9 of RNP-F/PSA-0043 for the ATWS Sequence Analysis. Operator actions are            UET analysis are for a Low Reactivity Core Model with discussed in tables and narrative.                 a maximum HZP MTC of +3.5 pcm/&#xba; F (HFP MTC WCAP-15831-P is used as the basis for              equivalent of -7.42 pcm/&#xba; F). Per ANP-2887 -cycle Moderator Temperature Coefficient (MTC)            reload report for cycle 27, the RNP MTC at HZP is -
Unfavorable Exposure Time (UET). No                0.35 pcm/&#xba; F at BOC and -11.29 pcm/&#xba; F at HFP. The explanation provided as to how this data reflects  WCAP analysis used a bounding MTC that was current plant operation, refueling cycles, or fuel positive. For RNP, the MTC is negative, therefore configuration.                                     bounded by the MTC values of the Low Reactivity Core Model used in the WCAP. Additionally, PSA is on the Revision 0A                                                                                                                                  Page 40 of 79

54002-CALC-01                                                                       Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic           Status                   Finding/Observation                                           Disposition ID)
Reload Review panel and will be evaluating current plant data to existing PSA analysis as indicated in Note under Table 4.9 of RNP-F/PSA-0043.
DA-A2 (1 and 2)   Definitions for data  Dispositioned 1) SSC boundaries are defined in                       Consideration of the AFW room cooling as an air handling
[Cat I-III is NOT analysis                            RNP-F/PSA-0077, Attachment 2. No boundary              unit does not fit the manner in which the components have MET]                                                  definition is defined for HVAC air handling units or   been modeled. Given the system configuration in the plant coolers. The AFW coolers are modeled as                the separation of the components is more appropriate.
separate fans and heat exchangers. These are           NUREG-CR/6928 provides values for the as modeled more appropriately grouped as an air handling unit     configuration and these are utilized.
per NUREG-CR-9628. The failure boundary for the Diesel Generator output breaker is not             The definition of the EDG component boundary is correct consistent with the system analysis modeling and       in that it includes the output breaker. The event the generic failure data. Attachment 2 has the         representing the failure of the output breakers breaker within the component boundary, but the         (PCB5217BNN and PCB5227BNN) to close will be set to breaker is modeled separately. Control room hand       zero in the model since failures of this component are switches are included in the component boundary        rolled into failures of the EDG.
per Attachment 2, even though they are not local as described in NUREG-CR/6968, this appears to         The control circuitry which is included in the NUREG-be a grey area, though. Logic for control valve        CR/6928 component boundary definitions is interpreted to automation is not within the component boundary,       include the control room hand switches as well as local though NUREG-CR/6968 states local                      switches. Therefore the definition of component boundaries instrumentation and controls are within the            as defined in Attachment 2 of the RNP data calculation boundary. For pumps, it appears that separate          RNP-F/PSA-0072 are consistent with the boundary cooling and lubrication systems are not within         definitions identified in NUREG-CR/6928.
scope, though these systems would be included in NUREG-CR/6928 if local. The peer review did not        Component boundaries for control valves as defined in include a comprehensive review of component            Attachment 2 of the RNP data calculation RNP-F/PSA-boundaries. The issues found, though individually      0072 are consistent with the boundary definitions identified not a finding, show a trend of inconsistencies         in NUREG-CR/6928 in that they specify that local switches, between the plant model and component                  contacts, relays, and control circuitry are included in the boundary definitions. 2) No definitions of failure     boundary. Logic for an automatic actuation is not identified modes were found. 3) No definitions of success         in NUREG-CR/6928 definitions and therefore is defined to criteria were found.                                   be outside of the control valve boundary in Attachment 2.
Attachment 2 of the RNP data calculation RNP-F/PSA-0072 specifies that, with regard to pumps, if a Revision 0A                                                                                                                                    Page 41 of 79

54002-CALC-01                                                  Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation      Topic    Status                Finding/Observation                                     Disposition ID) different system is providing additional cooling or lubrication then this component is considered to be separate and not within the boundaries of the pump being considered. NUREG-CR/6928 identifies that only local lubrication or cooling systems are to be included in pump boundaries and Attachment 2 boundary definitions are in agreement with this specification.
With regard to the battery chargers the input and output breakers are modeled separately due to the fact that plant specific data is utilized in the model for battery chargers.
Thus the manner in which the component is modeled is consistent with those definitions put forth in Attachment 2 of the RNP data calculation RNP-F/PSA-0072. Generic data values are utilized to model the breakers.
A review was made of other component boundaries and it was determined the boundaries detailed in Attachment 2 of the data calculation (RNP-F/PSA-0072) and in the model are appropriate when compared to those set forth in NUREG-CR/6928.
Wording has been added to Section 4.3 of the data calculation (RNP-F/PSA-0072) to provide more detailed information on the identification of events as failures. There are definitions in the RNP scoping documents that are used in the evaluation of functional failures for the Maintenance Rule. The MR database is then used to capture all failure events. These events are then evaluated and screening comments provided to determine if they count as failures towards the PSA.
Success criteria are comprehensively defined for each modeled system in the respective system notebooks (RNP-F/PSA-0018).
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54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic            Status                  Finding/Observation                                          Disposition ID)
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR DA-A2 is considered to be MET at CAT I-III.
DA-C1 (1)          Parameter estimation  Dispositioned Some generic data collected from a recognized      The spreadsheet which documents and aggregates the
[Cat I-III is MET]                                     source, NUREG-CR/6928. Other generic data          generic data (GENDATA RNP 2011.xlsx) identifies collected from a Progress Energy source called     which data source contributes for each component and
                                                       'Generic Data Aggregation.' Unable to find         failure mode. This includes the numerical values that documentation of process and methodology           are used for each source. The Section 3.2 of the data used for data collection and analysis for Generic  calculation (RNP-F/PSA-0072) provides references for Data Aggregation.' Generic data is not used for    each contributor to the generic data aggregation.
unavailability. A cursory review of Generic Data Aggregation showed that the failure rates were within reason based on similar failure rates in NUREG-CR/6928.
DA-C14 (1)         Intersystem           Dispositioned No inter-system coincident maintenance             PRA personnel reviewed EOOS entries for instances in
[Cat I-III is MET] Unavailability                      analysis was found to be performed.                which intersystem coincident maintenance occurred.
While examples of such events could be found, none were the result of planned, repetitive maintenance activities. Therefore, using the guidance of the ASME standard (ASME/ANS RA-Sa-2009) it was determined no additional events should be created for these occurrences.
DA-C3 (1)          Plant data screening  Dispositioned Plant specific data originates from Maintenance    The wording in Section 4.3 of the data calculation
[Cat I-III is MET]                                     Rule functional failure data. Justifications are   (RNP-F/PSA-0072) has been adjusted to articulate that provided for screening and disregarding plant-     the MR database is used to capture all failure events.
specific data in RNP_Fail_Data.xls. Section 4.3    These events are then evaluated and screening of RNP-F/PSA-0072 states that "A failure would     comments provided to determine if they count as be counted if a component tripped for no           failures towards the PSA. The statement from Section apparent cause and was later restarted with no     4.3 of the data calculation quoted in this F&O has been corrective activity. This type of event may not be removed from the calculation due to an inaccurate counted as a MR functional failure..." However, Revision 0A                                                                                                                                   Page 43 of 79

54002-CALC-01                                                                      Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic        Status                   Finding/Observation                                         Disposition ID) this type of failure would be screened out by the  representation of the process used in evaluating failure identification process that starts with MR failures.
functional failures.
DA-C4 (1)          Plant component    Dispositioned For maintenance rule functional failures and       This is primarily a documentation issue and wording
[Cat I-III is NOT  failure data                     unavailability events, the reviewer was unable     has been added to Section 4.3 of the data calculation MET]                                                to find system or component specific definitions   (RNP-F/PSA-0072) to provide more detailed of these MR performance criteria. ADM-NGGC-        information on the identification of events as failures.
0101 states that PSA should be considered as       There are definitions in the RNP scoping documents an input into MR performance criteria              that are used in the evaluation of functional failures for development. No evidence was found that MR         the Maintenance Rule. The MR database is then used System functional failure and unavailability       to capture all failure events. These events are then definitions have been compared to definitions or   evaluated and screening comments provided to scoping in PSA. This SR is not met.                determine if they count as failures towards the PSA.
RESOLUTION OF CAPABILITY CATEGORY No clear basis was discovered for identification   CLASSIFICATION of events as failures.                             The documentation has been updated and addressed accordingly. Therefore, SR DA-C4 is considered MET at CAT I-III.
DA-C6 (1)          Plant-specific     Dispositioned In practice, PI (computer point) data is used for  Further documentation has been added to Section 4.4
[Cat I-III is MET] component demands                estimating demands on standby components,          of the data calculation (RNP-F/PSA-0072) to where the data points exist. According to          demonstrate that post maintenance testing is excluded Section 4.4 of RNP-F/PSA-0072, estimates of        from the collection of demand data. The component operation based on normal operating practices,     data based on OSI-PI utilizes a one hour screening surveillance procedures (SPs), and System          value. Other demands are based on test procedures or Engineer input were used for the remaining         system engineer input and only AOVs and MOVs, Batteries, Battery Chargers,        include those demands specified during the test or Air Compressors, Dedicated Shutdown Diesel         during normal operations.
Generator, Diesel Driven Fire Pump, Motor Driven Fire Pump, and Deepwell Pumps. There is no evidence that demands from post maintenance testing are excluded. No unusual Revision 0A                                                                                                                                Page 44 of 79

54002-CALC-01                                                                  Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation             Topic         Status                  Finding/Observation                                         Disposition ID) number of demands as multiple demands in an hour are excluded.
DA-C8 (1)         Component          Dispositioned The flag events that represent the time that      Data was obtained and analyzed to determine the
[Cat I is MET]    configurations in                components were configured in their standby       runtime percentages of all possible alignment standby                          status do not appear to use actual plant data as  configurations for select plant systems. These plant a primary source.                                 systems were selected based on those for which data was available and those which utilized alignment flags in previous model revisions. New basic events were put into the model to replace the former alignment flags. The new events represent each possible running configuration and have a value equal to the percentage that the alignment was found to be operating.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With the incorporation of plant data into this analysis, SR DA-C8 is considered to be MET at CAT II/III is met.
DA-D4 (1)         Bayesian approach  Dispositioned No discussion is provided to demonstrate the      Section 5.1 of the data calculation (RNP-F/PSA-0072)
[Cat I is MET]                                     posterior distributions are reasonable given the  provides identification of ways in which the Bayesian relative weight of evidence provided by the prior results have been validated using guidance from the and the plant-specific data.                      standard. Both generic and plant specific data was reviewed and spreadsheet (RNP_Bayes_Data_R1.xlsx) calculations analyzed for appropriateness in the performance of the Bayesian update. Results of the update were checked for appropriateness using at least one of the methods put forth in the standard.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With the incorporation of the posterior distributions to the Bayesian approach, CAT II/III is considered to be MET.
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54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic             Status                  Finding/Observation                                            Disposition ID)
DA-D6 (1)           Common Cause           Dispositioned No evidence that a review was performed of           Performed a review for plant experience with common
[Cat I is MET]      Failure probabilities                generic common cause failure probabilities to        cause failures, and updated RNP-F/PSA-0073 Section be consistent with available plant experience.       2.0 to include this discussion.
There is no discussion of available plant experience with respect to common cause              RESOLUTION OF CAPABILITY CATEGORY failures in RNP-F/PSA-0073.                          CLASSIFICATION Incorporation of plant experience qualifies for meeting CAT II requirements for DA-D6.
DA-E2 (1)           Data analysis process  Dispositioned The PRA documentation does not show the              Development and discussion of the occurrences
[Cat I-III is MET]  documentation                        probabilities or basis for the probabilities for the represented by the basic events cited in the F&O may following basic events (note that the basic          be found in the Westinghouse Owners Group events are assigned probabilities in the fault       document WOG Risk-Informed ATWS Assessment tree): #ACRDMF (control rod fails due to             and Licensing Implementation Process (WCAP-mechanical binding), #CRDMF (insufficient rod        15831-P) which is cited in the RNP Success Criteria insertion), #RPS (failure of reactor trip),          calculation (RNP-F/PSA-0075). Additionally, EAMSAC (AMSAC failure), ESFAS (ESFAS                 information regarding to the performance of a fails), CCVENT (performing CV purge), and            containment purge may be found in the Technical GINRDOORSL (personnel hatch door gasket).            Specifications for RNP, Sections 3.3.6, 3.6, and 3.9.3.
No evidence was found that these basic event values are not appropriate.
HR-H1           (1) Operator recovery      Dispositioned Section E.5 RNP-F/PSA-0014 Rev. 1 discusses          As indicated in Section E.5 of RNP-F/PSA-0014, the
[Cat II is MET]     actions                              TYPE Cr: POST-INITIATOR RECOVERY                     only non-proceduralized (Type CR) human interactions ACTIONS; states that the only recovery human         included in the Robinson PSA model are in the Plant actions included in the Robinson PSA model           flooding analysis (RNP-F/PSA-0009). OPER-48 and are in the Plant flooding analysis, but Table A-1    OPER-51 are not Type CR.
indicates OPER-48 recovery of the loss of IA to the SG PORVs via the use of the Steam Dump N2 accumulator and OPER-51 Failure to recover RCP seal injection with alternate filter line. Documentation of operator recovery actions is not clear and consistent.
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54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation                Topic            Status                   Finding/Observation                                         Disposition ID)
HR-I1 (1)          HRA documentation       Dispositioned The Type A pre-initiator process is unique and     Updated Figure 1-1, Pre-Initiator Human Failure Events
[Cat I-III is MET]                                       therefore requires some detailed review to         Flow Chart of RNP-F/PSA-0071 Rev. 2 to provide identify how HFEs were                             additional detail about how HFEs are identified.
identified for inclusion in the models.
IE-A6 (1)          Initiating events from  Dispositioned Several IE's are evaluated with a fault tree that  Section 1.1.4.10 and 1.1.4.11 of RNP-F/PSA-0006,
[Cat II is MET]    multiple failures                     models the loss of a system. Such fault trees      Rev 3 has been updated to include consideration.
typically include common cause failures and various system alignments caused by test and maintenance unavailability. No consideration is documented for potential unique initiating events for common cause failures of both DC busses or both AC busses. The consequences of such initiating events would be more severe than an initiating event where a single bus fails.
IE-C3 (1)          Recovery action credit  Dispositioned The screening of the PORV opening due to           Should pressure transmitter PT-445 fail high, a
[Cat I-III is MET]                                       failure of PT-445 needs to be investigated as a    demand for the PCV-456 will be generated. In order for potential unique initiator.                        the valve to open, the P-11 interlock must be defeated (i.e., RCS pressure > 2000 psi). Should this occur, the opening of the valve will result in a decrease in RCS pressure and a reactor trip on low pressurizer pressure. This event is similar to a small LOCA with a frequency of 5E-4 per year. However, it can be mitigated if the operator closes either the PORV or its associated block valve. When mitigated, it would be similar to the reactor trip initiator of 2.53E-01 per year.
When the potential for operator recovery is included, the initiating frequency for this event is much less than the small LOCA initiating event frequency: From RNP-F/PSA-0072, RNP Failure Data Analysis, the probability of PORV Spurious Operation is 3.31E-07 per hour, PORV Fails to Close is 6.70E-4 per demand, PORV Block Valve Fails to Close is 1.64E-3 per demand and from RNP-F/PSA-0014 the potential for operator recovery is 2.6E-3. The frequency of a Revision 0A                                                                                                                                       Page 47 of 79

54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic          Status                  Finding/Observation                                            Disposition ID) spurious PORV opening and failure to reclose is thus 1.5E-5 per year (Note: 2 PORVs) which is small compared to the small LOCA frequency of 5E-4 per year. In addition, it does not result in any other significant impacts on safety systems. Therefore, a spurious PORV opening is not considered to require separate consideration in the RNP PRA model. RNP-F/PSA-0006, Section 1.1.4.6 is updated to include this justification.
IE-D3 (1)         Initiating event      Dispositioned A qualitative assessment of identified IE            Rationale for using lognormal distribution is
[Cat I-III is NOT analysis                            uncertainties is presented in RNP-F/PSA-0074.        documented in RNP-F/PSA-0049, Section 6.5 and MET]                                                  No basis provided for selection of uncertainty       Table 12 note 1 in RNP-F/PSA-0006 is updated distributions.                                       accordingly.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated to provide the appropriate basis for this methodology and addressed accordingly. Therefore, SR IE-D3 is considered to be MET at CAT I-III.
LE-C11 (1)        Containment failure   Dispositioned RNP-F/PSA-0047 credits containment spray             Level 2 operator actions were assessed in more detail
[Cat I is MET]    impacts on equipment                operation following containment failure for small    using industry-standard methodologies via the EPRI and operator actions                and late releases, but not for large early           HRA calculator. As part of that assessment,
[Associated with                                      releases. Manual isolation operator actions are      adjustments were made as appropriate to the LE-C12]                                               described in Sections 8.3.6 and 8.3.7. No            environmental conditions under the actions that would evidence of justification of credit for equipment    be performed. In this context, assessment of survivability or justification of credit for human   Containment Spray success is used to determine actions that could potentially be impacted by        whether a containment release (non-ISLOCA or SGTR) containment failure.                                 is scrubbed or unscrubbed. In many end-states, in the Containment Event Tree, this determines whether the release is a Large Early release or not. The success or failure of the Containment Spray system to perform its credited functions is determined through the Revision 0A                                                                                                                                    Page 48 of 79

54002-CALC-01                                                                     Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation             Topic          Status                  Finding/Observation                                          Disposition ID) development of failure and unavailability data using industry-approved sources and/or plant-specific data.
Furthermore, the Containment Spray system is an ECCS mitigation system and thus designed to sustain the radiation and temperature effects of post-accident conditions postulated in the design basis. No environmental conditions were identified which required the Containment Spray system to operate beyond their design basis. Thus, it is reasonable to conclude that the Containment Spray system would be available (within the constraints of the failure and unavailability data) during a post-accident containment failure.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION With adequate justification for credit of equipment survivability and human actions impacted by containment failure, this SR is considered to be MET at CAT II/III.
LE-C2 (1)         Operator actions    Dispositioned LERF-related operator actions following the         Level 2 operator actions were assessed in more detail
[Cat I is MET]    following onset of                onset of core damage are described in the HRA       using industry-standard methodologies via the EPRI core damage                       notebook (RNP-F/PSA-0014, Annex B) and              HRA calculator.
RNP-F/PSA-0047. Screening values were set to 0.1 to see if there was sensitivity of results. For RESOLUTION OF CAPABILITY CATEGORY example, operator action to mitigate ISLOCA is      CLASSIFICATION evaluated in Section 8.2 of RNP-F/PSA-0047.         Through analysis of human operator actions with HRA The probability of failure is assumed to be 0.1     calculator as a more thorough approach, CAT II/III is based on "available guidance and time               considered to be MET.
available." The scenarios for these operator actions were not adequately documented as consistent with applicable procedures. Need to ensure applicability to the current as built/as operated plant.
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54002-CALC-01                                                                        Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic           Status                   Finding/Observation                                          Disposition ID)
LE-C4 (1)         LERF model logic for  Dispositioned The accident sequences are based on generic         Assessment performed in RNP-F/PSA-0062.
[Cat I is MET]    accident progression                references and on plant specific MAAP analysis. Important mitigation actions in           Based on this assessment, and based on the fact that significant accident progression sequences,         SI-864A and SI-864B are in series in the RWST supply such as SG isolation for SGTR are not               line and are in close proximity to one another, it is modeled. Evidence of technical justification for    reasonable to conclude that the ISLOCA pathway to demonstrating the feasibility of mitigating         the RWST could, at worst case, be isolated by closing actions was not provided. Scrubbing is              one valve until it can be closed no further and then discussed for release categories in                 closing the other until it can be closed no further, and RNP-F/PSA-0048 and brief rationale is provided      so on going back and forth between the valves until the when it is credited. Inclusion of beneficial        line is isolated.
failures was not observed.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The assessment and justification provided above are considered to meet CAT II for SR LE-C4.
LE-C9 (1)         Human actions under   Dispositioned Four LERF HFEs documented in Sections 8.2,          Level 2 operator actions were assessed in more detail
[Cat II-III is    adverse environments                8.3.3, 8.3.6, 8.3.7, and 8.3.11 of                  using industry-standard methodologies via the EPRI MET]                                                  RNP-F/PSA-0047 are modeled. Evidence of             HRA calculator. As part of that assessment, justification for human actions under adverse       adjustments were made as appropriate to the environments was not found in the LERF HFE          environmental conditions under the actions would be documentation. One of the HFEs, OPER-ISOL,          performed.
represents local closure of manual valves given failure of containment isolation. The environmental conditions associated with these valves could potentially have an impact on the operator's ability to locally close the valves and needs to be addressed. Although this action was not credited, the documentation should be clear about this.
LE-D5 (1)         Secondary side        Dispositioned Based on a review of a number of calculation        Assumption of 10 cycles for SG PORVs is not used in
[Cat I is NOT     isolation capability                files, SGTR event tree, and fault tree for #RW, it  the Level 2 analysis. For induced SGTR analysis that MET]              analysis                            appears that a realistic secondary side isolation   was added to the Level 2 analysis, an assumption of capability analysis is performed. However, there    40 cycles was used. Since one stuck-open SG SRV is Revision 0A                                                                                                                                  Page 50 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation             Topic             Status                    Finding/Observation                                          Disposition ID) are a number of important issues that their           considered sufficient for a large early release, the resolutions may impact this conclusion. These         modeling of multiple SRVs sticking open is extraneous.
issues include a) A justification for the assumed number of cycles for the PORVs on the faulted         RESOLUTION OF CAPABILITY CATEGORY SG (which is 10) is not provided. Also, it seems      CLASSIFICATION that only 'fail to close' of one SRV is considered    Based on the above disposition LE-D5 is considered to in the fault tree. b) Additionally, it appears that   be MET at CAT II.
Operator action to isolate the faulted SG is not included as a potential contributor to the secondary side isolation failure probability.
Given that in the current model 1) fail to isolate probability is dominated by the probability of a PORV sticking open, and 2) the SGTR IE is one of the highest contributors to the LERF figure of merit, it is important to ensure that the above apparent issues are fully resolved.
LE-D6 (1)          Thermally-Induced     Dispositioned Based on a review of Sections 8.2.4 and 8.3.7         More detailed modeling of Induced SGTR was added
[Cat I is MET]     Steam Generator                     of RNP-F/PSA-0047 a simple representation of          to the level 2 analysis as described in section 8.2.2 of (SG) Tube Rupture                   thermally induced SGTR is performed on a              RNP-F/PSA-0047.
Analysis                            generic basis. Note that 1.200 Gap Analysis comment with respect to ISGTR is a valid              RESOLUTION OF CAPABILITY CATEGORY comment. The response to the comment seems            CLASSIFICATION to be adequate but could not be verified.             Based on the above disposition LE-D6 is considered to be MET at CAT II.
LE-E1 (1)          Parameter values for  Dispositioned The system level data are documented in the           Description of development of input parameters is
[Cat I-III is MET] equipment and                       Level 1 analysis and some Level 2-specific            provided in RNP-F/PSA-0047. In many cases, they are operator response                   parameters are documented in Attachment 2 of          based on expert judgment.
RNP-F/PSA-0047. The basis for some of the values was subjective (e.g., see Table 8-10) and others are based primarily on the IPE submittal and supporting MAAP runs for the Level 2 analysis (e.g., Table 8- 35). Although it is understood that the approach for estimating probabilities for some Level 2 parameters is Revision 0A                                                                                                                                     Page 51 of 79

54002-CALC-01                                                                        Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic            Status                   Finding/Observation                                        Disposition ID) soft, using IPE results without additional justification may not be as realistic as required for a CC II classification.
LE-F1 (1)          Parameter estimates   Dispositioned Relative contribution to LERF is presented in      Induced SGTR has been added to the Level 2 model
[Cat II-III is     for accident                        the quantification notebook (RNP-F/PSA-0077)       for RNP. It is the dominant contributor for large bypass MET]               progression                         and the source term development notebook           containment failure sequences. Dominant contributors phenomena                           (RNP-F/PSA-0048). Induced SGTR contribution        to LERF and containment failure are described in was not presented among the significant LERF       sections 3.16 and 3.17 of the quantification calculation contributor results presented in Section 3.16 of   (RNP-F/PSA-0077).
RNP-F/PSA-0077.
LE-F2 (1)          Documentation of      Dispositioned The updated quantification notebook                Review of LERF contributors and results of that review
[Cat I-III is NOT  Contributors to LERF                (RNP-F/PSA-0077) was examined to find              are described in section 3.3 of the quantification MET]               analysis                            evidence of a review of the LERF contributors      calculation.
for reasonableness. All that was found were entries for "Level 2 HNPSUM Changes" toward        RESOLUTION OF CAPABILITY CATEGORY the end of the Appendix A "Change Log for          CLASSIFICATION RNP MOR09." Clear documentation is needed          Based on the above disposition LE-F2 is considered to to know what LERF reasonableness reviews           be MET at CAT I-III.
were performed and what were the results of the reviews to ensure that the SR is fully met.
LE-F3 (1)          Documentation of      Dispositioned The uncertainty analysis, as documented in         Assumptions and Uncertainties related to LERF are
[Cat I-III is MET] LERF analysis                       RNP-F/PSA-0074, contains very little               revised and documented in RNP-F/PSA-0074.
documentation of identification of LERF
[Associated with                                       uncertainties. Only about 10% of all the LE-G4]                                                 assumptions and uncertainties documented in Table 1 of RNP-F/PSA-0074, seem to be LERF related, even though LERF analysis as whole is supported by many conservative analysis assumptions. Additionally most of the identified uncertainties are not related to the biggest contributor to the LERF figure or merit (i.e., a SGTR event). This reviewer found only one (Item 195).
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54002-CALC-01                                                                        Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation                Topic          Status                  Finding/Observation                                        Disposition ID)
LE-G5           (1) LERF analysis        Dispositioned Limitations in the LERF analysis that would       RNP-F/PSA-0047, Rev. 2, Section 8.4.2 states that No
[Cat I-III is NOT   limitations                        impact applications are not identified.           limitations were identified in Level 2 analysis (including MET]                                                                                                     LERF) that would impact applications. However, any assumptions and uncertainties identified in the analysis are addressed in RNP-F/PSA-0074, RNP Uncertainty Analysis.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR LE-G5 is considered to be MET at CAT I-III.
QU-B3 (1)           Truncation analysis  Dispositioned The current truncation levels of 1E-10 for CDF    An appropriate lower truncation is used in
[Cat I-III is NOT                                      and LERF are insufficient to estimate the mean    RNP-F/PSA-0077, Rev. 2, and Section 2.2 truncation MET]                                                   CDF or LERF required by QU-A3 and LE-E4.          study shows convergence for both CDF and LERF.
The truncation study in Section 2.2 of RNP-F/PSA-0077 shows that truncation occurs for       RESOLUTION OF CAPABILITY CATEGORY CDF, but does not establish that it occurs at 1E- CLASSIFICATION
: 10. The study does not show convergence for       Based on the above disposition, SR QU-B3 is LERF. SR QU-B3 describes an acceptable            considered to be MET at CAT I-III.
process for demonstrating convergence, where a successive decade reduction in truncation results in a CDF or LERF change of less than 5%. The RNP results show a decade reduction from 1E-10 to 1E-11 results in a change of approximately 10% for CDF and 30% for LERF, which does not meet the example process in the SR.
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54002-CALC-01                                                                    Evaluation of Risk Significance of Permanent ILRT Extension QU-D1 (1)         Review significant Dispositioned Documented cutset reviews were inadequate.        The following has been added to section 3.3 of the
[Cat I-III is NOT cutsets to verify                Section 3.3 of RNP-F/PSA-0077 explains that       quantification calculation (RNP-F/PSA-0077): Due to MET]              model logic                      the top 50 cutsets were reviewed for CDF and      the low truncation settings and large quantity of cutsets LERF. Section 1.1 defines a significant cutsets   generated it is not practical to evaluate each cutset in as one of the top 95% of cutsets or a cutset that the 95th percentile. (Approximately 3800 cutsets contributes more than 1% to CDF. The top 50       comprise the top 95 %.) All of the top fifty cutsets were cutsets for CDF are near 50% for all CDF.         reviewed followed by a review of select cutsets at LERF cutsets are assumed to have a similar        varying intervals leading up to the 95% level. The top limitation. Review of non-significant cutsets     fifteen cutsets represent 50% of the CDF, and cutset appeared to be adequate. RNP-F/PSA-0077           17 is the point at which each subsequent cutsets states that 'A sample of non-significant cutsets  contribute less than 1% to CDF. Cutsets at the bottom listed in Table B-2 (lower than the 95% of the    of the top 50 each only contribute approximately 0.2%
CDF) was reviewed for reasonableness and to       to total CDF. In addition to the top 50, the bottom 50 ensure recoveries were being applied.'            cutsets were evaluated to increase confidence that the generated MOR12 was yielding logical results. A list of the 50 top core damage cutsets is provided in Table B-1 in Appendix B. These cutsets are presented in order of descending frequency. A sample of non-significant cutsets listed in Table B-2(lower than the 95% of the CDF) was reviewed for reasonableness and to ensure recoveries were being applied. Part of this review for reasonableness was reviewing which accident sequences were contributing to CDF and LERF, comparing those results to the previous model with consideration of what model changes were made, and performing sensitivity testing as needed to determine the reason for the differences between the previous model and this one.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-D1 is considered to be MET at CAT I-III.
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54002-CALC-01                                                                       Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation              Topic            Status                   Finding/Observation                                        Disposition ID)
QU-D3 (1)         Review cutsets for     Dispositioned Section 3.3 of RNP-F/PSA-0077, Cutset Results      Section 3.3 of RNP-F/PSA-0077, states that flag
[Cat I-III is NOT flag events, mutually                Review (Significant & Non-significant) indicates   events settings, mutually exclusive event rules and MET]              exclusive events, and                that "a panel of PRA experts reviewed the          recovery rules are reviewed for correctness and logical recovery rule                        cutsets for modeling accuracy, operational         results.
accuracy                             consistency and to ensure the correct recoveries were being applied." But only found     RESOLUTION OF CAPABILITY CATEGORY limited evidence that results were reviewed to     CLASSIFICATION determine that flag event settings and mutually    Based on the disposition above, SR QU-D3 is exclusive event rules yield logical results. Found considered to be MET at CAT I-III.
model changes listed in Table A-1 of RNP-F/PSA-0077 that indicated that results had been reviewed.
QU-D4 (1)         Compare results to     Dispositioned RNP-F/PSA-0077 Section 3.4, Similar Plant          Section 3.4 of RNP-F/PSA-0077, is updated to provide
[Cat I is MET]    similar plants                       Review, Table 7 shows a comparison of              adequate comparison. Comparison with the LERF was Robinson CDF and LERF values to Turkey             conducted in the cut-set review process. However, this Point and Beaver Valley. Plant systems             detailed review was not documented in the current differences are compared. However, the             MOR.
comparison is inadequate and sources of specific differences (e.g., LOCA contribution)     RESOLUTION OF CAPABILITY CATEGORY are not identified.                                CLASSIFICATION Based on the above disposition, SR QU-D4 is considered to be MET at CAT II-III.
QU-D7 (1)         Review                 Dispositioned Importance was calculated at basic event level,    Section 3.13 of RNP-F/PSA-0077, documents
[Cat I-III is NOT component/basic                      but could not be found at component level.         Component Importance.
MET]              event importance                     Review did not find documentation to substantiate review of component importance.       RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-D7 is considered to be MET at CAT I-III.
QU-F2 (1)         Model integration      Dispositioned Several items identified as typically documented   1. System successes are listed in the system notebook
[Cat I-III is NOT process                              for SR QU-F2 were not adequately                   calculation (RNP-F/PSA-0018) and described in the MET]                                                   documented. Missing or inadequate                  success criteria notebook (RNP-F/PSA-0075).
documentation was identified for the following     2. A discussion of the sensitivity of accident sequences Revision 0A                                                                                                                                   Page 55 of 79

54002-CALC-01                                                                      Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation             Topic            Status                  Finding/Observation                                         Disposition ID) items: 1) description of the process used to      to operator actions has been added to section 2.7 of account for system successes, 2) discussion of    the quantification calculation (RNP-F/PSA-0077).
human actions that are the key factors in         3. Asymmetries for RNP are discussed in the system causing the accidents to be non-dominant, 3)      notebook calculation (RNP-F/PSA-0018). Asymmetries discussion of asymmetries in the model, 4) list   exist in particular on the AC Power system (and of mutually exclusive results, 5) cutset review   associated systems like DGs and DC power).
process, and 6) discussion of RRW for             4. Mutually-exclusive events are listed in the equipment or operator errors.                     RNP_MTX_12.xls file, and there is a cutset file which lists all the mutually-exclusive combinations in the fault tree. Both of these files are part of the quantification calculation (RNP-F/PSA-0077).
: 5. Cutset review process is described in section 3.3
: 6. No specific standard requirement to discussion review of RRW, but the review of F-V importance is discussed in the quantification calculation and this is considered adequate for RRW as well.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the above disposition, SR QU-F2 is considered to be MET at CAT I-III.
QU-F3 (1)         Document significant  Dispositioned Important basic events are documented in          Discussion of top accident sequences for SGTR,
[Cat II-III is    contributors to CDF                 Table 10 of RNP-F/PSA-0077. Important IE          ATWS and LOCA have been added to section 3.15 of MET]              results                             importances are documented Figure 18.             the quantification calculation (RNP-F/PSA-0077).
Detailed accident sequence descriptions are provided in Section 3.15. Some, but not all, of these accident sequences describe the top sequences. For example, no top sequence was described for SGTR, ATWS, or LOCA. Top event tree sequences are discussed in general, but the top cutsets within the event tree sequences are not described. Detailed description of significant accident sequences or functional failure groups could be improved to Revision 0A                                                                                                                                 Page 56 of 79

54002-CALC-01                                                                             Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic               Status                  Finding/Observation                                           Disposition ID) provide more useful detail for all significant contributors to CDF.
QU-F5 (1)          Document limitations      Dispositioned Some software and hardware limitations are         Section 1.3 of RNP-F/PSA-0077, describes limitations.
[Cat I-III is MET] of the quantification                   cited in RNP-F/PSA-0077 Section 2.2, but are       The limitations of the hardware and software have process                                 not described.                                     been discussed in section 2.2. One example of limitations is documented as being the truncation level and a discussion is included as to how better confidence is obtained in the truncation level.
QU-F6 (1)          Document                  Dispositioned Significant' definitions are in Section 1.1 of     Justification for not performing a review of all cutsets in
[Cat I-III is MET] quantitative definitions                RNP-F/PSA-0077. However, the definitions           the top 95% was provided in section 3.3 of the for significant basic                   were not consistently applied. The cutset results  quantification calculation (RNP-F/PSA-0077).
events, cutsets and                     review documented in Section 3.3 of RNP-accident sequences                      F/PSA-0077 was not performed for the top 95%,
as discussed in SR QU-D1. No justification provided for not performing a review of significant cutsets. Also, as discussed in QU-D1, it was not clear that RNP-F/PSA-0077 Table B-4 cutsets included those lower than 95% of the LERF.
SC-A1 (1)          Definition of core        Dispositioned The success criteria definitions are documented    Success criteria definitions documented in calculations
[Cat I-III is NOT  damage                                  in RNP-F/PSA-0075, RNP-F/PSA-0061,                 RNP-F/PSA-0075, RNP-F/PSA-0043, MET]                                                       RNP-F/PSA-0043 and RNP-F/PSA-0049. There           RNP-F/PSA-0061 and RNP-F/PSA-0049 have been is some inconsistency in the success criteria      revised to provide consistent statements throughout.
[Associated with                                           definitions used for MAAP analyses.                Definition for "briefly exceeding 1800 &deg;F" has been SC-A2]                                                                                                        revised to read if the temperature exceeds 1800 &deg;F for a short time (i.e., ~1 min) in RNP-F/PSA-0043 and RNP-F/PSA-0049 based on the MAAP case MLOCA_8-HPI.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION Based on the disposition above, SR SC-A1 is considered to be MET at CAT I-III.
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54002-CALC-01                                                                          Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation                Topic            Status                  Finding/Observation                                         Disposition ID)
SC-B3           (1) Definition of Success  Dispositioned RNP Success Criteria calculation                  The success criteria selected for the PSA for the
[Cat I-III is MET]  Criteria                             RNP-F/PSA-0075 and Accident Sequence              LBLOCAs is based on the success criteria specified for calculation RNP-F/PSA-0043 references the         design basis LOCAs in the UFSAR. This includes the
[Associated with                                         supporting thermal hydraulic analyses and/or      use of a single train of equipment for injection and SC-C2]                                                   other applicable analyses. MAAP was used to       recirculation, and injection from the 2 accumulators on confirm the success criteria for large break      the non-broken loops. To provide a sense of assurance LOCA in the PSA, which is based on the            that the UFSAR success criterion was applicable to the success criteria specified for design basis       PSA, the MAAP code was looked at but not credited to LBLOCAs in the UFSAR, according to                evaluate LBLOCA scenarios at RNP. Last paragraph in RNP-F/PSA-0075, Section D.3.1. Although the       Section D.3.1. of RNP-F/PSA-0075 is re-written to LBLOCA MAAP run was described as only a           clarify this.
confirmation of the assumed success criteria, MAAP is not appropriate for analysis of large break LOCAs. MAAP does not conserve momentum and can provide misleading unrealistic results for LBLOCA.
SC-C3 (1)           Document model         Dispositioned Sources of model uncertainty and related          The success criteria calc, RNP-F/PSA-0075, has been
[Cat I-III is NOT   uncertainty                          assumptions are inadequately identified and       reviewed to identify source of model uncertainties and MET]                                                     associated with the development of success        related assumptions. The identified assumptions and criteria.                                         uncertainties have been added in Table 1 of calc RNP-F/PSA-0074, RNP Uncertainty Analysis.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated and addressed accordingly. Therefore, SR SC-C3 is considered to be MET at CAT I-III.
SY-A11          (1) Failures in system     Dispositioned Based on a review of a number of system fault     A review of the systems revealed that an RWST
[Cat I-III is MET]  model affecting                      trees and system notebooks, it appears that       rupture event was not included in the model. Event operability                          most of the pertinent equipment failures are      HTKRWSTFN was developed and added to the model.
included in the system models. However, pipe      An evaluation of pipe rupture events was performed Revision 0A                                                                                                                                    Page 58 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation               Topic            Status                   Finding/Observation                                          Disposition ID) ruptures are not included as potential              using the criteria in SR SY-A15, and it was determined equipment failures based on low probability of      that there is adequate justification to discount pipe pipe rupture. This exclusion is acceptable per      failure events for those systems where pipe failures SY-A15 but two additional points need to be         were previously assumed to have a negligible impact covered: 1) pipe failure probabilities for events   on risk. A statement to that effect has been added to that result is the total loss of a system need to   the applicable system notebooks.
be compared with the random probability of system failure not only one component failure probability, 2) if the same fault tree is used for quantifying the IE frequency, the low failure probability/frequency may not be true any longer.
SY-A11 (2)         Failures in system     Dispositioned RWST tank rupture not accounted for in the          Added basic event HTKRWSTFN under gates J150
[Cat I-III is MET] model affecting                      fault tree.                                         and H044 for RWST rupture.
operability SY-A2 (1)          Accuracy of system     Dispositioned Based on a review of RNP-F/PSA-0018                 SR SY-A2 requires the following: COLLECT pertinent
[Cat I-III is MET] analysis compared to                 attachments, (e.g., Attachment A.1, A.9, and        information to ensure that the systems analysis as-built, as-operated                A.5), it is concluded that pertinent data appears   appropriately reflects the as-built and as-operated
[Associated with   plant                                to have been used. Additionally, each system        systems. Examples of such information include system SY-A3]                                                  notebook appears to include a list of all the       P&IDs, one-line diagrams, instrumentation and control pertinent references. However, a direct             drawings, spatial layout drawings, system operating relationship between the information in the         procedures, abnormal operating procedures, system notebook and the source of information       emergency procedures, success criteria calculations, is often not provided (e.g., The source             the final or updated SAR, technical specifications, document for Figure A.1.1 is not provided. Or       training information, system descriptions and related source of information for instrumentation and       design documents, actual system operating control is not stated).                             experience, and interviews with system engineers and operators.
While it could be considered a good practice, there is no standard requirement to provide direct references within the body of the system notebook documentation to support the statements made therein. Additionally, there is no standard requirement to differentiate Revision 0A                                                                                                                                      Page 59 of 79

54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 1 Internal Events PRA Peer Review - Facts and Observations SR (Observation                Topic           Status                   Finding/Observation                                          Disposition ID) between informational and direct references. All references are included in section 8.0 of each system notebook appendix.
It is acknowledged that traceability of system engineer review comments was deficient. In subsequent system notebook updates, system engineers were assigned NTMs in Passport in which to record their review comments.
As such, no changes are required to meet this SR.
SY-A23 (1)          Develop consistent    Dispositioned The RNP utilizes CAFTA type codes such that a      Added Section 2.8 Basic Events Naming Convention to
[Cat I-III is MET]  model nomenclature                  component failure mode and its associated          RNP-F/PSA-0049.
failure rate are applied consistently in multiple systems. Need documentation of the naming convention for basic event names.
SY-A8           (1) Establish component   Dispositioned The boundaries of the modeled components           The definition of the EDG component boundary is
[Cat I-III is MET]  boundaries,                         appear to match the definitions used to            correct in that it includes the output breaker. The event definitions and data                establish the component failure data. However,     representing the failure of the output breakers there is some evidence that the data boundary      (PCB5217BNN and PCB5227BNN) to close will be set definitions are not consistent with the definition to zero in the model since failures of this component of component boundary. For example, the            are rolled into failures of the EDG. A review was made definition of the component boundary for EDGs      of other component boundaries and it was determined does not include output breaker but the            the boundaries detailed in Attachment 2 of the data definition of the component boundary in the        calculation (RNP-F/PSA-0072) and in the model are data analysis includes output breaker.             appropriate when compared to those set forth in NUREG-CR/6928.
SY-B8           (1) Documentation of      Dispositioned The spatial hazards that may impact multiple       Additional discussion added to Section 3.2 of the
[Cat I-III is NOT   Spatial Hazards                     systems or redundant components in the same        System Notebook Appendices, as appropriate.
MET]                                                    system are not clearly identified.                 RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION The documentation has been updated and addressed accordingly. Therefore, SR SY-B8 is considered to be MET at CAT I-III.
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54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                           Topic                Status                            Finding                                          Disposition CF-A1-01          REVIEW the conditional failure     Dispositioned RNP-0153 is Attachment 9 of RNP-F/PRA-0094         RNP-0153 (Attachment 7 of P2217-1021-01-03)
(CAT III)         probabilities for fire-induced                   and describes cable failure analyzed for the RNP   has been updated to reflect all RNP information.
circuit failures                                 Fire PRA, RNP-0153. The third bullet of RNP-       Additionally, Reference 5.1 has been updated to
[Associated       and                                              0153, Section 3.0 states, "All cables are assumed  RNP Fire Safe Shutdown Program Database, with CF-A1]       ASSIGN the appropriate industry-                 to be Thermoset. This is consistent with the cable Rev 19.
wide generic values for risk-                    specifications for safety related cable at BNP and significant contributors based on                there are no substantive differences in failure the specific circuit configuration               probabilities for the two cable types for the under consideration                              purposes of this analysis." Reference 5.1 of RNP-0153 is "BNP Fire Safe Shutdown Program Database, Rev. 26." In response to a Peer Review question, it was advised that BNP should be RNP, that cable should be Thermo-plastic rather than Thermo-set and the RNP FSSPMD should be referenced instead.
The determination of meeting the requirements of CF-A1 depends on the information in RNP-0153.
As presently written, RNP-0153 leads to questions of whether it is really applicable to RNP.
Revise and update RNP-0153. Perform a confirmation that the balance of information in RNP-0153 is valid for H.B. Robinson.
CF-A2-01          CHARACTERIZE the uncertainty       Dispositioned No characterization of the uncertainty associated  The basic events associated with hot short associated with the applied                      with the applied conditional circuit failure       probabilities have been assigned an error factor (NOT MET)         conditional failure probability                  probabilities was documented in EPM Report         when the combined cutset is created using the P2217-1021-01-01, Robinson Fire PRA                UNCERT code. Although important for assigned per CF-A1.
Quantification Calculation, as captured in EC      determining the statistical uncertainty of the PRA 90905.                                             cutsets, the criteria needed for the LAR are based on the mean values which are not significantly impacted by the uncertainty analysis.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
Based on the above disposition, SR CF-A2 is considered to be MET at CAT I/II/III.
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SR                             Topic                  Status                           Finding                                           Disposition CS-A1-01             IDENTIFY cables whose fire-        Dispositioned Attachment 7 of RNP-F/PSA-0066 (Equipment           All cables identified in Attachment 7 of induced failure could adversely                  Selection) contains a list of equipment which is to RNP-F/PSA-0066 have been routed and added (CAT I/II/III)       affect selected equipment and/or                 be credited in the Fire PRA, but which does not     to FSSPMD with the exception of components credited functions in the Fire PRA               have cables identified and routed. A spot check     that were not installed prior to RNP-F/PSA-0066
[Associated          plant response model.                            review of the cable database (RNP FSSPMD            being completed.
SR: CS-A10,                                                           R21_Read Only QA Record.mdb) reveals that CS-A2]                                                                several of the components listed in Attachment 7 are not included in the cable database. For example, Attachment 7 includes "Fire Tag" CHG-C-INDICATING-LIGHT corresponding to PRA BE JILCHGPCTF; however, CHG-C-INDICATINGLIGHT is not in the cable database, hence BE JILCHGPCTF will never be affected by fires. Others identified by spot check include 480V-52/11A, 480V-52/13B, EDGA-AMMETER, etc. Thus, there are some PRA components which are being credited in the PRA which do not have cable routing incorporated into the cable database; hence, they will never fail.
Review Attachment 7 of RNP-F/PSA-0066, identify components which are not in the cable database, and update the cable database to include all credited PRA components.
CS-A11-01            If assumed cable routing used in   Dispositioned The scope and extent of assumed cable routing       As noted in RNP Change Package RNP-0152 the Fire PRA, IDENTIFY the                       for non-credited components (i.e., components       (Attachment 19 of P2217-1021-01-03), cable (CAT I/II/III)       scope and extent, and PROVIDE                    assumed to fail for all fires) is documented in     toning was used to confirm cable routes. There a basis for the assumed cable                    Table 4 of RNP-F/PSA-0066. Although a basis is      were some instances where cable toning was not routing.                                         not provided for the assumed routing, it is a       possible within a specific compartment generally accepted practice to omit certain         (embedded cable, etc.). In these instances, the systems which do not perform a safety function      cable was assumed to be failed throughout the (but may back up a safety system) and would         entire compartment that it was known to traverse require immense amount of work to manually          through.
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54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                           Topic                 Status                            Finding                                            Disposition
[Associated                                                        trace the cables. The credited components with SR: CS-C3]                                                         assumed routing are contained in the FSSPMD,          As noted in RNP change package RNP-0205, however, there is no documented basis for the         (Attachment 19 of P2217-1021-01-03), the routing, and it is unclear if the scope and extent is assumed cable route data determines the cable-understood. The basis for assumed routing is not      to-fire zone correlation (which is sufficient for documented. Furthermore, to understand, and to        NSCA), but does not determine the cable-to-be able to evaluate the uncertainty associated        raceway-to-fire zone correlation (which is needed with assumed routing some form of                     for PRA). Based on this assumption, any ignition documentation should be assembled describing          source within a given fire zone will impact all the scope and extent of the assumed routing. For      cables with assumed cable routes in the ignition example, are there fire compartments with a           sources fire zone.
significant amount of assumed routing, are there high significance fire compartments, systems, trains or components dominated by cable failure of cables which have an assumed routing, etc.
As part of an overall Task 3 documentation package, describe the scope and extent of assumed routing used in the RNP Fire PRA.
CS-C1-01          DOCUMENT the cable selection       Dispositioned There is no notebook encompassing Task 3              In the process at RNP, Fire Protection/NSCA and location methodology applied                 (Cable Selection) making review and update            develops and maintains the cable selection and (NOT MET)         in the Fire PRA in a manner that                 difficult. There are numerous change packages,        circuit analysis data. This data is then referenced facilitates Fire PRA applications,               and a database (FSSPMD) which is a repository         as inputs to the Component Selection and upgrades, and peer review.                       for the cable routing information; however, there     Quantification FPRA calculations. This process is no document explaining what tasks were             and associated results are easily reviewable, has performed, which procedures or guidelines were        been peer reviewed multiple times for our other employed, and in which document the analysis is       sites and found to be acceptable. There is no contained.                                            requirement to have a separate PRA notebook.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
With no change being made, RNP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be Revision 0A                                                                                                                                       Page 63 of 79

54002-CALC-01                                                                            Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                           Topic                 Status                            Finding                                           Disposition sufficiently resolved for SR CS-C1 to be assessed as CAT I/II/III is MET.
ES-A2-01          REVIEW power supply, interlock     Dispositioned Power supplies, support systems, interlock           A detailed review of the modeling of associated circuits, instrumentation, and                   circuits and instrumentation are included in the     circuits for all components functional states in the (NOT MET)         support system dependencies                      internal events PRA model. For fire PRA specific     fire PRA CAFTA model was performed. The details of the review are provided in Attachment 2 and IDENTIFY additional                          logic, examples indicate that not all support
[Associated                                                                                                             of Hughes Associates Inc. (HAI) equipment whose fire-induced                     system dependencies were completely                  1RCS04042.414.031-002, and revisions to the SR: ES-B4,        failure, including spurious                      considered. The area of concern is the lack of       fire PRA CAFTA model, the RR file table BE, and CS-A2]            actuation, could adversely affect                modeling of instrumentation power dependency         Component Selection Database (CSDB) table any of the equipment identified                  (self-identified by the utility). For example, see   PRA_SSEL as a result of this detailed review are per SR ES-A1.                                    gate HRAPORV-455C "PORV FAILS TO CLOSE               listed in 1RCS04042.414.031-001, Attachments DUE TO HRA", one of its inputs is an AND gate        3, 4, and 5, respectively. Revision 4 of the RNP fire PRA calculation for component selection for "INDICATIONS THAT PORV IS OPEN FAIL."
(RNP-F/PSA-0066) incorporates all of the For the indicators modeled here, if there is a loss  changes required to the fire PRA CAFTA model of power, the indicators will scale low, potentially as a result of the detailed review of the preventing the operators from taking necessary       associated circuits.
actions. For the indication discussed in the example above, a loss of power will cause the        Review credited components to ensure that indicators to scale low, potentially preventing the  power supplies, support systems, interlock circuits, and instrumentation dependencies are operators from taking necessary actions, and adequately captured.
thus could adversely impact fire risk.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
With the above described changes incorporated, this SR is considered to be MET at CAT I/II/III.
ES-A3-01          INCLUDE equipment whose fire-      Dispositioned The "RCP Seal No. 1 Leak-Off" and "SI test line"     RNP-0148, Multiple Spurious Operation Expert induced failure, not including                   are both ISLOCA paths that were screened from        Panel Report, did consider spurious isolation of (NOT MET)         spurious operation, contributes to               the Internal Events analysis; however these          the RCP Seal No. 1 leakoff valves. As noted in or causes                                        screened paths are not explicitly dispositioned in the report, 125V DC power will be removed within (a) fire-induced initiating events               the fire PRA component selection calculation.
treated in the Fire Safe Shutdown                These items were discussed with the host utility
                  / Appendix R analysis                            that provided the following input on the two Revision 0A                                                                                                                                      Page 64 of 79

54002-CALC-01                                                                               Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                  Status                            Finding                                            Disposition (b) Internal Events PRA initiators               ISLOCA paths mentioned above: The seal leak-         ten minutes as noted in RNP Dedicated as identified using the IE                       off is normally open and seal leak-off is part of     Shutdown Procedure (DSP-002).
requirements in Part 2 (including                the normal cooling path. Failure in either direction any gradations across capability                                                                       As noted in Attachment 3 of the RNP Component is non-consequential" and "The test line has 3+
categories in that standard) as                                                                        Selection Calculation, RCP Seal No. 1 Leak-Off modified per 4-2.5, or                           locked closed manual valves." Based on this feedback, it is judged that this is a documentation   is not consequential. Leakoff is not required to (c) unique fire-induced initiating events not addressed or                          issue with no impact on the analysis. This finding    maintain adequate seal cooling in the PRA otherwise screened from the                      is related to the consideration of previously         model. Furthermore, seal leakoff is not modeled above two analyses if SR IE-C4                   screened ISLOCA pathways from the Internal            in the Fire PRA.
in Part 2 cannot be met.                         Events PRA for inclusion in the Fire PRA.
Include, in the component selection calculation, a RESOLUTION OF CAPABILITY CATEGORY discussion of screened paths and why or why not CLASSIFICATION:
they contain components that warrant inclusion in the fire PRA.                                         Based on the above disposition, this SR is considered to be MET at CAT I/II/III.
FQ-B1-01             PERFORM the quantification in      Dispositioned In review of the fire PRA documentation there no      RNP-F/PSA-0077 documented proof of (CAT I/II/III)       accordance with HLR-QU-B and                     convergence study was performed in support of         convergence for CDF and LERF for the internal its SRs in Part 2                                the selection of truncation level for quantification. events PRA. As the fire PRA model is largely and                                                                                                    dependent on the internal events PRA, it can be DEVELOP a defined basis to                                                                             concluded that there is convergence for CDF and support the claim of non-                                                                              LERF in the fire PRA applicability of any of the requirements under HLR-QU-B in                                                                         The model was quantified based on a CDF Part 2.                                                                                                quantification of 1E-12. This is about seven orders of magnitude below the final CDF. This is generally accepted as a good bounding truncation level FQ-F1-01             DOCUMENT the CDF and LERF          Dispositioned The contents of the elements of applicable SRs        The RNP Fire PRA was developed using the (CAT I/II/III)       analyses in accordance with                      of Part 2 were addressed in the FQ and                Internal Events PRA. The Internal Events PRA is HLR-QU-F and HLR-LE-G and                        associated documents; however, no explicit            aligned with RG 1.200 and was peer reviewed.
their SRs in Part 2 with the                     connections were established in the documents         Therefore, the back-references associated with following clarifications:                        to associate with the "back-references"               requirements LE-G2, LE-G4 and LE-G5 are (a) SRs QU-F2 and QU-F3 of                       requirements LE-G2, LE-G4 and LE-G5.                  considered to be met.
Part 2 are to be met including Revision 0A                                                                                                                                          Page 65 of 79

54002-CALC-01                                                                                Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                   Status                             Finding                                       Disposition identification of which fire scenarios and which physical analysis units (consistent with the level of resolution of the Fire PRA such as fire area or fire compartment) are significant contributors (as defined in Part 1);
(b) SR QU-F4 of Part 2 is to be met consistent with 4-2.13 (c) SRs LE-G2 (uncertainty discussion) and LE-G4 of Part 2 are to be met consistent with 4-2.13, and DEVELOP a defined basis to support the claim of non-applicability of any of the requirements under these sections in Part 2.
FSS-B1-01            DEFINE and JUSTIFY the              Dispositioned The conditions and timing that lead to Main        Section 5.9.2 of P2217-1021-01-03 provides a (CAT I/II/III)       conditions that are assumed to                    Control Room (MCR) abandonment due to              discussion of Main Control Room abandonment cause MCR abandonment and/or                      environmental conditions and reliance on ex-       due to environment conditions.
[Associated          reliance on ex-control room                       control room actions are described in main SR: FSS-A6,          operator actions including remote                 control room analysis report (Report               The probability associated with main control room FSS-B2]              and/or alternate shutdown                         0004-0042-412-002).                                abandonment due to a loss of habitability has actions.                                                                                             been incorporated into the quantification of the Abandonment due to equipment damage and to         Fire PRA.
loss of habitability is based on the guidelines provided in Section 11.5.2.11 of NUREG/CR-6850. The analysis method described in NUREG/CR-6850 Appendix L is used to assess scenarios in the MCB.
This SR is considered met. A new Finding (FSS-B1-01) has been assigned to this SR because all conditions requiring reliance on Revision 0A                                                                                                                                      Page 66 of 79

54002-CALC-01                                                                                Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                            Topic                   Status                            Finding                                           Disposition remote/alternate shutdown are not identified.
Fires in the MCR or other location in the plant that may lead to loss of control room functions such that use of remote/alternate shutdown capability is required are not characterized and evaluated. Such scenarios for MCR habitability are identified and analyzed.
FSS-C2-01          CHARACTERIZE ignition source         Dispositioned Section 5.6.2 of the Quantification Calculation       Section 5.6.2 of the Fire PRA Quantification intensity using a realistic time-                  (EPM Report No. P2217-1021-01-01, Rev. 1,             Calculation discusses the growth profile used.
(CAT II/III)       dependent fire growth profile (i.e.,               dated February 2013 included in EC 90905) and         This can also be seen in the RNP_EVAL_Rev2 a time-dependent heat release                      in particular the third bullet under Figure 7 on      spreadsheet used during the quantification rate) for significant contributors                 Page 51 of 113 indicates that "The HGL                process. This spreadsheet includes a cable tray as appropriate to the ignition                     threshold assumes a constant HRR up to HGL            propagation model (HGL_Time worksheet).
source.                                            formation." The basis for using the time versus HGL based on a constant HRR and then                  Furthermore, Calculation NED-M/MECH-1009 compare it to a curve based on a variable HRR         also provides a time to damage based on fire has not been provided. At the top of Page 52 of       growth.
113 in the Quantification Calculation it states that, "The net effect of these uncertainties is generally a conservative time to HGL. For ignition sources that are high risk, more detailed fire modeling may be pursued on a case by case basis."
The intent of the Standard is that conservative =
Category I, and that realistic = Category II/III, therefore this SR is evaluated at Category I. An example of a more realistic analysis is Hughes Report Number: 0004-0042-000-001 for Fire Compartment 20.
Finding FSS-C2-01 has not been resolved. For each scenario, the ignition source intensity is characterized using a time-dependent heat release rate, consistent with the Category II/III requirement for this SR; however, the total heat Revision 0A                                                                                                                                         Page 67 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                         Topic                  Status                            Finding                                         Disposition release required to cause hot gas layer formation is based on a fire that is initiated at full peak intensity, consistent with the Category I requirement for this SR. Suggestion FSS-C2-02 has been prepared to suggest further justification and validation of the methodology used to determine the total heat release required to cause hot gas layer formation.
FSS-C4-01       If a severity factor is credited in Dispositioned The severity factors, as described in Section      Section 5.5.4.5 of the Quantification Calculation the analysis, ENSURE that                         5.5.4.5 of the Quantification Calculation (EPM     (EPM Report No. P2217-1021-01-03) has been (CAT II)                                                          Report No. P2217-1021-01-01, dated                 updated to provide more detail as to why specific
: a. the severity factor remains                 February 2013 included in EC 90905) are            severity factors are applied in the MakeSScen independent of other                        apparently evaluated using the calculation logic   module.
quantification factors                      in Attachment 20 of the Quantification Calculation. This is not explained in a clear and
: b. the severity factor reflects                concise way in Section 5.5.4.5, nor is any the fire event set used to                  reference to Attachment 20 found in that section.
estimate fire frequency                     The general reference to module "MakeSScen" and the fact that its contents are listed in
: c. the severity factor reflects Attachment 20 in the introductory paragraph of the conditions and Section 5.5.4 is not considered to be sufficient.
assumptions of the specific Further, the descriptions in Section 5.5.4.5 fire scenarios under (pages 41-42 of 113) of how the Severity Factor analysis, and determination is done do not appear to exactly
: d. a technical basis supporting                match with the programming logic found on the severity factors                       pages 7-12 of Attachment 20. This needs to be determination is provided.                  clarified and clearly documented to support future use, update, and peer review of the Severity Factor calculations.
In addition, the use of generic fire modeling data and severity factors for different ignition source Bins from NUREG/CR-6850 without considering mode of exposure and position (i.e., not just Revision 0A                                                                                                                                    Page 68 of 79

54002-CALC-01                                                                              Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                              Topic                 Status                           Finding                                              Disposition distance) of the targets relative to the fire source may not fully constitute "explicit consideration" in quantifying the severity factor such that it reflects the conditions and assumptions of the specific fire scenario under analysis.
FSS-C7-01            If multiple suppression paths are  Dispositioned Credit is given for both automatic sprinkler          No credit has been applied for manual actuation credited, EVALUATE and                           suppression and manual firefighting, but mutual       or recovery of fixed suppression systems. The (NOT MET)            PROPERLY MODEL                                   dependency on the common water supply system          common water supply at RNP is sufficient to dependencies among the                           has not been evaluated or properly modeled.           provide water to both the automatic suppression
[Associated          credited paths including                                                                               system as well as the manual firefighting. The SR: FSS-D6]          dependencies associated with                                                                           reliability/unavailability of the fire pump and recovery of a failed fire                                                                              associated sprinkler has already been accounted suppression system, if such                                                                            for in the non-suppression probability.
recovery is credited.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
Based on the above disposition, SR FSS-C7 is considered to be MET at CAT I/II/III.
FSS-D2-01            USE fire models that have          Dispositioned This finding recommends that the HGL                  Attachment J of the LAR provides a discussion sufficient capability to model the               calculation (RNP-M/MECH-1826) be subjected to         on the software used during the development of (CAT I/II/III)       conditions of interest and only                  validation and verification in order to establish its the RNP HGL Calculation (RNP-M/MECH-1826).
within known limits of                           technical basis and known limits of applicability.
[Associated          applicability.
SR: FSS-D6]
FSS-D7-01            In crediting fire detection and    Dispositioned There is a failure to meet the Category I             Currently, system performance is monitored and (NOT MET)            suppression systems, USE                         requirement of having systems installed and           maintained at a high level as part of the System generic estimates of total system                maintained in accordance with applicable codes        Health Reporting and System Notebook unavailability provided that                     and standards. The Main Turbine Lube Oil              processes. Outlier behavior with respect to (a) the credited system is                       Deluge system must be replaced to account for         system availability would be evident to the installed and maintained in                      system deficiencies identified in NCR-425437          system engineer and plant management through accordance with applicable                       where a simultaneous actuation of the Turbine         the health data (available for the previous codes and standards                              Lube Oil suppression system, along with the Revision 0A                                                                                                                                             Page 69 of 79

54002-CALC-01                                                                            Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                          Topic                  Status                           Finding                                           Disposition (b) the credited system is in a                   mezzanine and ground level sprinkler systems,       12 months), which indicates overall Excellent fully operable state during plant                 could place a higher system demand on the           (Green) performance.
operation, and                                    water supply than can be provided by a single fire (c) the system has not                            pump.                                               Furthermore, during the periods when key fire experienced outlier behavior                                                                          protection systems are unavailable due to testing relative to system unavailability.                This was not identified, nor is a comparison        and maintenance, compensatory actions are provided in the Fire PRA of all installed detection taken such that the risk associated with the and suppression systems vs. the corresponding       system being unavailable does not increase.
Code Compliance calculation.
RESOLUTION OF CAPABILITY CATEGORY Finding FSS-D7-01 has not been resolved.            CLASSIFICATION:
Because evidence is not provided to support that    Using plant-specific information to quantify total credited detection/suppressions systems are         unavailability factors is a CAT III requirement and installed and maintained in accordance with         was not done.
applicable codes and standards. System health report for period Q2-2013 for systems               With no change being made, RNP considers the 6185/6181/6175/6195/6205/6180 notes that age,       risk results from the Fire PRA to be creditable for obsolescence and replacement part procurement       the NFPA 805 application and this finding to be is an issue. This system health report also notes   sufficiently resolved for SR FSS-D7 to be that "There are LTAMs budgeted for 2014 and         assessed as CAT II is MET.
2015 which study and replace the detection, CO2, and Halon Systems." This report suggests that some of the fire protection systems at RNP may be experiencing outlier behavior relative to system unavailability and may not be in a fully operable state during plant operation.
Consequently, this SR is still considered to be not met.
FSS-D7-03       In crediting fire detection and     Dispositioned Evidence needs to be provided to support that       Currently, system performance is monitored and (NOT MET)       suppression systems, USE                          credited detection/suppressions systems are         maintained at a high level as part of the System generic estimates of total system                 installed and maintained in accordance with         Health Reporting and System Notebook unavailability provided that (a)                                                                      processes. Outlier behavior with respect to applicable codes and standards. System health the credited system is installed                                                                      system availability would be evident to the and maintained in accordance                      report for period Q2-2013 for systems               system engineer and plant management through with applicable codes and                         6185/6181/6175/6195/6205/6180 notes that age,       the health data (available for the previous 12 standards, and (b) the credited                   obsolescence and replacement part procurement       months), which indicates overall Excellent system is in a fully operable state               is an issue for multiple fire protection systems.   (Green) performance.
during plant operation.                           This system health report also notes that "There Revision 0A                                                                                                                                     Page 70 of 79

54002-CALC-01                                                                          Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                 Status                        Finding                                         Disposition are LTAMs budgeted for 2014 and 2015 which        Furthermore, during the periods when key fire study and replace the detection, CO2, and Halon   protection systems are unavailable due to testing Systems." This report suggests that some of the   and maintenance, compensatory actions are taken such that the risk associated with the fire protection systems at RNP may be system being unavailable does not increase.
experiencing outlier behavior relative to system unavailability and may not be in a fully operable RESOLUTION OF CAPABILITY CATEGORY state during plant operation.                     CLASSIFICATION:
Using plant-specific information to quantify total unavailability factors is a CAT III requirement and was not done.
With no change being made, RNP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application and this finding to be sufficiently resolved for SR FSS-D7 to be assessed as CAT II is MET.
FSS-E1-01           For any fire modeling parameters   Open       Section 4.3 of Calculation No. P2217-2100-00,     Section 4.3 of Calculation P2217-2100-01-01 will (CAT I/II/II)       not covered by the requirements               Fire Scenario Data, RNP-F/PSA-0079, dated         be updated appropriately at a later date. This is a of HLR-FSS-C or HLR-FSS-D,                    January 2013 contains information about fire      documentation issue that will not impact fire USE plant-specific parameter modeling parameters that were used. However,      scenario development or quantification.
estimates for fire modeling if available, or use generic                     Section 4.4 through 4.7 should be completed information modified as                       because they are missing information about other discussed in SR FSS-E2; USE                   relevant fire modeling parameters.
generic information for the remaining parameter estimates.                Add relevant information to the report.
Finding FSS-E1-01 has not been resolved.
Sections 4.3 through 4.7 still make reference to databases for the parameters used in the fire modeling. These parameters should be added to the report.
FSS-E3-01           PROVIDE a mean value of, and          Open    No statistical representation of uncertainty      RNP used the HRRs and applied them using the (CAT I)             statistical representation of, the            intervals (e.g., NUREG/CR-6850 Table E-1 or       guidance found in NUREG/CR-6850. As uncertainty intervals for the Revision 0A                                                                                                                                   Page 71 of 79

54002-CALC-01                                                                          Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                          Topic                 Status                           Finding                                          Disposition parameters used for modeling                    G-1 for HRR, Tables E-2 through E-9 for severity  NUREG/CR-6850 is the consensus methodology, the significant fire scenarios.                 factor) is documented for the mean values of      a detailed uncertainty analysis on these parameter estimates used for fire modeling the    parameters is not needed and does not add to significant fire scenarios.                       the credibility of the results. The majority of applied values are based on the 98th and 75th percentile fires from NUREG/CR-6850, and the ZOIs are applied conservatively. It is not believed that reducing these values would allow the use of reduced impacts for the applications being pursued.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
Although no change has yet been made that would improve the Capability Category assessments, RNP considers the risk results from the Fire PRA to be creditable for the NFPA 805 application because documenting the statistical representation of uncertainty intervals will not change the quantified risk metrics.
FSS-E4-01         PROVIDE a characterization of     Dispositioned There is some assumed cable routing applied       As noted in RNP Change Package RNP-0152 (NOT MET)         the uncertainties associated with               directly in FSSPMD. These routings are identified (Attachment 19 of P2217-1021-01-03), cable cases where cable routing has                   in RNP-0152. Components with unknown routing      toning was used to confirm cable routes. There
[Associated       been assumed based on SRs                                                                         were some instances where cable toning was not are otherwise assumed to fail in the PRA model.
SR: FSS-A3]       CS-A10 and/or CS-A11.                                                                             possible within a specific compartment However, there is no discussion that              (embedded cable, etc.). In these instances, the characterizes the uncertainties associated with   cable was assumed to be failed throughout the cases where cable routing has been assumed. A     entire compartment that it was known to traverse more detailed characterization of the             through.
uncertainties associated with cases where cable routing has been assumed is needed for SR         As noted in RNP change package RNP-0205 (Attachment 19 of P2217-1021-01-03), the FSS-E4 to be met.
assumed cable route data determines the cable-to-fire zone correlation (which is sufficient for NSCA), but does not determine the cable-to_raceway-to-fire zone correlation (which Revision 0A                                                                                                                                    Page 72 of 79

54002-CALC-01                                                                               Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                              Topic                  Status                           Finding                                          Disposition Provide a detailed uncertainty characterization    is needed for PRA). Based on this assumption, discussion in appropriate reports related to       any ignition source within a given fire zone will circuits with assumed routing.                     impact all cables with assumed cable routes in the ignition sources fire zone.
RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
Based on the above disposition, SR FSS-EF-04 is considered to be MET at CAT I/II/III.
FSS-F3-02            In crediting fire detection and     Dispositioned Section 2.4 of the Structural Steel report         The unavailability of the Turbine Lube Oil Deluge (CAT I/II/III)       suppression systems, USE                          (P2217-2300-01-03) states that "It is assumed      system has been updated to 0.05 in Rev. 3 of generic estimates of total system                 that an unavailability value of 0.01 will be       P2217-1021-01-03 (Fire PRA Quantification
[Associated          unavailability provided that (a) bounding and conservative for the deluge           Calculation). This change was made based on SR: FSS-D7]          the credited system is installed and maintained in accordance                      sprinkler system." What is the basis for this? Has engineering judgment.
with applicable codes and                         this value been confirmed against the plant-standards, and (b) the credited                   specific experience for availability of the system is in a fully operable state               detection/suppression systems?
during plant operation.
FSS-G1-01            APPLY all the supporting            Dispositioned Fire modeling performed in support of the Multi-   A detailed multi-compartment analysis was (NOT MET)            requirements listed in SRs FSS-                   Compartment Analysis (MCA) failed to consider      performed and is documented in C1 through FSS-C8 for fire                        the acute effect of hot gas flow through openings  RNP-F/PSA-0089. The Fire PRA Quantification
[Associated modeling of single physical                       and ducts on local potential targets. In addition, Calculation only updated results based on new SR: FSS-G2, FSS-G3]              analysis units to the modeling of                 the consideration and evaluation of additional     CCDPs and a new HGL Calculation. The open multi-compartment fire scenarios.                 aspects of multi-compartment fire scenarios was    Turbine Building was discussed in the HGL not documented and/or done, i.e., the fact that    Calculation (RNP-M/MECH-1826). The detailed the Turbine Building has no exterior walls was     review of hot gas flow through openings has not not identified and the impact of this evaluated. been performed. The impact of hot gas flow Treatment of MCA in only 7 pages of the            through openings and ducts on local targets is Quantification Calculation does not appear to be   expected to be minimal.
adequate. Furthermore, in their evaluation of compartments subject to a hot gas layer, RNP       RESOLUTION OF CAPABILITY CATEGORY used a criterion of 1E-07 for total ignition       CLASSIFICATION:
Based on the above disposition, SR FSS-G1-01 frequency. Based on this, RNP excluded all but is considered to be MET at CAT I/II/III.
two compartments. A review of the table that was Revision 0A                                                                                                                                        Page 73 of 79

54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                        Topic                 Status                            Finding                                           Disposition subject to the review showed there were two additional compartments that met the criterion for inclusion. This is an error rate of 100%.
FSS-G6-02       QUANTIFY the risk                 Dispositioned Table B-5 of this report defines lists fourteen (14) Section 5.10 of P2217-1021-01-03 provides a (CAT I)         contribution of any selected                    unscreened multi-compartment fire scenarios.         discussion of the multi-compartment scenarios multi-compartment fire scenarios                Attachment E of the same report offers risk          quantified in the Fire PRA. There are five in a manner consistent with the associated with these scenario. The                  scenarios that do not screen currently based on FQ requirements.
Quantification Report (P2217-1021-01-02)             an updated HGL Calculation Section 5.10 states that "Based on updated Hot       (RNP-M/MECH-1826, Rev. 0). These scenarios Gas Layer frequencies, all of the fire               have been quantified in a manner consistent with compartments previously analyzed in the Multi-       the FQ requirements.
Compartment Analysis (Reference 3.14), have now screen." Reference 3.14 is                       RESOLUTION OF CAPABILITY CATEGORY RNP-F/PSA-0089 Rev 0 and it does not support         CLASSIFICATION:
Based on the above disposition, SR FSS-G6 is the statement that all MC fire is screened.
considered to be Met at CAT II/III.
Finding FSS-G6-01 has been resolved. Hughes Calculation No. 1RCS04042.414.031, Revision 0 (Multi-Compartment Analysis for the Fire Probabilistic Risk Assessment at Robinson Nuclear Plant), provides a detailed multi-compartment analysis. This analysis has been included in Section 5.10 of the RNP Fire PRA Quantification Calculation. Based on updated Hot Gas Layer data, the scenarios identified in the Hughes calculation screen from further analysis.
FSS-H2-01       DOCUMENT a basis for target       Dispositioned The fire modeling in Hughes Report No.               Plant modifications in addition to intumastic cable (CAT I)         damage mechanisms and                           0004-0042-000-001 for Fire Compartment 20            coating are incorporated into the Fire PRA thresholds used in the analysis,                credits fire coating. Category II of this SR         Calculation (EPM Report No.
including references for any                                                                         P2217-1021-01-03). These plant modifications requires documentation of the references for any plant-specific or target-specific                                                                    are discussed in Attachment S of the LAR. The performance criteria applied in                 plant-specific or target-specific performance        ten minute time until cable damage (per the analysis.                                   criteria applied in the analysis, and a basis for Revision 0A                                                                                                                                    Page 74 of 79

54002-CALC-01                                                                           Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                         Topic                   Status                           Finding                                          Disposition target damage mechanisms and thresholds used       NUREG/CR-6850) is achieved via plant in the analysis, which has not been provided.      modifications discussed in Attachment S.
Finding FSS-H2-01 has not been resolved.
Plant-specific documentation should be provided    RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION:
for the performance criteria used to evaluate Based on the above disposition, SR FSS-H2 is damage to coated cables and cable trays with       considered to be Met at CAT II/III.
solid bottoms.
HRA-C1-01         For each selected fire scenario,    Dispositioned The fire-specific operator actions are evaluated   All fire response actions have been updated such QUANTIFY the HEPs for all                         using only the Cause-Based Decision Tree           that they are quantified using the CBDTM and (CAT II)          HFEs and ACOUNT FOR                               method (CBDTM). The HRAs from the Internal         HCR/ORE combination method similar to Internal relevant fire-related effects using               Events methods used both CBDTM and                 Events operator actions. P2217-1022-01-03
[Associated       detailed analyses for significant                 HCR/ORE to address the cognitive risk of the       documents the Fire HRA.
with FQ-C1]       HFEs and conservative estimates                   operator actions. The HCR/ORE method is (e.g., screening values) for non-                 generally the dominant risk value for actions with significant HFEs, in accordance                   short system time windows or very long median with the SRs for HLR-HR-G in                      response times. Events 0FIREOMA01 and Part 2 set forth under at least                   0FIREOMA02 are events with short time windows Capability Category II, with the                  where the HCR/ORE method would be the following clarification:                          dominant cognitive risk. The cognitive risk is underestimated by an order of magnitude.
: a. Attention is to be given to how the fire situation alters                The HRA needs to be revised to better address any previous assessments in                  the cognitive risk portion for each HEP.
non-fire analyses as to the influencing factors and the timing considerations covered in SRs HR-G3, HR-G4, and HR-G5 in Part 2 and Revision 0A                                                                                                                                   Page 75 of 79

54002-CALC-01                                                                                   Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                    Status                              Finding                                          Disposition
: b. Develop a defined basis to support the claim of non-applicability of any of the requirements under HLR-HR-G in Part 2.
IGN-A7-01            USE a plant-wide consistent          Dispositioned In general the SR was met. There are areas such      Transient fire scenarios have been postulated in methodology based on                               as the transient ignition sources related            each fire compartment as noted in Hughes (CAT I/II/III)       parameters that are expected to                    identification are still required, some of these is  Calculation No. 1RCS04042.414.031-002.
influence the likelihood of ignition               already self-identified as part of the "lead review" to apportion high-level ignition                   activities as indicated in the -0067 notebook. For frequencies (e.g., plant-wide                      instance, fire zone 22 and others are identified in values) to estimate physical                       the "lead reviews" as additional effort is needed.
analysis unit or ignition source level frequencies.                                 Address the transient ignition sources for application fire zones. Providing a complete and through ignition sources identification is important.
IGN-A9-01            POSTULATE the possibility of         Dispositioned RNP did postulate transient combustibles for all     Transient fire ignition frequency has been transient combustible fires for all                physical analysis units except for one, FC490.       assigned to FC490. Transient Influence Factors (CAT I/II/III)       physical analysis units regardless                 This physical analysis unit is the deepwell pump     of Low(1), Low(1), and Low(1) replaced factors of the administrative restrictions.                D enclosure. RNP needs to provide justification      No(0), No(0), and No(0) for maintenance, for why transient combustibles are not postulated    occupancy, and storage, since entry to FC490, for FC490.                                           which requires using a crane to remove the concrete enclosure, is not precluded.
PP-B6 (01)           ENSURE                               Dispositioned Underground cabling was not addressed in             The cable routing database system FSSPMD RNP-F/PSA-0067. The RNP project team noted a         was reviewed determining which manholes had (CAT I/II/III)       a. that collectively, the defined                comment was included in this calculation that a      cables within the scope of the Fire PRA physical analysis units                       Yard fire zone should be included in the next      associated with them. The manholes where encompass all locations                       revision of the report.                              cables were identified have been included as plant partitioning elements (i.e., Physical Analysis Revision 0A                                                                                                                                            Page 76 of 79

54002-CALC-01                                                                         Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                         Topic                  Status                           Finding                                         Disposition within the global analysis                                                                   Units) in the Fire PRAs. The following manholes boundary                                                                                     have been included as plant partitioning and                                                                                              elements: MH M-34, MH M-35, MH-1, and MH-34. The cables that were identified as routed
: b. that defined physical                                                                        through MH-35, MH-36 and MH have been analysis units do not overlap                                                                included in the Intake Structure Fire Compartment, FC290. The cable loading for the manholes have been assessed following the same approach documented in the RNP combustible loading calc. The approach consists of multiplying the factor of 5,515 BTU/ln-ft to each linear ft of cable. Under this approach, a total of 496350 BTUs have been estimated for each manhole. This assumes nine 10 long exposed cables per manhole. Nine cables is the average number of cables identified in the manholes.
PRM-B11-01        MODEL all operator actions and    Dispositioned Main control room abandonment is discussed in    Fire response and Main Control Room operator influences in                          Section 7.1.4 of Calculation P2217-1022-01-01,   Abandonment HFEs have not been incorporated (NOT MET)         accordance with the HRA                         "Fire Human Reliability Analysis" and Table 7.1  in the combined Internal Events/Fire PRA. The element of this Standard.
lists the human failure events (HFEs) associated Main Control Room Abandonment HFEs from
[Associated                                                       with main control room abandonment. However,     Hughes Calculation No.
with HRA-B2                                                       these HFEs have not yet been incorporated into   RSC-CALKNX-2013-0301 has been incorporated and PRM-B2]                                                       the RNP fire PRA model. Without the HFEs for     into the recovery rule files used during the main control room abandonment, the RNP fire      quantification process as appropriate based on PRA is incomplete.                               the fire compartment being quantified.
Incorporate the main control room abandonment    RESOLUTION OF CAPABILITY CATEGORY HFEs into the fire PRA model.                    CLASSIFICATION:
Based on the abovementioned details, this SR is considered to be MET at CAT I/II/III.
Revision 0A                                                                                                                                  Page 77 of 79

54002-CALC-01                                                                                  Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                             Topic                   Status                             Finding                                        Disposition PRM-B15-01           MODEL any new accident              Dispositioned Quantification of the RNP fire PRA model            Following additional refinements to the model, progressions beyond the onset of                  indicates that the fire-related LERF is about       LERF is more in line with typical results.
(CAT I/II/III)       core damage identified per                        90 percent of the fire-related CDF. This result is PRM-B13 to determine the                          exceptionally high and has been self-identified by fire-induced LERF in accordance                   the utility. The result is produced by MSO events with HLR-LE-A, HLR-LE-B,                          in the switchgear and cable spreading rooms that HLR-LE-C, and HLR-LE-D and                        cause a core damaging accident sequence and their SRs in Part 2 with the                      also result in the spurious opening of several following clarifications:                         containment isolation valves. Fire-related LERF (a) All the SRs under HLR-LE-A,                   for RNP is an unusually high proportion of the HLR-LE-B, HLR-LE-C, and HLR-                      fire-related CDF. Typically, fire-related LERF is LE-D in Part 2 are to be                          10 - 25 percent of the fire-related CDF. At RNP, addressed in the context of fire                  fire-related LERF is about 90 percent of the fire-scenarios including effects on                    related CDF.
system operability / functionality, operator actions, accident                        Investigate ways to remove conservatisms from progression, and possible                         the RNP fire PRA model, particularly for the containment failures accounting                   MSOs in the switchgear and cable spreading for fire damage to equipment and                  rooms.
associated cabling.
(b) LE-C2 and LE-C6 in Part 2 are to be met in a manner consistent with 4-2.10.
(c) LE-C6 in Part 2 is to be met in a manner consistent with PRM-B9 above.
(d) LE-C8 in Part 2 is to be met in a manner consistent with PRM-B6 above.
and DEVELOP a defined basis to support the claim of nonapplicability of any of these requirements in Part 2.
PRM-C1-02            DOCUMENT the Fire PRA plant         Dispositioned Documentation of development of the plant           The development of the plant response model is response model in a manner                        response model is spread over several               documented in Attachment 9 of RNP-/PSA-0066 (CAT I/II/III)       consistent with HLR-IE-D,                         calculations and documents. The primary             via a model change log. RNP-F/PSA-0066 also HLRAS-C, HLR-SC-C, HLR-SY-                        documents are RNP-F/PSA-0066 (component Revision 0A                                                                                                                                         Page 78 of 79

54002-CALC-01                                                                                 Evaluation of Risk Significance of Permanent ILRT Extension Table 2 Fire PRA Peer Review - Facts and Observations (with RESOLUTION OF CAPABILITY CATEGORY CLASSIFICATION For Non-CAT II)
SR                              Topic                  Status                             Finding                                         Disposition C, and HLR-DA-E and their SRs                     selection) and P2217-1021-01-01 (quantification). discusses assumptions made during the in Part 2 as well as 4-2.10 with                  Thus, key elements of the PRM development           development of the model.
the following clarifications:                     (e.g., assumptions that are made while (a) HLR-IE-D in Part 2 is to be                   developing the model) being dispersed and met in a manner consistent with                   difficult to assimilate and comprehend. The that required under HLR-IGN-B                     dispersion of pertinent information about of this Standard.                                 development of the PRM makes understanding of (b) Document any defined bases                    the development difficult for reviewers and for to support the claim of                           utility staff who will make future modifications to nonapplicability of any of the                    the RNP fire PRA model and documentation.
referenced requirements in Part 2 beyond that already covered by                    Develop a PRM-specific calculation or notebook the clarifications in this section.               which combines the pertinent portions of PRM documentation from existing documents or calculations.
UNC-A2               INCLUDE the treatment of            Dispositioned To comply with Section 2 applicable SRs intent      Section 7.0 of the RNP Fire PRA Quantification uncertainties, including their                    for UNC, a final documentation of importance        Calculation, provides an analysis of uncertainty (CAT I/II/III)       documentation, as called out in                   rankings considering various sensitivity studies    regarding the quantification of the RNP Fire PRA.
SRs PRM-A4, FQ-F1, IGN-A10,                       along with applicable sequences is suggested to     Importance rankings are also discussed in this IGN-B5, FSS-E3, FSS-E4, FSS-                      be included in the UNC discussion in an             section. Additionally, Section 9.0 provides a H5, FSS-H9, and CF-A2 and that                    independent document.                               discussion of sensitivities that were evaluated.
required by performing Part 2 referenced requirements throughout this Standard.
Revision 0A                                                                                                                                          Page 79 of 79

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 7 Pages (including this cover page)
E-C AUGMENT PLAN TABLES FOR THE SECOND TEN-YEAR INTERVAL

H. B. Robinson Steam Electric Plant, Unit No. 2 IWE/IWL Second Ten-Year Interval Inservice Inspection Plan 11.0 Attachment E-C AUGMENT Plan Tables This Attachment contains the E-C AUGMENT Plan Tables for the Second Ten-Year Interval RNP-PM-007                                                     Revision 5                Page 9 of 12

H. B. ROBINSON UNIT 2                    CATEGORY - E-C AUGMENT                                 SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary       Scope     Drawing Item No     Comp ID       Method    Component Description                         Period Class       System                  Code Case/Relief Request                    1   2 3                         Comments E-C         579800        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 240-Y  GENERAL - PANEL ADJACENT TO 240-X(L) & 240-Z(R)                  1128 MC          8010          DETAILED  RR-01 E-C         584500        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 248-BB GENERAL - PANEL ADJACENT TO 248-AA(L) & 248-CC(R)                1128 MC          8010          DETAILED  RR-01 E-C         590600        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 256-SS GENERAL - PANEL ADJACENT TO 256-RR(L) & 256-TT(R)                1128 MC          8010          DETAILED  RR-01 E-C         563700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 264-GG GENERAL - PANEL ADJACENT TO 264-FF(L) & 264-HH(R)                1128 MC          8010          DETAILED  RR-01 E-C         597600        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 268-DD GENERAL - PANEL ADJACENT TO 268-CC(L) & 268-EE(R)                1128 MC          8010          DETAILED  RR-01 E-C         597700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 268-EE GENERAL - PANEL ADJACENT TO 268-DD(L) & 268-FF(R)                1128 MC          8010          DETAILED  RR-01 E-C         597900        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 268-GG GENERAL - PANEL ADJACENT TO 268-FF(L) & 268-HH(R)                1128 MC          8010          DETAILED  RR-01 E-C         601000        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 272-FF GENERAL - PANEL ADJACENT TO 272-EE(L) & 272-GG(R)                1128 MC          8010          DETAILED  RR-01 E-C         601100        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 272-GG GENERAL - PANEL ADJACENT TO 272-FF(L) & 272-HH(R)                1128 MC          8010          DETAILED  RR-01 E-C         602300        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 272-TT GENERAL - PANEL ADJACENT TO 272-SS(L) & 272-UU(R)                1128 MC          8010          DETAILED  RR-01 E-C         569800        IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-A  GENERAL - PANEL ADJACENT TO 280-CCC(L) & 280-B(R)                CALCULATION RNP-C/STRU-1128 MC          8010          DETAILED  RR-01 E-C         570200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-E  GENERAL - PANEL ADJACENT TO 280-D(L) & 280-F(R)
MC          8010          DETAILED  RR-01 RNP-PM-007                                                           Revision 5                                                      Page 1 of 5

H. B. ROBINSON UNIT 2                     CATEGORY - E-C AUGMENT                                 SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary        Scope     Drawing Item No     Comp ID        Method    Component Description                         Period Class       System                   Code Case/Relief Request                    1   2 3                          Comments E-C         570400         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-G   GENERAL - PANEL ADJACENT TO 280-F(L) & 280-H(R)
MC          8010           DETAILED  RR-01 E-C         570700         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-L   GENERAL - PANEL ADJACENT TO 280-K(L) & 280-M1(R)
MC          8010           DETAILED  RR-01 E-C         570800         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-M1  GENERAL - PANEL ADJACENT TO 280-L(L) & 280-M2(R)
MC          8010           DETAILED  RR-01 E-C         570900         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-M2  GENERAL - PANEL ADJACENT TO 280-M1(L) & 280-M3(R)
MC          8010           DETAILED  RR-01 E-C         631200         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 280-P   GENERAL - PANEL ADJACENT TO 280-O(L) & 280-Q(R)                  1128 MC          8010           DETAILED  RR-01 E-C         633400         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-LL  GENERAL - PANEL ADJACENT TO 280-KK(L) & 280-MM(R)
MC          8010           DETAILED  RR-01 E-C         633500         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 E4.11       L-WALL 280-MM  GENERAL - PANEL ADJACENT TO 280-LL(L) & 280-NN(R)
MC          8010           DETAILED  RR-01 E-C         572800         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      BULGES AND PITTING EVALUATED UNDER EC 72699, RNP E4.11       L-WALL 280-SS  GENERAL - PANEL ADJACENT TO 280-RR(L) & 280-TT(R)                CALCULATION RNP-C/STRU-1128 AND RNP-C/STRU-1130 MC          8010           DETAILED  RR-01 E-C         572900         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      BULGES AND PITTING EVALUATED UNDER EC 72699, RNP E4.11       L-WALL 280-TT  GENERAL - PANEL ADJACENT TO 280-SS(L) & 280-UU(R)                CALCULATION RNP-C/STRU-1128 AND RNP-C/STRU-1130 MC          8010           DETAILED  RR-01 E-C         573000         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-UU  GENERAL - PANEL ADJACENT TO 280-TT(L) & 280-VV(R)                CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         573100         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-VV  GENERAL - PANEL ADJACENT TO 280-UU(L) & 280-WW(R)                CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         573400         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-AAA GENERAL - PANEL ADJACENT TO 280-ZZ(L) & 280-BBB(R)               CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 RNP-PM-007                                                            Revision 5                                                      Page 2 of 5

H. B. ROBINSON UNIT 2                     CATEGORY - E-C AUGMENT                                 SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary        Scope     Drawing Item No     Comp ID        Method    Component Description                         Period Class       System                   Code Case/Relief Request                    1   2 3                         Comments E-C         573500         IWE       HBR2-10618 SHT 195, FIGURE 14               c   H      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 280-BBB GENERAL - PANEL ADJACENT TO 280-AAA(L) & 280-CCC(R)              CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         571200         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 284-B   GENERAL - PANEL ADJACENT TO 284-A(L) & 284-C(R)                  CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         571800         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 284-H   GENERAL - PANEL ADJACENT TO 284-G(L) & 284-I(R)                  1128 MC          8010           DETAILED  RR-01 E-C         571900         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 284-I   GENERAL - PANEL ADJACENT TO 284-H(L) & 284-J(R)                  1128 MC          8010           DETAILED  RR-01 E-C         575300         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 284-UU  GENERAL - PANEL ADJACENT TO 284-TT(L) & 284-VV(R)                1128 MC          8010           DETAILED  RR-01 E-C         637000         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 284-VV  GENERAL - PANEL ADJACENT TO 284-UU(L) & 284-WW(R)                1128 MC          8010           DETAILED  RR-01 E-C         575500         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 284-BBB GENERAL - PANEL ADJACENT TO 284-AAA(L) & 284-CCC(R)              CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         573700         IWE       HBR2-10618 SHT 195, FIGURE 14               c   h      PITTING EVALUATED UNDER EC 72699 AND RNP E4.11       L-WALL 284-CCC GENERAL - PANEL ADJACENT TO 284-BBB(L) & 284-A(R)                CALCULATION RNP-C/STRU-1128 MC          8010           DETAILED  RR-01 E-C         638100         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-G   GENERAL - PANEL ADJACENT TO 288-F(L) & 288-H(R)                  1128 MC          8010           DETAILED  RR-01 E-C         638300         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-I   GENERAL - PANEL ADJACENT TO 288-H(L) & 288-J(R)                  1128 MC          8010           DETAILED  RR-01 E-C         639300         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-P   GENERAL - PANEL ADJACENT TO 288-O(L) & 288-Q(R)                  1128 MC          8010           DETAILED  RR-01 E-C         642300         IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-TT  GENERAL - PANEL ADJACENT TO 288-SS(L) & 288-UU(R)                1128 MC          8010           DETAILED  RR-01 RNP-PM-007                                                            Revision 5                                                     Page 3 of 5

H. B. ROBINSON UNIT 2                    CATEGORY - E-C AUGMENT                                SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary       Scope     Drawing Item No     Comp ID       Method    Component Description                         Period Class       System                  Code Case/Relief Request                    1   2 3                        Comments E-C         642400        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-UU GENERAL - PANEL ADJACENT TO 288-TT(L) & 288-VV(R)                1128 MC          8010          DETAILED  RR-01 E-C         642500        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 288-VV GENERAL - PANEL ADJACENT TO 288-UU(L) & 288-WW(R)                1128 MC          8010          DETAILED  RR-01 E-C         644100        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 292-I  GENERAL - PANEL ADJACENT TO 292-H(L) & 292-J(R)                  1128 MC          8010          DETAILED  RR-01 E-C         648000        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 292-UU GENERAL - PANEL ADJACENT TO 292-TT(L) & 292-VV(R)                1128 MC          8010          DETAILED  RR-01 E-C         648100        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 292-VV GENERAL - PANEL ADJACENT TO 292-UU(L) & 292-WW(R)                1128 MC          8010          DETAILED  RR-01 E-C         648200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 292-WW GENERAL - PANEL ADJACENT TO 292-VV(L) & 292-XX(R)                1128 MC          8010          DETAILED  RR-01 E-C         649500        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 296-G  GENERAL - PANEL ADJACENT TO 296-F(L) & 296-H(R)                  1128 MC          8010          DETAILED  RR-01 E-C         649600        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 296-H  GENERAL - PANEL ADJACENT TO 296-G(L) & 296-I(R)                  1128 MC          8010          DETAILED  RR-01 E-C         649700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 296-I  GENERAL - PANEL ADJACENT TO 296-H(L) & 296-J(R)                  1128 MC          8010          DETAILED  RR-01 E-C         650700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 296-P  GENERAL - PANEL ADJACENT TO 296-O(L) & 296-Q(R)                  1128 MC          8010          DETAILED  RR-01 E-C         655400        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 300-H  GENERAL - PANEL ADJACENT TO 300-G(L) & 300-I(R)                  1128 MC          8010          DETAILED  RR-01 E-C         659700        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 300-VV GENERAL - PANEL ADJACENT TO 300-UU(L) & 300-WW(R)                1128 MC          8010          DETAILED  RR-01 RNP-PM-007                                                           Revision 5                                                    Page 4 of 5

H. B. ROBINSON UNIT 2                    CATEGORY - E-C AUGMENT                                 SECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLAN Category    Summary       Scope     Drawing Item No     Comp ID       Method    Component Description                         Period Class       System                  Code Case/Relief Request                    1   2 3                         Comments E-C         661300        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 304-I  GENERAL - PANEL ADJACENT TO 304-H(L) & 304-J(R)                  1128 MC          8010          DETAILED  RR-01 E-C         665200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130 E4.11       L-WALL 304-TT GENERAL - PANEL ADJACENT TO 304-SS(L) & 304-UU(R)
MC          8010          DETAILED  RR-01 E-C         655300        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 304-UU GENERAL - PANEL ADJACENT TO 304-TT(L) & 304-VV(R)                1128 AND BULGE EVALUATED IN RNP CALCULATION RNP-MC          8010          DETAILED  RR-01                                                  C/STRU-1130 E-C         655400        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130 E4.11       L-WALL 304-VV GENERAL - PANEL ADJACENT TO 304-UU(L) & 304-WW(R)
MC          8010          DETAILED  RR-01 E-C         666900        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-H  GENERAL - PANEL ADJACENT TO 308-G(L) & 308-I(R)                  1128 MC          8010          DETAILED  RR-01 E-C         668200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-R  GENERAL - PANEL ADJACENT TO 308-Q(L) & 308-S(R)                  1128 MC          8010          DETAILED  RR-01 E-C         668900        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-Y  GENERAL - PANEL ADJACENT TO 308-X(L) & 308-Z(R)                  1128 MC          8010          DETAILED  RR-01 E-C         671000        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-TT GENERAL - PANEL ADJACENT TO 308-SS(L) & 308-UU(R)                1128 AND BULGE EVALUATED IN RNP CALCULATION RNP-MC          8010          DETAILED  RR-01                                                  C/STRU-1130 E-C         575600        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-UU GENERAL - PANEL ADJACENT TO 308-TT(L) & 308-VV(R)                1128 MC          8010          DETAILED  RR-01 E-C         671100        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-VV GENERAL - PANEL ADJACENT TO 308-UU(L) & 308-WW(R)                1128 MC          8010          DETAILED  RR-01 E-C         671200        IWE       HBR2-10618 SHT 195, FIGURE 14                   c    h CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-E4.11       L-WALL 308-WW GENERAL - PANEL ADJACENT TO 308-VV(L) & 308-XX(R)                1128 MC          8010          DETAILED  RR-01 RNP-PM-007                                                           Revision 5                                                      Page 5 of 5

U. S. Nuclear Regulatory Commission  to Serial: RNP-RA/15-0090 2 Pages (including this cover page)
CONTAINMENT LINER DETAIL AT MOISTURE BARRIER

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==REFERENCES:==
: 1) G-190343
: 2) G-190353
: 3) G-190358
: 4) G-190359 fl      .~ ~                          - - - FLOOR  SLAB             5) RELIEF REQUEST IWE/IWL-02 FIGURE 1 EL 226'-0" -   --
                                                 .       })   ~~~~~~~D.Oi~::X:X:X:Z:~ X"       TOP OF MAT LINER BASE MAT - - -
0    ,,                EC 65306
                                                                                         - ~" TO ~" LINER TRANSITION REV IMTE              DESCRIPTION PROF'ESSIONAL ENGINEER:
OUAl.ITY lMl: SAFETY RELATED PROGRESS ENERGY                   ~~     rogress ENGINEERING SECTION                 ~J Ener Pl.ANT: ROBINSON PLANT - UNIT 2              SCAl.E: NONE TITLE:            EXAMINATION BOUNDARY OF CONTAINMENT LINER (DETAIL AT MOISTURE BARRIER)
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