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#REDIRECT [[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..]]
| number = ML15323A085
| issue date = 11/19/2015
| title = H.B. Robinson, Unit 2 - 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 In
| author name = Glover R M
| 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, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 172
}}
 
=Text=
{{#Wiki_filter:ENERGY'° 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 R. Michael Glover H. B. Robinson Steam Electric Plant Unit 2 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 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 O CFR 50, Appendix J, Integrated Leak Rate Test Interval and Type C Leak Rate Testing Frequency
 
==Dear Sir/Madam:==
In accordance with the provisions of 1 O 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 1 O 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. Attachment 5 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 1 O 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 ,, 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 testing except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.  --------NOTE------- SR 3.0.2 is not applicable ------------------------  However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criterion is < 0.6 La for the Type B and Type C tests. In accordance with 10 CFR 50, Appendix J, Option A, as modified by 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 and C leakage rate testing except for containment air lock testing, in accordance with 10 CFR 50, --------NOTE------- SR 3.0.2 is not applicable Enclosure                                                                                                                                                                Page 4 of 60  Appendix J, Option A, as modified by approved exemptions.  ------------------------  However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criterion is < 0.6 La for the Type B and Type C tests. the Containment Leakage Rate Testing Program. In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved 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 leakage rate testing, in accordance with the Containment Leakage Rate testing Program  In accordance with the 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 leakage rate testing, in accordance with the Containment Leakage Rate testing Program  In accordance with the 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 Attachment 1. 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, Option A, as modified by approved exemptions. -------------------------------------------------------------------     
--------NOTE------- SR 3.0.2 is not applicable ------------------------  Perform required air lock leakage rate testing in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. In accordance with 10 CFR 50, 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, Appendix J, Option A, as modified by approved exemptions. -------------------------------------------------------------------          --------NOTE------- SR 3.0.2 is not applicable ------------------------  Perform required air lock leakage rate testing in accordance with 10 CFR 50, Appendix J, Option A,  as modified by approved exemptions. the Containment Leakage Rate Testing Program. In accordance with 10 CFR 50, Appendix J, Option A, as 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. Attachment 4 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 licensee's 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, (ML 12138A041), 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  (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  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 2nd 5-Year Inspection 2005 3rd 5-Year Inspection 2010 4th 5-Year Inspection 2015 5th 5-Year Inspection 2020 6th 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 Request Number Affected Component(s) Examination Category Item No. Examination Area Alternative Examinations Status IWE/IWL RR-01 Containment Liner E-A E1.11 Insulated portions of the Containment Liner General/Detailed visual exam when panels are removed for maintenance purposes and 1/3 of the lowest elevation panels per Period, 100% for the Interval Granted TAC No.
(MD8509)  Dated 4/30/2009 IWE/IWL RR-02 Containment Moisture Barrier E-A E1.30 Containment moisture barrier General/Detailed visual exam when panels are removed for maintenance purposes and 1/3 of the lowest elevation panels per Period, 100% for the Interval Granted TAC No.
(MD8509)  Dated 4/30/2009  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 (% Weight per Day) Results Adjusted to Pa (% Weight per Day) May 1974 0.013 1 0.013 Feb. 1978 0.035 2 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 Fall 2005 RO-24 Spring 2007 RO-25 Fall 2008 RO-26 Summer 2010 RO-27 Spring 2012 RO-28 Fall 2013 RO-29 Spring 2015 As-Left 11985 sccm 18392 sccm 22972 sccm 16983 sccm 10436 sccm 9844 sccm 7122 sccm Fraction of 0.6 La 0.131 0.201 0.251 0.186 0.114 0.108 0.079  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 cycle's 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 Number / Component Initial LLRT SCCM Admin Limit SCCM As-left SCCM Cause of Failure Corrective Action Scheduled Interval 2013 RO-28 4  WD-1713 WD-1793 Primary System Vent Header  2800  2000  736  Note 1  Note 1  Appendix J Option A 42  V12-12 V12-13 Containment Vacuum Relief  3000  3000  53  Note 2  Note 2  Appendix J Option A 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 valve's 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 valve's 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 1970's 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 Fall 2005 RO-24 Spring 2007 RO-25 Fall 2008 RO-26 Summer 2010 RO-27 Spring 2012 RO-28 Fall 2013 RO-29 Summer 2015  0  sccm  1355.46  sccm  3002.01 sccm  2723.21sccm  2478.37 sccm  2054 sccm  1687 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 licensee should use the definition in the NEI TR 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002. (Refer to SE Section 3.1.1.1.)  HBRSEP2 will utilize the definition in NEI 94-01 Revision 2-A, Section 5.0. Reference Table 3.2.3-1, ILRT Test Results, Notes (3) and (4) The licensee submits a schedule of containment inspections to be performed prior to and between Type A tests. (Refer to SE Section 3.1.1.3.)  Reference Tables 3.2.2-1, 3.2.2-2 and 3.5-1 of this submittal. The licensee addresses the areas of the containment structure potentially subjected to degradation. (Refer to SE Section 3.1.3.)  Reference Section 3.2.2, Inaccessible Class MC Areas, of this submittal. 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 inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4.) HBRSEP2 steam generator replacements were completed in 1984 using the equipment hatch. 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 than 15 years. If a licensee has to utilize the provision of Section 9.1 of NEI TR 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. (Refer to SE Section 3.1.1.2.) HBRSEP2 will follow the requirements of NEI 94-01 Revision 2-A, Section 9.1. In accordance with the requirements of 94-01 Revision 2-A, SER Section 3.1.1.2, HBRSEP2 will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.
Enclosure                                                                                                                                                                Page 44 of 60    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. For plants licensed under 10 CFR Part 52, applications requesting a permanent extension 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. Not applicable. HBRSEP2 was not licensed under 10 CFR Part 52.
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 cycle's 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 plant's 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 (From Section 4.2 of SE) HBRSEP2 Response  1.The licensee submits documentation indicating that the technical adequacy of their PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension. The technical  adequacy of the HBRSEP2 PRA is consistent with the requirements of Regulatory Guide 1.200 as is relevant to this ILRT interval extension, as detailed in Attachment 4 of this submittal,  "PRA Risk Assessment for Extending ILRT Interval to 15 Years," Attachment 1. 2. The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small, and consistent with the clarification provided in Section 3.2.4.5 of this SE. Specifically, a small increase in population dose should be defined as an increase 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 should be defined as a value marginally greater than that accepted in a previous one-time ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point. EPRI Report No. 1009325, Revision 2-A, incorporates these population dose and CCFP acceptance guidelines, and these guidelines have been used for the HBRSEP2 plant specific assessments. The increase in population dose is 0.020 person-rem/year.
The increase in CCFP is 0.829%. The increase proved to be below 1.5 percentage points and thus is considered to be small.
Enclosure                                                                                                                                                                Page 49 of 60         
 
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. 3. The methodology in EPRI Report No. 1009325, Revision 2, is acceptable except for the calculation of the increase in expected population dose (per year of reactor operation). In order to make the methodology acceptable, the average leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La  EPRI Report No. 1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has been used in the HBRSEP2 plant specific risk assessment. 4. A licensee amendment request (LAR) is required in instances where containment over-pressure is relied upon for emergency core cooling system (ECCS) performance. For HBRSEP2, containment over-pressure is NOT relied upon for emergency core cooling system (ECCS) performance.
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. Attachment 4 of this submittal,  "PRA Risk Assessment for Extending ILRT Interval to 15 Years," Attachment 1, 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 Owner's 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. Attachment 4 of this submittal, "PRA Risk Assessment for Extending ILRT Interval to 15 Years," Attachment 1, 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 -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) 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 - 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 Commission's 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 Attachment 1 to Serial:  RNP-RA/15-0090 5 Pages (including this cover page) 
 
PROPOSED TECHNICAL SPECIFICATION CHANGES Containment  3.6.1    HBRSEP Unit No. 2 3.6-2 Amendment No. 176 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY  SR  3.6.1.1 Perform required Type B and C leakage rate testing except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. The leakage rate acceptance criterion is  1.0 La. However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criterion is < 0.6 La for the Type B and Type C tests.  --------NOTE-------- SR 3.0.2 is not applicable ------------------------
In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions  SR  3.6.1.2 Verify containment structural integrity in accordance with the Containment Tendon Surveillance Program. In accordance with the Containment Tendon Surveillance Program  SR  3.6.1.3 Perform required visual examinations and Type A leakage rate testing, in accordance with the Containment Leakage Rate Testing Program. In accordance with the Containment Leakage Rate Testing Program Containment Air Lock  3.6.2    HBRSEP Unit No. 2 3.6-5 Amendment No. 176 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 door is locked closed.
Once per 31 days  C. Containment air lock inoperable for reasons other than Condition A or B. C.1I  Initiate action to evaluate overall containment leakage rate per LCO 3.6.1. AND C.2  Verify a door is closed in the air lock. AND  C.3  Restore air lock to OPERABLE status. Immediately 
 
1 hour 24 hours  D. Required Action and associated Completion Time not met. D.1  Be in MODE 3. AND  D.2  Be in MODE 5. 6 hours 36 hours Containment Air Lock  3.6.2    HBRSEP Unit No. 2 3.6-6 Amendment No. 176 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 of SR 3.6.1.1, in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.  ----------------------------------------------------------------------
Perform required air lock leakage rate testing in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. 
 
---------NOTE------- SR 3.0.2 is not applicable
------------------------
 
In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. SR  3.6.2.2 Verify only one door in the air lock can be  opened at a time. 24 months Programs and Manuals 5.5 5.5  Programs and Manuals (continued) HBRSEP 5.0-22 Amendment No. 219 5.5.16 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 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 theApril 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 firstunit startup following testing in accordance with this program, the leakagerate 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. 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:
U. S. Nuclear Regulatory Commission Attachment 2 to Serial:  RNP-RA/15-0090 5 Pages (including this cover page)
REVISED TECHNICAL SPECIFICATION PAGES Containment  3.6.1    HBRSEP Unit No. 2 3.6-2 Amendment No. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY  SR  3.6.1.1 Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program.
In accordance with the Containment Leakage Rate Testing Program  SR  3.6.1.2 Verify containment structural integrity in accordance with the Containment Tendon Surveillance Program. 
 
In accordance with the Containment Tendon Surveillance Program Containment Air Lock  3.6.2    HBRSEP Unit No. 2 3.6-5 Amendment No. 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 door is locked closed.
Once per 31 days  C. Containment air lock inoperable for reasons other than Condition A or B. C.1  Initiate action to evaluate overall containment leakage rate per LCO 3.6.1. AND  C.2  Verify a door is closed in the air lock. AND C.3  Restore air lock to OPERABLE status. Immediately 
 
1 hour 24 hours  D. Required Action and associated Completion Time not met. D.1  Be in MODE 3.
AND  D.2  Be in MODE 5. 6 hours 36 hours Containment Air Lock  3.6.2    HBRSEP Unit No. 2 3.6-6 Amendment No. 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 accordance with the Containment Leakage Rate Testing Program.     
 
In accordance with the Containment Leakage Rate Testing Program. SR  3.6.2.2 Verify only one door in the air lock can be  opened at a time. 24 months Programs and Manuals  5.5  5.5  Programs and Manuals  (continued)      HBRSEP 5.0-22 Amendment No. 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. La a. b. La 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:
U. S. Nuclear Regulatory Commission Attachment 3 to Serial:  RNP-RA/15-0090 11 Pages (including this cover page)
PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES SR Applicability B 3.0  BASES  (continued)  HBRSEP Unit No. 2 B 3.0-14 Revision No. 52  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."
SR Applicability B 3.0  BASES  (continued)  HBRSEP Unit No. 2 B 3.0-15 Revision No. 52  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, Containment  B 3.6.1 B 3.6  CONTAINMENT SYSTEMS
 
B 3.6.1  Containment 
 
BASES (continued)
HBRSEP Unit No. 2 B 3.6-1 Revision No. 0  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 following a Design Basis Accident (DBA). Additionally, this structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.
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.
The concrete reactor building is required for structural integrity of the containment under DBA conditions. The steel liner and its penetrations establish the leakage limiting boundary of the containment. Maintaining the containment OPERABLE limits the leakage of fission product radioactivity from the containment to the environment.
The isolation devices for the penetrations in the containment boundary 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";
Containment B 3.6.1 BASES (continued) HBRSEP Unit No. 2 B 3.6-2 Revision No. 34 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; andd.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 La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure (Pa) resulting from the LOCA. At HBRSEP, Unit 2, Pa is specified as the containment design pressure of 42 psi, which exceeds the calculated peak LOCA containment pressure. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on all containment leakage rate testing. La is assumed to be 0.1% per day in the safety analysis at Pa = 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.
Containment B 3.6.1 BASES  (continued) (continued) HBRSEP Unit No. 2 B 3.6-3 Revision No. 0 LCO Containment OPERABILITY is maintained by limiting leakage to  1.0 La, except prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test. At this time, the combined Type B and C leakage must be < 0.6 La and the overall Type A leakage must be < 0.75 La. 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 the acceptance criteria of Appendix J. 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.
Containment B 3.6.1 BASES (continued) HBRSEP Unit No. 2 B 3.6-4 Revision No. 0 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 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, leakage test is required to be < 0.6 La for combined Type B and C leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an  1.0 La. At  1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by Appendix J, 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. SR  3.6.1.2 This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program.
Containment  B 3.6.1 BASES
 
HBRSEP Unit No. 2 B 3.6-5 Revision No. 0  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 La for overall Type A leakage. At  1.0 La 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. 
: 2. UFSAR, Section 6.2.
Containment Air Lock  B 3.6.2 BASES
 
(continued)
HBRSEP Unit No. 2 B 3.6-7 Revision No.34 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.
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."
Containment Air Lock  B 3.6.2 BASES (continued)
HBRSEP Unit No. 2 B 3.6-11 Revision No. 0  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 leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by Appendix J (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 in determining the overall containment leakage rate.
Containment Air Lock  B 3.6.2 BASES
 
HBRSEP Unit No. 2 B 3.6-12 Revision No. 0  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. 
: 2. UFSAR, Paragraph 6.9.2.
U. S. Nuclear Regulatory Commission Attachment 4 to Serial:  RNP-RA/15-0090 80 Pages (including this cover page)
EVALUATION OF RISK SIGNIFICANCE OF PERMANENT ILRT EXTENSION   
 
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U. S. Nuclear Regulatory Commission Attachment 5 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 RNP-PM-007 Revision 5 Page 9 of 12    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-007Revision 5Page 1 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsE-C579800IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 240-YPANEL ADJACENT TO 240-X(L) & 240-Z(R)MC8010RR-01E-C584500IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 248-BBPANEL ADJACENT TO 248-AA(L) & 248-CC(R)MC8010RR-01E-C590600IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 256-SSPANEL ADJACENT TO 256-RR(L) & 256-TT(R)MC8010RR-01E-C563700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 264-GGPANEL ADJACENT TO 264-FF(L) & 264-HH(R)MC8010RR-01E-C597600IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 268-DDPANEL ADJACENT TO 268-CC(L) & 268-EE(R)MC8010RR-01E-C597700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 268-EEPANEL ADJACENT TO 268-DD(L) & 268-FF(R)MC8010RR-01E-C597900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 268-GGPANEL ADJACENT TO 268-FF(L) & 268-HH(R)MC8010RR-01E-C601000IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 272-FFPANEL ADJACENT TO 272-EE(L) & 272-GG(R)MC8010RR-01E-C601100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 272-GGPANEL ADJACENT TO 272-FF(L) & 272-HH(R)MC8010RR-01E-C602300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 272-TTPANEL ADJACENT TO 272-SS(L) & 272-UU(R)MC8010RR-01E-C569800IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-APANEL ADJACENT TO 280-CCC(L) & 280-B(R)MC8010RR-01E-C570200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-EPANEL ADJACENT TO 280-D(L) & 280-F(R)MC8010RR-01CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128PeriodH. B. ROBINSON UNIT 2PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDEVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILED RNP-PM-007Revision 5Page 2 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsPeriodH. B. ROBINSON UNIT 2E-C570400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-GPANEL ADJACENT TO 280-F(L) & 280-H(R)MC8010RR-01E-C570700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-LPANEL ADJACENT TO 280-K(L) & 280-M1(R)MC8010RR-01E-C570800IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-M1PANEL ADJACENT TO 280-L(L) & 280-M2(R)MC8010RR-01E-C570900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-M2PANEL ADJACENT TO 280-M1(L) & 280-M3(R)MC8010RR-01E-C631200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-PPANEL ADJACENT TO 280-O(L) & 280-Q(R)MC8010RR-01E-C633400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-LLPANEL ADJACENT TO 280-KK(L) & 280-MM(R)MC8010RR-01E-C633500IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-MMPANEL ADJACENT TO 280-LL(L) & 280-NN(R)MC8010RR-01E-C572800IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-SSPANEL ADJACENT TO 280-RR(L) & 280-TT(R)MC8010RR-01E-C572900IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-TTPANEL ADJACENT TO 280-SS(L) & 280-UU(R)MC8010RR-01E-C573000IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-UUPANEL ADJACENT TO 280-TT(L) & 280-VV(R)MC8010RR-01E-C573100IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-VVPANEL ADJACENT TO 280-UU(L) & 280-WW(R)MC8010RR-01E-C573400IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 280-AAAPANEL ADJACENT TO 280-ZZ(L) & 280-BBB(R)MC8010RR-01GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDBULGES AND PITTING EVALUATED UNDER EC 72699, RNP CALCULATION RNP-C/STRU-1128 AND RNP-C/STRU-1130PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128BULGES AND PITTING EVALUATED UNDER EC 72699, RNP CALCULATION RNP-C/STRU-1128 AND RNP-C/STRU-1130CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDGENERAL - DETAILEDEVALUATED IN RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDPITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILED RNP-PM-007Revision 5Page 3 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsPeriodH. B. ROBINSON UNIT 2E-C573500IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-BBBPANEL ADJACENT TO 280-AAA(L) & 280-CCC(R)MC8010RR-01E-C571200IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 284-BPANEL ADJACENT TO 284-A(L) & 284-C(R)MC8010RR-01E-C571800IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 284-HPANEL ADJACENT TO 284-G(L) & 284-I(R)MC8010RR-01E-C571900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 284-IPANEL ADJACENT TO 284-H(L) & 284-J(R)MC8010RR-01E-C575300IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 284-UUPANEL ADJACENT TO 284-TT(L) & 284-VV(R)MC8010RR-01E-C637000IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 284-VVPANEL ADJACENT TO 284-UU(L) & 284-WW(R)MC8010RR-01E-C575500IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 284-BBBPANEL ADJACENT TO 284-AAA(L) & 284-CCC(R)MC8010RR-01E-C573700IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 284-CCCPANEL ADJACENT TO 284-BBB(L) & 284-A(R)MC8010RR-01E-C638100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-GPANEL ADJACENT TO 288-F(L) & 288-H(R)MC8010RR-01E-C638300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-IPANEL ADJACENT TO 288-H(L) & 288-J(R)MC8010RR-01E-C639300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-PPANEL ADJACENT TO 288-O(L) & 288-Q(R)MC8010RR-01E-C642300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-TTPANEL ADJACENT TO 288-SS(L) & 288-UU(R)MC8010RR-01GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 RNP-PM-007Revision 5Page 4 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsPeriodH. B. ROBINSON UNIT 2E-C642400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-UUPANEL ADJACENT TO 288-TT(L) & 288-VV(R)MC8010RR-01E-C642500IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-VVPANEL ADJACENT TO 288-UU(L) & 288-WW(R)MC8010RR-01E-C644100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 292-IPANEL ADJACENT TO 292-H(L) & 292-J(R)MC8010RR-01E-C648000IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 292-UUPANEL ADJACENT TO 292-TT(L) & 292-VV(R)MC8010RR-01E-C648100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 292-VVPANEL ADJACENT TO 292-UU(L) & 292-WW(R)MC8010RR-01E-C648200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 292-WWPANEL ADJACENT TO 292-VV(L) & 292-XX(R)MC8010RR-01E-C649500IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 296-GPANEL ADJACENT TO 296-F(L) & 296-H(R)MC8010RR-01E-C649600IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 296-HPANEL ADJACENT TO 296-G(L) & 296-I(R)MC8010RR-01E-C649700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 296-IPANEL ADJACENT TO 296-H(L) & 296-J(R)MC8010RR-01E-C650700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 296-PPANEL ADJACENT TO 296-O(L) & 296-Q(R)MC8010RR-01E-C655400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 300-HPANEL ADJACENT TO 300-G(L) & 300-I(R)MC8010RR-01E-C659700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 300-VVPANEL ADJACENT TO 300-UU(L) & 300-WW(R)MC8010RR-01GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILED RNP-PM-007Revision 5Page 5 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsPeriodH. B. ROBINSON UNIT 2E-C661300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 304-IPANEL ADJACENT TO 304-H(L) & 304-J(R)MC8010RR-01E-C665200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 304-TTPANEL ADJACENT TO 304-SS(L) & 304-UU(R)MC8010RR-01E-C655300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 304-UUPANEL ADJACENT TO 304-TT(L) & 304-VV(R)MC8010RR-01E-C655400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 304-VVPANEL ADJACENT TO 304-UU(L) & 304-WW(R)MC8010RR-01E-C666900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-HPANEL ADJACENT TO 308-G(L) & 308-I(R)MC8010RR-01E-C668200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-RPANEL ADJACENT TO 308-Q(L) & 308-S(R)MC8010RR-01E-C668900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-YPANEL ADJACENT TO 308-X(L) & 308-Z(R)MC8010RR-01E-C671000IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-TTPANEL ADJACENT TO 308-SS(L) & 308-UU(R)MC8010RR-01E-C575600IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-UUPANEL ADJACENT TO 308-TT(L) & 308-VV(R)MC8010RR-01E-C671100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-VVPANEL ADJACENT TO 308-UU(L) & 308-WW(R)MC8010RR-01E-C671200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-WWPANEL ADJACENT TO 308-VV(L) & 308-XX(R)MC8010RR-01GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 AND BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 AND BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILED U. S. Nuclear Regulatory Commission Attachment 6 to Serial:  RNP-RA/15-0090 2 Pages (including this cover page)
CONTAINMENT LINER DETAIL AT MOISTURE BARRIER   
 
F'llE: GROUTED TENDON (HIGH STRENGTH STEEL BARS INSIDE s* DIAMETER PIPE) CONTAINMENT ---CONCRETE WALL EL 226'-0" ---..... BASE MAT ---.. J:,. . ' v ; . . : . ** *A t .. , ** . I A ' .,, . ' -. , 4-. . ' I> .. CONTAINMENT LINER (NOTE 1) rcc:MB) X MOISTURE BARRIER IN CONCRETE NOTCH fl ---FLOOR SLAB . }) TOP OF MAT LINER TO LINER TRANSITION LEGEND INACCESSIBLE FOR VISUAL INSPECTION IWE BOUNDARY FOR INFORMATION ( ) ISi COMPONENT IDENTIFIER NOTES: 1) THE CONTAINMENT LINER IS THICK FROM EL. 226'-0" TO EL. 253'-0". THICK FROM EL. 253'-0" TO 352'-0" ANO THICK ABOVE 352'-0".
 
==REFERENCES:==
: 1) G-190343 2) G-190353 3) G-190358 4) G-190359 5) RELIEF REQUEST IWE/IWL-02 FIGURE 1 0 ,, EC 65306 REV IMTE DESCRIPTION PROF'ESSIONAL ENGINEER: OUAl.ITY lMl: SAFETY RELATED PROGRESS ENERGY rogress ENGINEERING SECTION Ener Pl.ANT: ROBINSON PLANT -UNIT 2 SCAl.E: NONE TITLE: EXAMINATION BOUNDARY OF CONTAINMENT LINER (DETAIL AT MOISTURE BARRIER) FIGURE 16 ENERGY'&deg; 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 R. Michael Glover H. B. Robinson Steam Electric Plant Unit 2 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 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 O CFR 50, Appendix J, Integrated Leak Rate Test Interval and Type C Leak Rate Testing Frequency
 
==Dear Sir/Madam:==
In accordance with the provisions of 1 O 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 1 O 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. Attachment 5 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 1 O 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 ,, 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 testing except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.  --------NOTE------- SR 3.0.2 is not applicable ------------------------  However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criterion is < 0.6 La for the Type B and Type C tests. In accordance with 10 CFR 50, Appendix J, Option A, as modified by 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 and C leakage rate testing except for containment air lock testing, in accordance with 10 CFR 50, --------NOTE------- SR 3.0.2 is not applicable Enclosure                                                                                                                                                                Page 4 of 60  Appendix J, Option A, as modified by approved exemptions.  ------------------------  However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criterion is < 0.6 La for the Type B and Type C tests. the Containment Leakage Rate Testing Program. In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved 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 leakage rate testing, in accordance with the Containment Leakage Rate testing Program  In accordance with the 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 leakage rate testing, in accordance with the Containment Leakage Rate testing Program  In accordance with the 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 Attachment 1. 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, Option A, as modified by approved exemptions. -------------------------------------------------------------------     
--------NOTE------- SR 3.0.2 is not applicable ------------------------  Perform required air lock leakage rate testing in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. In accordance with 10 CFR 50, 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, Appendix J, Option A, as modified by approved exemptions. -------------------------------------------------------------------          --------NOTE------- SR 3.0.2 is not applicable ------------------------  Perform required air lock leakage rate testing in accordance with 10 CFR 50, Appendix J, Option A,  as modified by approved exemptions. the Containment Leakage Rate Testing Program. In accordance with 10 CFR 50, Appendix J, Option A, as 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. Attachment 4 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 licensee's 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, (ML 12138A041), 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  (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  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 2nd 5-Year Inspection 2005 3rd 5-Year Inspection 2010 4th 5-Year Inspection 2015 5th 5-Year Inspection 2020 6th 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 Request Number Affected Component(s) Examination Category Item No. Examination Area Alternative Examinations Status IWE/IWL RR-01 Containment Liner E-A E1.11 Insulated portions of the Containment Liner General/Detailed visual exam when panels are removed for maintenance purposes and 1/3 of the lowest elevation panels per Period, 100% for the Interval Granted TAC No.
(MD8509)  Dated 4/30/2009 IWE/IWL RR-02 Containment Moisture Barrier E-A E1.30 Containment moisture barrier General/Detailed visual exam when panels are removed for maintenance purposes and 1/3 of the lowest elevation panels per Period, 100% for the Interval Granted TAC No.
(MD8509)  Dated 4/30/2009  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 (% Weight per Day) Results Adjusted to Pa (% Weight per Day) May 1974 0.013 1 0.013 Feb. 1978 0.035 2 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 Fall 2005 RO-24 Spring 2007 RO-25 Fall 2008 RO-26 Summer 2010 RO-27 Spring 2012 RO-28 Fall 2013 RO-29 Spring 2015 As-Left 11985 sccm 18392 sccm 22972 sccm 16983 sccm 10436 sccm 9844 sccm 7122 sccm Fraction of 0.6 La 0.131 0.201 0.251 0.186 0.114 0.108 0.079  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 cycle's 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 Number / Component Initial LLRT SCCM Admin Limit SCCM As-left SCCM Cause of Failure Corrective Action Scheduled Interval 2013 RO-28 4  WD-1713 WD-1793 Primary System Vent Header  2800  2000  736  Note 1  Note 1  Appendix J Option A 42  V12-12 V12-13 Containment Vacuum Relief  3000  3000  53  Note 2  Note 2  Appendix J Option A 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 valve's 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 valve's 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 1970's 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 Fall 2005 RO-24 Spring 2007 RO-25 Fall 2008 RO-26 Summer 2010 RO-27 Spring 2012 RO-28 Fall 2013 RO-29 Summer 2015  0  sccm  1355.46  sccm  3002.01 sccm  2723.21sccm  2478.37 sccm  2054 sccm  1687 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 licensee should use the definition in the NEI TR 94-01, Revision 2, in lieu of that in ANSI/ANS-56.8-2002. (Refer to SE Section 3.1.1.1.)  HBRSEP2 will utilize the definition in NEI 94-01 Revision 2-A, Section 5.0. Reference Table 3.2.3-1, ILRT Test Results, Notes (3) and (4) The licensee submits a schedule of containment inspections to be performed prior to and between Type A tests. (Refer to SE Section 3.1.1.3.)  Reference Tables 3.2.2-1, 3.2.2-2 and 3.5-1 of this submittal. The licensee addresses the areas of the containment structure potentially subjected to degradation. (Refer to SE Section 3.1.3.)  Reference Section 3.2.2, Inaccessible Class MC Areas, of this submittal. 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 inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4.) HBRSEP2 steam generator replacements were completed in 1984 using the equipment hatch. 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 than 15 years. If a licensee has to utilize the provision of Section 9.1 of NEI TR 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. (Refer to SE Section 3.1.1.2.) HBRSEP2 will follow the requirements of NEI 94-01 Revision 2-A, Section 9.1. In accordance with the requirements of 94-01 Revision 2-A, SER Section 3.1.1.2, HBRSEP2 will also demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.
Enclosure                                                                                                                                                                Page 44 of 60    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. For plants licensed under 10 CFR Part 52, applications requesting a permanent extension 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. Not applicable. HBRSEP2 was not licensed under 10 CFR Part 52.
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 cycle's 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 plant's 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 (From Section 4.2 of SE) HBRSEP2 Response  1.The licensee submits documentation indicating that the technical adequacy of their PRA is consistent with the requirements of RG 1.200 relevant to the ILRT extension. The technical  adequacy of the HBRSEP2 PRA is consistent with the requirements of Regulatory Guide 1.200 as is relevant to this ILRT interval extension, as detailed in Attachment 4 of this submittal,  "PRA Risk Assessment for Extending ILRT Interval to 15 Years," Attachment 1. 2. The licensee submits documentation indicating that the estimated risk increase associated with permanently extending the ILRT surveillance interval to 15 years is small, and consistent with the clarification provided in Section 3.2.4.5 of this SE. Specifically, a small increase in population dose should be defined as an increase 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 should be defined as a value marginally greater than that accepted in a previous one-time ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1.5 percentage point. EPRI Report No. 1009325, Revision 2-A, incorporates these population dose and CCFP acceptance guidelines, and these guidelines have been used for the HBRSEP2 plant specific assessments. The increase in population dose is 0.020 person-rem/year.
The increase in CCFP is 0.829%. The increase proved to be below 1.5 percentage points and thus is considered to be small.
Enclosure                                                                                                                                                                Page 49 of 60         
 
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. 3. The methodology in EPRI Report No. 1009325, Revision 2, is acceptable except for the calculation of the increase in expected population dose (per year of reactor operation). In order to make the methodology acceptable, the average leak rate accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La  EPRI Report No. 1009325, Revision 2-A, incorporated the use of 100 La as the average leak rate for the pre-existing containment large leakage rate accident case (accident case 3b), and this value has been used in the HBRSEP2 plant specific risk assessment. 4. A licensee amendment request (LAR) is required in instances where containment over-pressure is relied upon for emergency core cooling system (ECCS) performance. For HBRSEP2, containment over-pressure is NOT relied upon for emergency core cooling system (ECCS) performance.
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. Attachment 4 of this submittal,  "PRA Risk Assessment for Extending ILRT Interval to 15 Years," Attachment 1, 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 Owner's 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. Attachment 4 of this submittal, "PRA Risk Assessment for Extending ILRT Interval to 15 Years," Attachment 1, 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 -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) 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 - 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 Commission's 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 Attachment 1 to Serial:  RNP-RA/15-0090 5 Pages (including this cover page) 
 
PROPOSED TECHNICAL SPECIFICATION CHANGES Containment  3.6.1    HBRSEP Unit No. 2 3.6-2 Amendment No. 176 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY  SR  3.6.1.1 Perform required Type B and C leakage rate testing except for containment air lock testing, in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. The leakage rate acceptance criterion is  1.0 La. However, during the first unit startup following testing performed in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions, the leakage rate acceptance criterion is < 0.6 La for the Type B and Type C tests.  --------NOTE-------- SR 3.0.2 is not applicable ------------------------
In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions  SR  3.6.1.2 Verify containment structural integrity in accordance with the Containment Tendon Surveillance Program. In accordance with the Containment Tendon Surveillance Program  SR  3.6.1.3 Perform required visual examinations and Type A leakage rate testing, in accordance with the Containment Leakage Rate Testing Program. In accordance with the Containment Leakage Rate Testing Program Containment Air Lock  3.6.2    HBRSEP Unit No. 2 3.6-5 Amendment No. 176 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 door is locked closed.
Once per 31 days  C. Containment air lock inoperable for reasons other than Condition A or B. C.1I  Initiate action to evaluate overall containment leakage rate per LCO 3.6.1. AND C.2  Verify a door is closed in the air lock. AND  C.3  Restore air lock to OPERABLE status. Immediately 
 
1 hour 24 hours  D. Required Action and associated Completion Time not met. D.1  Be in MODE 3. AND  D.2  Be in MODE 5. 6 hours 36 hours Containment Air Lock  3.6.2    HBRSEP Unit No. 2 3.6-6 Amendment No. 176 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 of SR 3.6.1.1, in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions.  ----------------------------------------------------------------------
Perform required air lock leakage rate testing in accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. 
 
---------NOTE------- SR 3.0.2 is not applicable
------------------------
 
In accordance with 10 CFR 50, Appendix J, Option A, as modified by approved exemptions. SR  3.6.2.2 Verify only one door in the air lock can be  opened at a time. 24 months Programs and Manuals 5.5 5.5  Programs and Manuals (continued) HBRSEP 5.0-22 Amendment No. 219 5.5.16 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 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 theApril 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 firstunit startup following testing in accordance with this program, the leakagerate 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. 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:
U. S. Nuclear Regulatory Commission Attachment 2 to Serial:  RNP-RA/15-0090 5 Pages (including this cover page)
REVISED TECHNICAL SPECIFICATION PAGES Containment  3.6.1    HBRSEP Unit No. 2 3.6-2 Amendment No. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY  SR  3.6.1.1 Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program.
In accordance with the Containment Leakage Rate Testing Program  SR  3.6.1.2 Verify containment structural integrity in accordance with the Containment Tendon Surveillance Program. 
 
In accordance with the Containment Tendon Surveillance Program Containment Air Lock  3.6.2    HBRSEP Unit No. 2 3.6-5 Amendment No. 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 door is locked closed.
Once per 31 days  C. Containment air lock inoperable for reasons other than Condition A or B. C.1  Initiate action to evaluate overall containment leakage rate per LCO 3.6.1. AND  C.2  Verify a door is closed in the air lock. AND C.3  Restore air lock to OPERABLE status. Immediately 
 
1 hour 24 hours  D. Required Action and associated Completion Time not met. D.1  Be in MODE 3.
AND  D.2  Be in MODE 5. 6 hours 36 hours Containment Air Lock  3.6.2    HBRSEP Unit No. 2 3.6-6 Amendment No. 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 accordance with the Containment Leakage Rate Testing Program.     
 
In accordance with the Containment Leakage Rate Testing Program. SR  3.6.2.2 Verify only one door in the air lock can be  opened at a time. 24 months Programs and Manuals  5.5  5.5  Programs and Manuals  (continued)      HBRSEP 5.0-22 Amendment No. 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. La a. b. La 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:
U. S. Nuclear Regulatory Commission Attachment 3 to Serial:  RNP-RA/15-0090 11 Pages (including this cover page)
PROPOSED TECHNICAL SPECIFICATIONS BASES CHANGES SR Applicability B 3.0  BASES  (continued)  HBRSEP Unit No. 2 B 3.0-14 Revision No. 52  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."
SR Applicability B 3.0  BASES  (continued)  HBRSEP Unit No. 2 B 3.0-15 Revision No. 52  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, Containment  B 3.6.1 B 3.6  CONTAINMENT SYSTEMS
 
B 3.6.1  Containment 
 
BASES (continued)
HBRSEP Unit No. 2 B 3.6-1 Revision No. 0  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 following a Design Basis Accident (DBA). Additionally, this structure provides shielding from the fission products that may be present in the containment atmosphere following accident conditions.
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.
The concrete reactor building is required for structural integrity of the containment under DBA conditions. The steel liner and its penetrations establish the leakage limiting boundary of the containment. Maintaining the containment OPERABLE limits the leakage of fission product radioactivity from the containment to the environment.
The isolation devices for the penetrations in the containment boundary 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";
Containment B 3.6.1 BASES (continued) HBRSEP Unit No. 2 B 3.6-2 Revision No. 34 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; andd.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 La: the maximum allowable containment leakage rate at the calculated peak containment internal pressure (Pa) resulting from the LOCA. At HBRSEP, Unit 2, Pa is specified as the containment design pressure of 42 psi, which exceeds the calculated peak LOCA containment pressure. The allowable leakage rate represented by La forms the basis for the acceptance criteria imposed on all containment leakage rate testing. La is assumed to be 0.1% per day in the safety analysis at Pa = 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.
Containment B 3.6.1 BASES  (continued) (continued) HBRSEP Unit No. 2 B 3.6-3 Revision No. 0 LCO Containment OPERABILITY is maintained by limiting leakage to  1.0 La, except prior to the first startup after performing a required 10 CFR 50, Appendix J, leakage test. At this time, the combined Type B and C leakage must be < 0.6 La and the overall Type A leakage must be < 0.75 La. 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 the acceptance criteria of Appendix J. 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.
Containment B 3.6.1 BASES (continued) HBRSEP Unit No. 2 B 3.6-4 Revision No. 0 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 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, leakage test is required to be < 0.6 La for combined Type B and C leakage. At all other times between required leakage rate tests, the acceptance criteria is based on an  1.0 La. At  1.0 La the offsite dose consequences are bounded by the assumptions of the safety analysis. SR Frequencies are as required by Appendix J, 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. SR  3.6.1.2 This SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program.
Containment  B 3.6.1 BASES
 
HBRSEP Unit No. 2 B 3.6-5 Revision No. 0  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 La for overall Type A leakage. At  1.0 La 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. 
: 2. UFSAR, Section 6.2.
Containment Air Lock  B 3.6.2 BASES
 
(continued)
HBRSEP Unit No. 2 B 3.6-7 Revision No.34 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.
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."
Containment Air Lock  B 3.6.2 BASES (continued)
HBRSEP Unit No. 2 B 3.6-11 Revision No. 0  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 leakage rate testing requirements with regard to air lock leakage (Type B leakage tests). The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by Appendix J (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 in determining the overall containment leakage rate.
Containment Air Lock  B 3.6.2 BASES
 
HBRSEP Unit No. 2 B 3.6-12 Revision No. 0  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. 
: 2. UFSAR, Paragraph 6.9.2.
U. S. Nuclear Regulatory Commission Attachment 4 to Serial:  RNP-RA/15-0090 80 Pages (including this cover page)
EVALUATION OF RISK SIGNIFICANCE OF PERMANENT ILRT EXTENSION   
 
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U. S. Nuclear Regulatory Commission Attachment 5 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 RNP-PM-007 Revision 5 Page 9 of 12    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-007Revision 5Page 1 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsE-C579800IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 240-YPANEL ADJACENT TO 240-X(L) & 240-Z(R)MC8010RR-01E-C584500IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 248-BBPANEL ADJACENT TO 248-AA(L) & 248-CC(R)MC8010RR-01E-C590600IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 256-SSPANEL ADJACENT TO 256-RR(L) & 256-TT(R)MC8010RR-01E-C563700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 264-GGPANEL ADJACENT TO 264-FF(L) & 264-HH(R)MC8010RR-01E-C597600IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 268-DDPANEL ADJACENT TO 268-CC(L) & 268-EE(R)MC8010RR-01E-C597700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 268-EEPANEL ADJACENT TO 268-DD(L) & 268-FF(R)MC8010RR-01E-C597900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 268-GGPANEL ADJACENT TO 268-FF(L) & 268-HH(R)MC8010RR-01E-C601000IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 272-FFPANEL ADJACENT TO 272-EE(L) & 272-GG(R)MC8010RR-01E-C601100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 272-GGPANEL ADJACENT TO 272-FF(L) & 272-HH(R)MC8010RR-01E-C602300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 272-TTPANEL ADJACENT TO 272-SS(L) & 272-UU(R)MC8010RR-01E-C569800IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-APANEL ADJACENT TO 280-CCC(L) & 280-B(R)MC8010RR-01E-C570200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-EPANEL ADJACENT TO 280-D(L) & 280-F(R)MC8010RR-01CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128PeriodH. B. ROBINSON UNIT 2PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDEVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILED RNP-PM-007Revision 5Page 2 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsPeriodH. B. ROBINSON UNIT 2E-C570400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-GPANEL ADJACENT TO 280-F(L) & 280-H(R)MC8010RR-01E-C570700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-LPANEL ADJACENT TO 280-K(L) & 280-M1(R)MC8010RR-01E-C570800IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-M1PANEL ADJACENT TO 280-L(L) & 280-M2(R)MC8010RR-01E-C570900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-M2PANEL ADJACENT TO 280-M1(L) & 280-M3(R)MC8010RR-01E-C631200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-PPANEL ADJACENT TO 280-O(L) & 280-Q(R)MC8010RR-01E-C633400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-LLPANEL ADJACENT TO 280-KK(L) & 280-MM(R)MC8010RR-01E-C633500IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 280-MMPANEL ADJACENT TO 280-LL(L) & 280-NN(R)MC8010RR-01E-C572800IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-SSPANEL ADJACENT TO 280-RR(L) & 280-TT(R)MC8010RR-01E-C572900IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-TTPANEL ADJACENT TO 280-SS(L) & 280-UU(R)MC8010RR-01E-C573000IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-UUPANEL ADJACENT TO 280-TT(L) & 280-VV(R)MC8010RR-01E-C573100IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-VVPANEL ADJACENT TO 280-UU(L) & 280-WW(R)MC8010RR-01E-C573400IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 280-AAAPANEL ADJACENT TO 280-ZZ(L) & 280-BBB(R)MC8010RR-01GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDBULGES AND PITTING EVALUATED UNDER EC 72699, RNP CALCULATION RNP-C/STRU-1128 AND RNP-C/STRU-1130PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128BULGES AND PITTING EVALUATED UNDER EC 72699, RNP CALCULATION RNP-C/STRU-1128 AND RNP-C/STRU-1130CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDGENERAL - DETAILEDEVALUATED IN RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDPITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILED RNP-PM-007Revision 5Page 3 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsPeriodH. B. ROBINSON UNIT 2E-C573500IWEHBR2-10618 SHT 195, FIGURE 14cHE4.11L-WALL 280-BBBPANEL ADJACENT TO 280-AAA(L) & 280-CCC(R)MC8010RR-01E-C571200IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 284-BPANEL ADJACENT TO 284-A(L) & 284-C(R)MC8010RR-01E-C571800IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 284-HPANEL ADJACENT TO 284-G(L) & 284-I(R)MC8010RR-01E-C571900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 284-IPANEL ADJACENT TO 284-H(L) & 284-J(R)MC8010RR-01E-C575300IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 284-UUPANEL ADJACENT TO 284-TT(L) & 284-VV(R)MC8010RR-01E-C637000IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 284-VVPANEL ADJACENT TO 284-UU(L) & 284-WW(R)MC8010RR-01E-C575500IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 284-BBBPANEL ADJACENT TO 284-AAA(L) & 284-CCC(R)MC8010RR-01E-C573700IWEHBR2-10618 SHT 195, FIGURE 14chE4.11L-WALL 284-CCCPANEL ADJACENT TO 284-BBB(L) & 284-A(R)MC8010RR-01E-C638100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-GPANEL ADJACENT TO 288-F(L) & 288-H(R)MC8010RR-01E-C638300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-IPANEL ADJACENT TO 288-H(L) & 288-J(R)MC8010RR-01E-C639300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-PPANEL ADJACENT TO 288-O(L) & 288-Q(R)MC8010RR-01E-C642300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-TTPANEL ADJACENT TO 288-SS(L) & 288-UU(R)MC8010RR-01GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128PITTING EVALUATED UNDER EC 72699 AND RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 RNP-PM-007Revision 5Page 4 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsPeriodH. B. ROBINSON UNIT 2E-C642400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-UUPANEL ADJACENT TO 288-TT(L) & 288-VV(R)MC8010RR-01E-C642500IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 288-VVPANEL ADJACENT TO 288-UU(L) & 288-WW(R)MC8010RR-01E-C644100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 292-IPANEL ADJACENT TO 292-H(L) & 292-J(R)MC8010RR-01E-C648000IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 292-UUPANEL ADJACENT TO 292-TT(L) & 292-VV(R)MC8010RR-01E-C648100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 292-VVPANEL ADJACENT TO 292-UU(L) & 292-WW(R)MC8010RR-01E-C648200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 292-WWPANEL ADJACENT TO 292-VV(L) & 292-XX(R)MC8010RR-01E-C649500IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 296-GPANEL ADJACENT TO 296-F(L) & 296-H(R)MC8010RR-01E-C649600IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 296-HPANEL ADJACENT TO 296-G(L) & 296-I(R)MC8010RR-01E-C649700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 296-IPANEL ADJACENT TO 296-H(L) & 296-J(R)MC8010RR-01E-C650700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 296-PPANEL ADJACENT TO 296-O(L) & 296-Q(R)MC8010RR-01E-C655400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 300-HPANEL ADJACENT TO 300-G(L) & 300-I(R)MC8010RR-01E-C659700IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 300-VVPANEL ADJACENT TO 300-UU(L) & 300-WW(R)MC8010RR-01GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILED RNP-PM-007Revision 5Page 5 of 5CATEGORY - E-C AUGMENTSECOND TEN-YEAR INTERVAL INSERVICE IWE/IWL INSPECTION PLANCategorySummaryScopeDrawingItem NoComp IDMethodComponent DescriptionClassSystemCode Case/Relief Request 123CommentsPeriodH. B. ROBINSON UNIT 2E-C661300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 304-IPANEL ADJACENT TO 304-H(L) & 304-J(R)MC8010RR-01E-C665200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 304-TTPANEL ADJACENT TO 304-SS(L) & 304-UU(R)MC8010RR-01E-C655300IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 304-UUPANEL ADJACENT TO 304-TT(L) & 304-VV(R)MC8010RR-01E-C655400IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 304-VVPANEL ADJACENT TO 304-UU(L) & 304-WW(R)MC8010RR-01E-C666900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-HPANEL ADJACENT TO 308-G(L) & 308-I(R)MC8010RR-01E-C668200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-RPANEL ADJACENT TO 308-Q(L) & 308-S(R)MC8010RR-01E-C668900IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-YPANEL ADJACENT TO 308-X(L) & 308-Z(R)MC8010RR-01E-C671000IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-TTPANEL ADJACENT TO 308-SS(L) & 308-UU(R)MC8010RR-01E-C575600IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-UUPANEL ADJACENT TO 308-TT(L) & 308-VV(R)MC8010RR-01E-C671100IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-VVPANEL ADJACENT TO 308-UU(L) & 308-WW(R)MC8010RR-01E-C671200IWEHBR2-10618 SHT 195, FIGURE 14 chE4.11L-WALL 308-WWPANEL ADJACENT TO 308-VV(L) & 308-XX(R)MC8010RR-01GENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDGENERAL - DETAILEDCORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 AND BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128 AND BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130BULGE EVALUATED IN RNP CALCULATION RNP-C/STRU-1130CORROSION EVALUATED IN RNP CALCULATION RNP-C/STRU-1128GENERAL - DETAILED U. S. Nuclear Regulatory Commission Attachment 6 to Serial:  RNP-RA/15-0090 2 Pages (including this cover page)
CONTAINMENT LINER DETAIL AT MOISTURE BARRIER   
 
F'llE: GROUTED TENDON (HIGH STRENGTH STEEL BARS INSIDE s* DIAMETER PIPE) CONTAINMENT ---CONCRETE WALL EL 226'-0" ---..... BASE MAT ---.. J:,. . ' v ; . . : . ** *A t .. , ** . I A ' .,, . ' -. , 4-. . ' I> .. CONTAINMENT LINER (NOTE 1) rcc:MB) X MOISTURE BARRIER IN CONCRETE NOTCH fl ---FLOOR SLAB . }) TOP OF MAT LINER TO LINER TRANSITION LEGEND INACCESSIBLE FOR VISUAL INSPECTION IWE BOUNDARY FOR INFORMATION ( ) ISi COMPONENT IDENTIFIER NOTES: 1) THE CONTAINMENT LINER IS THICK FROM EL. 226'-0" TO EL. 253'-0". THICK FROM EL. 253'-0" TO 352'-0" ANO THICK ABOVE 352'-0".
 
==REFERENCES:==
: 1) G-190343 2) G-190353 3) G-190358 4) G-190359 5) RELIEF REQUEST IWE/IWL-02 FIGURE 1 0 ,, EC 65306 REV IMTE DESCRIPTION PROF'ESSIONAL ENGINEER: OUAl.ITY lMl: SAFETY RELATED PROGRESS ENERGY rogress ENGINEERING SECTION Ener Pl.ANT: ROBINSON PLANT -UNIT 2 SCAl.E: NONE TITLE: EXAMINATION BOUNDARY OF CONTAINMENT LINER (DETAIL AT MOISTURE BARRIER) FIGURE 16}}

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