GNRO-2016/00062, License Amendment Request (LAR) for One Cycle Extension of Appendix J Type a Integrated Leakage Rate Test and Drywell Bypass

License Amendment Request (LAR) for One Cycle Extension of Appendix J Type a Integrated Leakage Rate Test and Drywell Bypass

ML16364A338
Person / Time
Site:	Grand Gulf
Issue date:	12/29/2016
From:	Fallacara V Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF9461, GNRO-2016/00062
Download: ML16364A338 (86)

Text

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'b?SEntergy GNRO-2016/00062 December 29,2016 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Entergy Operations, Inc.

P. O. Box 756 Port Gibson, MS 39150 Vincent Fallacara Site Vice President Tel. (601) 437-2129

Dear Sir or Madam:

License Amendment Request(LAR) for One Cycle Extension Of AppendixJ Type A Integrated Leakage Rate Test and Drywell Bypass Test Interval Docket No. 50-416 License No. NPF-29 Pursuant to 10 CFR 50.90, Entergy Operation Inc. (Entergy) is submitting a request for an amendment to the Technical Specifications (TS) for the Grand Gulf Nuclear Station Unit 1 (GGNS).

The proposed LAR would allow for a one cycle extension to the 10-year frequency of the GGNS Unit 1 containment leakage rate test (i.e., Integrated Leakage Rate Test (ILRT) or Type A test) and the drywell bypass leak rate test (DWBT). These tests are required by TS 5.5.12 "Containment Leakage Rate Testing Program," and Technical Specification Surveillance Requirement (SR) 3.6.5.1.1 respectively. The proposed change would permit the existing ILRT and DWBT frequency to be extended from 10 years to 11.5 years.

The proposed change

would, based on current refueling outage (RFO) projected schedules, allow Entergy to minimize the impact of the ILRT and DWBT on critical path outage activities by not having to perform the GGNS Unit 1 ILRT and DWBT prior to the expiration of the GGNS Unit 1 10-year interval for both tests. Currently, the ILRT and DWBT are to be performed approximately seven months prior to the 10th year anniversary of the completion of the last ILRT and DWBT (October 19, 2008). If granted, this revision would extend the period from 10 years to 11.5 years between successive tests. In terms of RFOs, this extension would move the performance of the next ILRT and DWBT from the scheduled spring 2018 End of Cycle 21 RFO to the spring 2020 End of Cycle 22 RFO.

The enclosure contains a description of the proposed changes, the supporting technical analyses, and the significant hazards considerations determination. of the enclosure provides the existing TS pages marked up to show the proposed changes. of the enclosure provides the retyped TS pages.

There are no regulatory commitments in this submittal.

GNRO-2016/00062 Page 2 of 3 Entergy has determined that this LAR does not involve a significant hazards consideration as determined per 10 CFR 50.92.

Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

Grand Gulf Nuclear Station has identified this change as affecting activities planned during the upcoming refueling outage (RF21) and on that basis requests approval of this proposed change no later than February 1, 2018, with an implementation date of February 18, 2018.

The requested approval date and implementation period will enable GGNS to optimize refueling outage planning and activities.

This request is to permit the deferral of the performance of the ILRT and DWBT until the 2020 refueling outage based upon the site's ILRT performance history and the evaluated low risk of the 18 month extension timeframe per the probabilistic determination.

Although this request is neither exigent nor emergency, your prompt review is requested.

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

This letter contains no new commitments. If you have any questions or require additional information, please contact James Nadeau at 601-437-2103.

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

Executed on the zs" day of December, 2016.

Attachment(s):

1.

2.

3.

cc: with Attachment(s)

Description and Assessment of Technical Specifications Changes Proposed Technical Specification Changes (Marked-Up)

Revised Technical Specification Pages (Clean)

Mr. John P. Boska, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0-8-C2 Washington, DC 20555

GNRO-2016/00062 Page 3 of 3 cc: without Attachment(s)

Mr. Mark Dapas Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 1600 East Lamar Boulevard Arlington, TX 76011-4511 U.S. Nuclear Regulatory Commission ATTN: Mr. James Kim, NRR/DORL (w/2)

Mail Stop OWFN/8 B1 11555 Rockville Pike Rockville, MD 20852-2738 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 Dr. Mary Currier, M.D., M.P.H State Health Officer Mississippi Department of Health P.O. Box 1700 Jackson, MS 39215-1700 Email: mary.currier@msdh.ms.gov to GNRO-2016/00062 Description and Assessment of Technical Specification Changes

GNRO-2016/00062 EVALUATION OF THE PROPOSED CHANGE

Subject:

License Amendment Request (LAR)for One Cycle Extension Of Appendix J Type A Integrated Leakage RateTest and Drywell Bypass Test Interval 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusion

6.0 REFERENCES

Attachments:

1. Technical Specification Pages (Mark-up) 2.

Retyped Technical Specification Pages

GNRO-2016/00062 1.0

SUMMARY

DESCRIPTION Entergy Operations, Inc. (Entergy) requests an amendment to Operating License (OL) No.

NPF-29 for the Grand Gulf Nuclear Station, Unit 1 (GGNS) to allow for a one cycle extension to the 1O-year frequency of the GGNS Unit 1 containment leakage rate test (Le., Integrated Leakage Rate Test (ILRT) or Type A test) and the drywell bypass leak rate test (DWBT). These tests are required by TS 5.5.12 "Containment Leakage Rate Testing Program,"

and Technical Specification Surveillance Requirement (SR) 3.6.5.1.1 respectively. The proposed change would permit the existing ILRT and DWBT frequency to be extended from 10 years to 11.5 years.

Currently, the ILRT and DWBT are to be performed approximately seven months prior to the 10th year anniversary of the completion of the last ILRT and DWBT (October 19, 2008). If granted, this revision would extend the period from 10 years to 11.5 years between successive tests. In terms of RFOs, this extension would move the performance of the next ILRT and DWBT from the scheduled spring 2018 End of Cycle 21 RFO to the spring 2020 End of Cycle 22 RFO.

2.0 DETAILED DESCRIPTION 2.1 Proposed Change The GGNS TS 5.5.12, "10 CFR 50, Appendix J, Testing Program," currently states, in part:

"This program establishes the leakage rate testing program of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be implemented in accordance with the Safety Evaluation issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087) as modified by the Safety Evaluation issued for Amendment No. 135 to the Operating License. For Type B and Type C local leakage rate testing, this program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," date July 2012.

Consistent with standard scheduling practices for Technical Specifications surveillances, intervals for the recommended surveillance frequency for Type A testing may be extended by up to 25 percent of the test interval, not to exceed 15 months. The calculated peak containment internal pressure, Pa, is 12.1 psig."

The proposed changes to GGNS TS 5.5.12, 10 CFR 50, Appendix J Testing Program, will be the administrative change to add the performance of the next Type A test no later than End of Cycle 22 RFO. The proposed change, shown in bold text, will revise TS 5.5.12, as follows, to state, in part:

"This program establishes the leakage rate testing program of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be implemented in accordance with the Safety Evaluation issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087) as modified by the Safety Evaluation issued for Amendment No. 135 to the Operating License, except that the next Type A test performed after the October 19, 2008 Type A test shall be performed no later than

GNRO-2016/00062 the plant restart after the End of Cycle 22 Refueling Outage. For Type B and Type C local leakage rate testing, this program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012.

Consistent with standard scheduling practices for Technical Specifications surveillances, intervals for the recommended surveillance frequency for Type A testing may be extended by up to 25 percent of the test interval, not to exceed 15 months. The calculated peak containment internal pressure, Pa, is 12.1 psig.

Surveillance Requirements, SR 3.6.5.1.1, Drywell, currently states, in part:

SR 3.6.5.1.1 SURVEILLANCE FREQUENCY Verify that bypass leakage is less than or equal to the bypass leakage limit.

However, during the first unit startup following drywell bypass leak rate testing performed in accordance with this SR, the acceptance criterion is leakage :::; 10%

of the bypass leakage limit.

120 months The proposed changes to GGNS SR 3.6.5.1.1 will be the administrative change to add the performance of the next DWBT test no later than End of Cycle 22 RFO. The proposed change, shown in bold text, will revise TS SR 3.6.5.1.1, as follows, to state, in part:

SR 3.6.5.1.1.

SURVEILLANCE Verify that bypass leakage is less than or equal to the bypass leakage limit.

However, during the first unit startup following drywell bypass leak rate testing performed in accordance with this SR, the acceptance criterion is teakaqe s 10%

of the bypass leakage limit.

FREQUENCY 120 months, except that the next drywell bypass leak rate test performed after the October 19, 2008 test shall be performed no later than the plant restart after the End of Cycle 22 Refueling Outage.

The mark-ups of TS 5.5.12 and SR 3.6.5.1.1 are provided in Attachment 1. The retyped TS pages for TS 5.5.12 and SR 3.6.5.1.1 are provided in Attachment 2.

There are no regulatory commitments being made in this LAR.

3.0 TECHNICAL EVALUATION

3.1 Justification for the Technical Specification Change

GNRO-2016/00062 3.1.1 Chronology of Testing Requirements of 10 CFR 50, Appendix J 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 Band 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, Regulatory Guide (RG) 1.163 (Reference 1) was issued. The RG endorsed Nuclear Energy Institute (NEI) 94-01, Revision 0 (Reference 4), 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 (Le., 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 5) and Electric Power Research Institute (EPRI) TR-104285 (Reference 6) 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. II 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 ILRTintervals 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

GNRO-2016/00062 leakage rate surveillance testing frequencies.

Justification for extending test intervals is based on the performance history and risk insights.

In 2012, NEI 94-01, Revision 3-A (Reference 2), 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 (Reference 1) and NRC SERs of June 25, 2008 (Reference 7) and June 8, 2012 (Reference 8) 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 of 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.

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."

3.1.2 Current GGNS 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." On April 6, 1998, the NRC approved License Amendment No. 135 for GGNS authorizing the implementation of 10 CFR Part 50, Appendix J, Option B for Types A, Band C tests (Reference 13).

Current Technical Specification 5.5.12 requires that a program be established to comply with the containment leakage rate testing requirements of 10 CFR 50.54(0) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions.

The program is

GNRO-2016/00062 required to be in accordance with the guidelines contained in the Safety Evaluation issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087) as modified by the Safety Evaluation issued for Amendment No. 135 to the Operating License. For Type B and Type C local leakage rate testing, this program shall be in accordance with the guidelines contained in NEI 94 01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,II date July 2012. Consistent with standard scheduling practices for Technical Specifications required surveillances, intervals for the recommended surveillance frequency for Type A testing may be extended by up to 25 percent of the test interval, not to exceed 15 months. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 12.1 psig.

Type Band C Local Leak Rate Testing during the Operating Cycle 20 was at EPU containment accident pressure of 14.8 psig, which encompassed all components in the Appendix J program.

Testing showed no change due to pressure, which would be the similar expectation of Type A testing at the EPU pressure.

The differences between Amendment No. 135 and "guidance provided in RG 1.163 are delineated in the NRC's SE for Amendment No. 135.

RG 1.163 endorses, with certain exceptions, NEI 94-01, Revision 0, (Reference 4) as an acceptable method for complying with the provisions of Appendix J, Option B.

3.1.3 GGNS 10 CFR 50 Appendix J, Option B Licensing History April 26, 1995 The NRC granted exemption to GGNS from the requirements of 10 CFR Part 50, Appendix J, Section 111.0, to permit the selection of containment leakage rate testing intervals for components on the basis of performance (Reference 10).

GGNS proposed changes to the frequency of performing Types A, B, and C tests including changes to the frequency of leakage rate testing of air locks. The exemption was to remain in effect until Refueling Outaqe 9.

Exemption from Section 111.0.1 (a):

Type A tests shall be performed on a 10-year interval provided that the two previous consecutive Type A tests, performed on the test interval specified in Appendix J (three tests, at approximately equal intervals, in a 10-year period), have been successful.

If a Type A test is failed, and the failure is not due to a Type B or C component, acceptable performance shall be re-established by performing a Type A test within 48-months of the unsuccessful Type A test.

Following a successful Type A test, the surveillance frequency may be returned to once per 10 years.

In addition, the licensee must perform general inspections of the accessible interior and exterior surfaces of the containment structures, as specified in Section V.A of Appendix J, at the Type A test interval specified in Appendix J, even when no Type A test is required during that outage.

This exemption shall be valid from the beginning of Refueling Outage 7 to the first startup following Refueling Outage 9.

Exemption from Sections 111.0.2 and 111.0.3 of Appendix J:

Type Band C testing shall be performed according to the following algorithm.

After two successful consecutive tests, performed at the Appendix J test interval of no more than two years, a Type B or C component may be tested once every 5 years.

If this test or a

GNRO-2016/00062 subsequent test is a failure, the test interval for this component shall revert to a 2-year interval until the component passes two consecutive tests.

The 5-year interval may then be resumed.

Main steam isolation valves, feedwater valves and containment system supply and exhaust isolation valves shall remain on a 2-year test interval. Any change will require prior review and approval by the NRC. The exemption shall be valid from the beginning of Refueling Outage 7 to the first startup following Refueling Outage 9.

Exemption from Section 111.0.2 (b)(i) and (b)(iii):

Air locks may be leakage rate tested at intervals of no more than 2 years. If an air lock fails a leakage rate test, the air lock shall then be required to pass two consecutive leakage rate tests at a test interval of 6 months prior to returning to the 2-year test interval.

Following opening of an air lock door when containment integrity is required, the air lock shall be tested at least every 30 days. If an air lock fails a leakage rate test following opening of an air lock door when containment integrity is required, the air lock shall be required to pass two consecutive leakage rate tests at a test interval of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to returning to the 30-day interval.

Since the Grand Gulf air lock doors have testable seals, testing the seals fulfills the 30-day test requirement.

This exemption shall be valid from the beginning of Refueling Outage 7 to the first startup following Refueling Outage 9.

August 1, 1996 The NRC issued License Amendment No.

126 for GGNS (Reference 11).

The amendment revised and deleted surveillance requirements, notes, and action statements involved with the requirements for the drywell leak rate testing, and the air lock leakage and interlock testing in TS 3.6.5.1 (Orywell), 3.6.5.2 (Orywell air lock), and 3.6.5.3 (Orywell Isolation Valves).

October 18, 1996 The NRC approved License Amendment No. 128 for GGNS (Reference 12).

The amendment revised the TS to modify the frequency requirements in surveillance requirement (SR) 3.6.1.3.5 on the leakage rate testing for each containment purge isolation valve with resilient seals to permit these valves to be leakage rate tested on a performance basis in accordance with 10 CFR 50, Appendix J.

April 6, 1998 The NRC approved License Amendment No. 135 for GGNS (Reference 13).

The amendment revised the TS to permit the implementation of the containment leak rate testing provisions of 10 CFR Part 50, Appendix J, Option B.

Specifically, this revision established a 10 CFR 50, Appendix J, Testing Program, and added this program to the TS.

This program references the NRC's SER on the GGNS' exemption to Appendix J, dated April 26, 1995 (Reference 10), as a method acceptable to the NRC for complying with Option B. This included changes to existing TS SRs 3.6.1.1.1, 3.6.1.2.1, 3.6.1.3.5, 3.6.1.3.8, 3.6.1.3.9, and addition of the "10 CFR Part 50, Appendix J, Testing Program" as TS 5.5.12. The applicable TS Bases were also modified.

As stated in the NRC's SE for Amendment No.135 (Reference 13), the NRC's April 26, 1995, SER (Reference 10) limited the test intervals for Types Band C testing to 5 years.

GGNS had opted to extend the Type B test interval to 10 years and keep the Type C interval at its present value of 5 years. This was consistent with RG 1.163.

GNRO-2016/00062 In addition, according to Reference 13, GGNS also opted to use alternative testing or analysis in lieu of as-found tests when maintenance is performed.

RG 1.163 does not endorse use of alternative testing or analysis in lieu of as-found testing. However, GGNS stated it was current practice to use valve operation test and evaluation system (VOTES) testing in lieu of a local leakage rate test (LLRT) for maintenance that does not affect leak-tightness, which GGNS defined as maintenance that affects only the valve actuator.

GGNS stated that an LLRT would only be performed if VOTES test detected a degraded thrust value, which could indicate seat leakage. This position is consistent with Appendix J, Option B and was acceptable to the NRC.

In addition, GGNS also proposed that following opening of an air lock door when containment integrity is required, the air locks shall be tested at least every 30 days. This 30-day test requirement may be satisfied by testing the air lock door seals.

The NRC found this acceptable, since the differences between the GGNS proposal and the testing mandated by NEI 94-01 are not significant.

The NRC determined that the use of the guidance of the April 26, 1995, SER is consistent with the intent of RG 1.163 (Reference 1) and was therefore acceptable. (GNRI-98/00028)

March 14, 2001 The NRC approved License Amendment No. 145 for GGNS (Reference 14).

The amendment consisted of changes to the facility OL and TS for a full-scope implementation of the alternative source term (AST).

Among these changes was a revision to TS SR 3.6.1.3.8, Main Steam Isolation Valve (MSIV) Leakage Rate, was amended to increase the maximum allowable leak rate to less than or equal to 100 standard cubic feet per hour (scfh) per main steam line (MSL) with a total leak rate through all four MSLs of less than or equal to 250 scfh (from less than or equal to 100 scfh through all four MSLs).

This amendment also revised TS 1.1, "Definitions," to reference new dose conversion factors and to increase the maximum allowable primary containment leakage rate from 0.437 percent to 0.682 percent of primary containment air weight per day. This value is based on 0.35 percent per day from the containment leak and an additional 100 scfh (0.087 percent per day) through the steam lines. (GNRI-2001/0032)

January 28, 2004 The NRC approved License Amendment No. 164 regarding the one-time extension of the integrated leak rate test (ILRT) and drywell bypass test interval for GGNS (TAC NO.

MB8940).

(Reference

15) The amendment changes the administrative TS 5.5.12 regarding containment integrated leakage rate testing (ILRT) and TS 3.6.5.1.1 regarding drywell bypass leak rate testing (DWBT).

The change would allow for a one-time extension of the interval from 10 to 15 years for performance of the next ILRT and DWBT.

This change would add an exception to the commitment to implement the containment ILRT program in accordance with the Safety Evaluation (SE) issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087), as modified by the SE issued for Amendment No. 135 to the Operating License.

Specifically, GGNS revised TS 5.5.12 by adding to the end of the second sentence the following:

", except that the next Type A test performed after the November 24, 1993 Type A test shall be performed no later than November 23, 2008."

GNRO-2016/00062 In addition, GGNS also revised TS 3.6.5.1.1 by adding an exception to the Frequency requirement of 120 months that states:

", except that the next drywell bypass leak rate test performed after the November 24, 1993 test shall be performed no later than November 23, 2008."

These changes represented a one-time deferral of the ILRT and the DWBT by up to five additional years.

July 12, 2005 The NRC approved License Amendment No. 168 for GGNS (Reference 16).

The amendment revised the air lock surveillance test acceptance criteria to be consistent with the NRC approved industry TS Task Force (TSTF) change to the Standard TS, TSTF-52, entitled, "Implement 10 CFR Part 50, Appendix J, Option B."

In summary, GGNS proposed to adopt the containment air lock leakage rate specified as a percentage of the maximum allowable primary containment leakage La, in the ISTS rather than the absolute leakage rate previously specified in the GGNS TS.

August 24, 2007 The NRC approved License Amendment No. 176 for GGNS (Reference 17).

This amendment revised the GGNS TS to allow certain types of relief valves to be used to isolate a containment penetration flow path without being deactivated under specific criteria.

The NRC has allowed similar types of penetrations and valves to be excluded from the scope of Appendix J containment leakage testing through issuance of 10 CFR 50.69, "Risk-Informed Categorization and Treatment of Structures,

Systems, and Components for Nuclear Power Reactors."

The basis for these approvals was that containment leakage through these types of penetrations and valves were determined to not contribute in a significant way to diminishing safety or increasing risk. (Reference 17; GNRI-2007/00101)

July 18, 2012 The NRC approved License Amendment No.191 for GGNS to increase the maximum steady-state reactor core power level by approximately 15%

from the original licensed thermal power level of 3,833 MWt [i.e., extended power uprate (EPU)]. (Reference 18)

The license was amended to including a new license condition 2.C.(44) for the SRs related to leak rate tests associated with TS 5.5.12 [10 CFR 50, Appendix J, Testing Program] are not required to be performed until their next scheduled performance dates.

These tests will be performed at the EPU calculated peak containment pressure or within EPU drywell bypass leakage limits, as appropriate.

This amendment also changed TS 5.5.12 calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, from 11.5 psig to 14.8 psig.

December 26, 2013 The NRC approved License Amendment No. 197 for GGNS (Reference 19).

The amendment revised the TS for GGNS to support operation with 24-month fuel cycles.

Specifically, the amendment revised the frequency of certain TS surveillance requirements from 18 months to 24 months in accordance with Generic Letter 91-04, "Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle," dated April 2, 1991. (GNRI-2013/00187)

GNRO-2016/00062 August 31, 2015 The NRC approved License Amendment No. 205 for GGNS (Reference 27). The amendment TSs for GGNS to allow plant operation from the currently licensed Maximum Extended Load Line Limit Analysis (MELLLA) domain to plant operation in the expanded MELLLA Plus (MELLLA+) domain under the previously approved extended power uprate condition of 4408 megawatts thermal rated core thermal power.

This amendment also changed TS 5.5.12 calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, from 14.8 to 12.1 psig.

February 17, 2016 The NRC approved License Amendment No. 209 for GGNS (Reference 20). The amendment revised the GGNS TS to allow for a permanent extension of the Type C leakage rate testing frequency and reduction of the Type Band C grace intervals that are required by GGNS TS 5.5.12, "10 CFR 50, Appendix J, Testing Program," by including a reference to Nuclear Energy Institute (NEI) Topical Report, NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012. In addition, the amendment changed TS 5.5.12 and Surveillance Requirement (SR) 3.6.5.1.1 by deleting the information regarding the performance of the last ILRT/DWBT that have already occurred. (GNRI-2016/00020) 3.2 System Descriptions GGNS is designed with a General Electric Company boiling water reactor (BWR) enclosed by a Mark III type containment.

The drywell is enclosed within the primary containment and is designed to divert the energy released during a design-basis, large-break loss of coolant accident (LOCA). The drywell communicates with the primary containment through a series of horizontal vents in the drywell wall. The vents are covered both inside and outside the drywell by water from the annular shaped suppression pool.

The pool forms a seal between the drywell and the primary containment.

The drywell contains the reactor coolant system and other high energy piping systems.

The GGNS Updated Final Safety Analysis Report (UFSAR), Section 6.2 describes the primary containment in detail.

3.2.1 Containment Building Description The Containment structure is designed to house the primary nuclear system and is part of the containment system whose functional requirement is the control of the release of radioactivity from a primary nuclear system.

The containment consists of three basic parts: a flat circular foundation mat, a right circular cylinder, and a hemispherical dome.

The containment cylindrical wall, dome, and foundation mat are constructed of cast-in-place, conventionally reinforced concrete.

For the most part, the Containment wall and foundation mat are separated by a 2-inch gap (which is filled with a compressible joint filler material) from the auxiliary building, to preclude significant interaction of these Category I structures during seismic disturbances. Type Band C Local Leak Rate Testing during the Operating Cycle and Refueling Outage 20 was at EPU containment Accident pressure of 14.87 psig, which encompassed all components in the Appendix J program.

Testing showed no change due to pressure which would be the similar expectation of Type A testing at the EPU pressure.

GNRO-2016/00062 3.2.2 Dimensions of Containment:

Inside diameter (10): 124 ft. 0 in. (based on cylindrical wall inside radius of 62 ft.)

Height of cylinder (top of foundation mat to dome spring line): 144 ft. 9 in.

Inside radius of cylindrical wall: 62 ft. 0 in.

Thickness of cylindrical walls: 3 ft. 6 in. (4 ft. 0 in. only in localized areas)

Inside radius of dome: 62 ft. 0 in.

Thickness of dome: 2 ft. 6 in.

Foundation mat thickness: 9 ft. 6 in.

Containment internal design pressure: 15 psig Containment airspace design temperature: 185 of Suppression pool design temperature: 210°F 3.2.3 Containment Penetrations and Attachments Two personnel airlocks (Upper and Lower) and an equipment hatch provide access to the Containment structure. Containment airlocks are tested in accordance with 10 CFR Part 50 Appendix J, Option B.

Each containment airlock door has two inflatable seals that are maintained at a nominal pressure of 70 psig.

Opening an airlock door, however, requires for its seals to be deflated.

Before the other door on the same airlock can be opened, this door must be closed and its seals must be re-inflated up to the 60 psig nominal interlock setpoint. This interlock ensures the pressure integrity of containment is maintained up to 56 psig when the airlocks are in use.

For the containment personnel locks, the airlock design incorporates provisions for testing between the door seals and between the doors (reference UFSAR Figure 6.2-85).

The provisions are (a.) Testing of annulus between seals and (b.) overall airlock pressure test.

Both tests can be run at a pressure of Pa.

Personnel air lock and equipment hatch openings penetrate the drywell cylindrical wall as shown on UFSAR Figures 3.8-58 and 3.8-61. Each of the two doors on the personnel air lock is fitted with two inflatable rubber seals to ensure the leak-tightness of the lock. The pressure within the seals can be monitored during normal operation to further ensure the integrity of the lock.

A horizontal fuel transfer tube penetration is provided at one end of the refueling pool to transfer fuel elements between the Containment and the Auxiliary Building. The location of the transfer tube penetration is shown on UFSAR Figure 3.8-1.

Piping penetrating the containment has been equipped with test connections and test vents or has other provisions to allow periodic leak rate testing to ensure that leakage is

GNRO-2016/00062 within the acceptable limit as defined by the Technical Specifications and Appendix J of 10 CFR 50, as described in UFSAR 6.2.6.

Typical mechanical and control systems penetrations are shown on UFSAR Figure 3.8-60.

These penetrations are detailed and designed to be leak-tight. During normal operation, the leakage past these penetrations will be negligible.

3.3 Traditional Engineering Considerations 3.3.1 Integrated Leakage Rate Testing (ILRT) History Previous Type A tests confirmed that the GGNS 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 (IWElIWL) program and maintenance rule monitoring provide confidence in containment integrity.

To date, five operational Type A tests have been performed on GGNS. There is considerable margin between these Type A test results and the TS 5.5.12 limit of 0.75 La (0.5115%

Weight per Day), where La is equal to 0.682%

Weight per Day of the containment air mass at the peak accident pressure. These test results demonstrate that GGNS has Iowa leakage Containment.

Table 3.3.1-1, Integrated Leakage Rate Testing (ILRT) History Test Date 950/0 UCL As-Left Leakage weight %

per day January 5, 1982' 0.083 0.083 November 4, 1985' 0.141 0.145 April 16, 1989' 0.129 0.133 November 21, 1993' 0.127 0.210 October 19, 20082 0.208 0.248 Notes:

1. The acceptance criterion for the ILRT is 0.75 La. At the time of these tests the limit was 0.328 wt%/day. The first test was a preoperational test performed under a special test procedure. Commercial operation began on July 1, 1985.

2.

The current limit would be 0.5115 wt%/day. This was changed as part of the implementation of the alternate source term (Reference 14).

3.3.2. Containment Leakage Rate Testing Program - Types Band C Testing Program GGNS Types Band C testing program requires testing of electrical penetrations, airlocks, hatches, flanges, and containment isolation valves in accordance with the Safety Evaluation issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087) as modified by the Safety Evaluation issued for Amendment No.

135 to the Operating License. For Type B and Type C local leakage rate testing, this program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part

GNRO-2016/00062 50, J," date July 2012.

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 Band 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 5.5.12 and SR 3.6.1.1.1 program requirements, the allowable maximum pathway total Types Band C leakage is 0.6 La where 0.6 La equals approximately 198,000 sccm (La equal approximately 330,000 sccm).

As discussed in NUREG-1493 (Reference 5), Types Band C tests can identify the vast majority of all potential containment leakage paths. Types Band C testing will continue to provide a high degree of assurance that containment integrity is maintained.

A review of the Types Band C test results from 2005 through 2016 for GGNS 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-Found minimum pathway leak rate average for GGNS shows an average of 9.93%

of 0.6 La with a high of 18.060/0 of 0.6 La or 0.1084 La.

The As-Left maximum pathway leak rate average for GGNS shows an average of 33.23%

of 0.6 La with a high of 61.63%

of 0.6 La or 0.3701 La.

Table 3.3.2-1 provides the LLRT data trend summaries for GGNS since 2005 and encompasses previous ILRTs.

This summary shows that there has been no As-Found failure that resulted in exceeding the TS 5.5.12 and SR 3.6.1.1.1 limit of 0.6 La (198,000 sccm) and demonstrates a history of successful tests. The As-Found minimum pathway summations represent the high quality of maintenance of Types Band C tested components while the As-Left maximum pathway summations represent the effective management of the Containment Leakage Rate Testing Program by the program owner.

Table 3.3.2-1, Types Band C LLRT Combined As-Found/As-Left Trend Summary RFO 2005 2007 2008 2010 2012 2014 2016 AFMin Path 5918 12,885 18,984 18,057 24,453 21,595 35,760 (seem)

Fraction of 7.41 La (%)

1.79 3.9 5.75 5.47 6.54 10.84 AL Max Path 20,288 18,389 57,793 69,850 93,069 79,014 122,136 (seem)

Fraction of La (%)

6.15 5.57 17.51 21.17 28.2 23.94 37.01 ALMin Path 3,027 2,189 23,457 25,065 30,415 35,054 35,474 (seem)

Fraction of 0.66 La (%)

0.92 7.11 7.60 9.22 10.62 10.75

GNRO-2016/00062 The following Table 3.3.2-2 identifies the components that have not demonstrated acceptable performance during the previous two outages for GGNS:

GNRO-2016/00062 Table 3.3.2-2: Types Band C LLRT Program Implementation Review Component As-Admin As-Left Cause of Scheduled Found Limit Corrective Action

[Penetration No.]

SCCM SCCM SCCM Failure Interval 2014 RF19 1B21F016 Valve identified Interval Main Steam Line Drain 7000 4780 200 as degraded in Valve replaced under remained at 24

[19]

RF18, disk WO 311176-01 months damaged 1G36F106 Valve identified Valve repaired under Interval RWCU Backwash Xfer Pump Disch 9500 1040 370 as degraded in WO 00235386 remained at 24

[49]

RF18 months 1E12F028B Valve identified Valve repaired under Interval Low Pressure Core Injection From RHR B 8800 4680 4160 as degraded in WO 00310025 remained at 24

[21]

RF18 (1 )

months 1P71 F149 Disk and seat Valve was replaced in Interval 1270 100 0

remained at 24 Chilled Water Return had scratches RF19 months

[39]

NOTES:

(1) During the post-LLRT for 1E12F028B under WO 00310025, 1E12F028B passed the LLRT with leakage of 4160 sccm against an acceptance criterion of 4680 sccm. Although the post LLRT passed the acceptance criteria, this valve should be repaired or replaced in RF20.

GNRO-20 16/00062 Table 3.3.2-2: Types Band C LLRT Program Implementation Review (continued)

Component As-Admin As-left Cause of Scheduled found Limit Corrective Action

[Penetration No.]

SCCM SCCM SCCM Failure Interval 2016 RF20 Adjusted 1833F125

>20000 260 2

limit/torque Valve repaired under Interval set at 24

[81]

setting on WO 439571 months 833F125to reduce leakage.

GNRO-2016/00062 3.3.3 Type B and Type C Tested Components on Extended Intervals The percentage of the total number of GGNS Type B tested components (78) that are on 120-month extended performance-based test interval is 65%.

The percentage of the total number of GGNS Type C tested components (151) that are on 60-month extended performance-based test interval is 58%.

3.3.4 Drywell Bypass Leakage Rate Testing The current interval for the GGNS drywell bypass test (DWBT) surveillance is one in ten years. Since the DWBT and the ILRT require some of the same equipment and steps and since they both have been on the same schedule frequency, it is desirable to extend the DWBT also. This will allow the two tests to remain on the same test frequency.

The DWBT verifies that pre-existing drywell bypass leakage does not exceed the maximum allowed leakage. The DWBT acceptance criterion in the Tech Specs is <10% of the analyzed design limit. The design bypass limit is used to establish the timing of automatic containment sprays following a LOCA. The sprays effectively control the containment pressure to less than its design limit (15 psi) by suppressing the steam from the drywell that bypasses the suppression pool.

This Surveillance ensures that the actual drywell bypass leakage is less than or equal to the acceptable A/...Jk design value of 0.8 fe assumed in the safety analysis. SR 3.6.5.1.1, specifically it states:

Verify bypass leakage is less than or equal to the bypass leakage limit.

However, during the first unit startup following drywell bypass leak rate testing performed in accordance with this SR, the acceptance criterion is leakage s 10%

of the bypass leakage limit.

As described in Section 3.1.2 of the NRC SER dated 8/1/96 addressing the increase in surveillance interval, the significant (90%) margin below the design limit allows for degradation of the drywell integrity until the next test (Reference 11).

The value that is 10% of 0.80 ft2 (0.08 fe) corresponds to the maximum allowable drywell bypass leakage rate of 3,072 scfm.

A summary of the DWBT results are presented in Table 3.3.3-1 below:

Table 3.3.4-1, Drywell Bypass Leakage Rate Testing (DWBT) History Outage Mo/DayNear Drvwell ByPass Leakage Pre-Op 1/5/1982 610.45 scfm 3/14/1983 1621.00 scfm 6/14/1984 2599.17 scfm 11/6/1985 2315.00 scfm RF01 11/11/1986 1568.40 scfm RF02 12/30/1987 1500.30 scfm

GNRO-2016/00062 Table 3.3.4-1, Drywell Bypass Leakage Rate Testing (DWBT) History Outage Mo/DayNear Drvwell ByPass Leakage RF03 4/15/1989 1631.01 scfm RF04 11/20/1990 1591.00 scfm RF05 5/31/1992 618.20 scfm RF06 11/24/1993 868.73 scfm RF16 10/19/2008 0.019 fe 3.3.5 Monitoring Drywell Leakage On August 1, 1996, the NRC issued an amendment to the GGNS Facility Operating License (TAC NO. M94176), that revised the TS to allow a performance-based drywell bypass leakage surveillance test. The NRC requested that GGNS monitor the drywell for significant leakage during operation. GGNS committed to assess the leaktightness of the drywell at least once each operating cycle.

The assessment is actually performed every quarter by running the drywell purge compressor to pressurize the drywell. The drywell purge compressors are part of an engineered safety system which forces air from the primary containment into the drywell.

The compressors are required to be operated for at least 15 minutes every quarter and an assessment is performed in conjunction with these surveillances.

There are two compressors, each of which can provide approximately 1000 scfm. The assessment considers whether a compressor is capable of increasing the pressure in the drywell. The Staff concluded in the Safety Evaluation that the proposed method provided reasonable assurance the TS value of drywell bypass leakage would not be exceeded. This regular monitoring of drywell leakage helps to ensure that there is no significant undetected degradation of the drywell.

3.3.6 Supplemental Inspections GGNS SR 3.6.5.1.2 requires the visual inspection of the exposed accessible interior and exterior surfaces of the drywell at a frequency of once prior to performance of each Type A test required by SR 3.6.1.1.1.

The exposed accessible drywell interior and exterior surfaces are inspected to ensure there are no apparent physical defects that would prevent the drywell from performing its intended function. This SR ensures that drywell structural integrity is maintained. The Frequency was chosen so that the interior and exterior surfaces of the drywell can be inspected in conjunction with the inspections of the primary containment required by 10 CFR 50, Appendix J. Due to the passive nature of the drywell structure, the specified Frequency is sufficient to identify component degradation that may affect drywell structural integrity.

3.3.7 Service Level I Coatings Assessment Purpose The purpose of the Service Level I Coatings Assessment program is to monitor the

GNRO-2016/00062 condition of Service Level I coatings and provide an effective method to assess coating condition through visual inspections. To identify degraded or damaged coatings and provide a means for repair of identified problem area.

Significance and Use A coating monitoring program provides early identification and detection of potential problems in coating systems. Degraded coatings have the potential to fail if they are not upgraded/repaired by a maintenance program. Failure of coating material and rust may generate debris under design basis accident conditions that could adversely affect the performance of post-accident safety systems, such as ECCS Suction strainers.

Establishment of a ongoing inservice monitoring program allows for planning and scheduling of priority coating activities to ensure the integrity and performance of service level one coating systems.

Frequency of Inspections Inspections of coatings in the drywell are to be performed during Refueling Outages.

Containment inspection may be performed during operation. Time between inspections will not exceed 24 months.

Records and Past History of Existing Coatings The last two performance monitoring reports pertaining to the service level one coating systems should be reviewed prior to the monitoring process.

Inspection Plan Perform a walk-through visual inspection on all readily accessible coated surfaces if historical information is not available. After the walkthrough, detailed visual inspection shall be performed on previously designated areas and on areas documented as possible deficiencies by the initial walk-through inspection.

Where defects exist on the containment boundary, the following rules apply:

Notify the containment Responsible Professional Engineer (Design Engineering Programs) of location and extent of degradation potentially affecting the containment boundary such as blisters, cracks, corrosion, which affects the base metal.

VT-3 Visual Exam must be performed before removing coatings or surface preparation as per ASME Section XI. Following repair or reapplication of coating, a copy of the Coatings Inspection Report shall be attached to the VT-3 Report.

Identification of visible defects such as blistering, cracking, flaking/peeling, rusting, and physical damage.

Blistering - compare any blistering found to the blistering pictorial standards of coating defects (refer to test method 0714) and record size and frequency, if blisters are larger than those on the comparison photographs, measure, record size and extent of surface area affected. Photograph area and report if the blisters are intact.

Cracking - can be limited to one layer of coating or extend through to the substrate. Measure the length of the crack or if extensive cracking has occurred,

GNRO-2016/00062 measure the size of the area affected. Determine if cracking is isolated or is part of a pattern. Record depth of crack length, and pattern of crack on the inspection report, photograph the area affected.

Flaking Peeling Delamination - Measure the size of peels and note pattern formed.

Carefully check to see if lifting can be achieved beyond the obvious peeled area.

Note all observations on the inspection report and photograph the area affected.

Rusting - compare with the pictorial standards ASTM of test method 0610 to determine the degree of rusting. Try to determine the source of rusting, is it surface stain caused by rust in another location or is it a failure of the coating allowing the substrate to rust.

Photograph the affected area and record observations.

If no defects are found - mark "coatings intact, no defects" on the inspection reports.

If portions of the coating cannot be inspected - note the specific areas on the location map-inspection report, along with the reason why the inspection cannot be conducted.

Written or photographic documentation, or both of coating inspection area, failures, and defects shall be made and the process of documentation will be determined by the inspection coordinator. Practice ASTM 04121 provides one method to obtain consistent comparable close up photographs.

For coating surfaces determined to be suspect, defective or deficient, one or more physical tests, such as dry film thickness (test methods ASTM 01186 and SSPC-PA-2),

adhesion ASTM (Test Methods 03359 and 04541), and continuity (NACE RP0188-88) may be performed and evaluated by the coating specialist. Samples may be gathered, and the size and extent of defective patterns may be described.

Evaluation The inspection reports should be evaluated by responsible qualified evaluation personnel.

The evaluation personnel shall prepare a report that includes a summary of findings and recommendations for future surveillance or repair; this would include an analysis of the reasons or suspected reasons for failure. The repair work should be prioritized into large and small defective areas. A recommended corrective plan or action must be provided for the large, (area larger than 1 sq. ft.), defective areas so that the plant can repair these areas, if required during the same outage.

Condition Reports must be written on Nonconforming Items. A coating failure such as loss of adhesion, delamination, blistering, flaking, etc. is considered nonconforming. Damaged coating areas are considered a maintenance item and will be addressed by a Work Request.

3.3.8 Containment Inservice Inspection Program Introduction This Program Section contains the details of the ASME Section XI, Division 1, Containment Inservice Inspection (CISI)

Program for GGNS.

Implementation of a Containment Inservice Inspection Program in accordance with the requirements of ASME

GNRO-2016/00062 Section XI, Division 1, is mandated by 10 CFR 50.55a.

This program is applicable to the third 120-month Containment Inservice Inspection Interval. The ASME Section XI Code of Record for the CISI Program during the third interval is the 2001 Edition with the 2003 Addenda. The Containment Inservice Inspection Interval designation is based on the Inservice Inspection Interval.

The scope of this Program Section includes the examination and testing of ASME Class CC and MC components and their integral attachments.

Background

The Grand Gulf Nuclear Station containment system is a Mark III pressure suppression containment system consisting of a drywell, vapor suppression pool and a primary containment structure. The cylindrical, reinforced concrete primary containment structure forms the containment pressure boundary and encloses the suppression pool and the drywell. The inspections associated with this program are limited to the primary containment structure and its appurtenances. Inspections of the drywell are outside the scope of this program plan.

The initial Containment lSI Program commenced on June 2, 1997 and continued through June 1, 2007. The interval was extended until May 31, 2008, as permitted under IWA-2430(d).

Code of Federal Regulations Requirements The Code of Federal Regulations Final Rule that affects the lSI Program Update for GGNS is the 10CFR50.55a Final Rule published September 29, 2005 (70FR188).

70FR188 incorporated by reference ASME Section XI, 2001 Edition with 2003 Addenda in paragraph (b)(2) and was effective November 1,2004.

Scope of the Containment lSI Program The scope of the Containment lSI Program shall include class MC and Class CC components to include those items required under 10CFR50.55a(g)(4)(v) to be treated as either class MC or class CC.

10CFR50.55a(g)(4)(v) requires:

(A)

Metal containment pressure retaining components and their integral attachments must meet the inservice inspection,

repair, and replacement requirements applicable to components which are classified as ASME Code Class MC; (B) Metallic shell and penetration liners which are pressure retaining components and their integral attachments in concrete containments must meet the inservice inspection, repair, and replacement requirements applicable to components which are classified as ASME Code Class MC; and (C) Concrete containment pressure retaining components and their integral attachments, and the post-tensioning systems of concrete containments must meet the inservice inspection, repair, and replacement requirements applicable to components, which are classified as ASME Code Class CC In accordance with IWA-1320 and IWE-1100, Class MC pressure retaining components

GNRO-2016/00062 and their integral attachments and metallic shell and penetration liners of Class CC pressure retaining components and their integral attachments are subject to the requirements of Subsection IWE.

In accordance with IWA-1320 and IWL-1100, Class CC reinforced concrete and post-tensioning systems of concrete containments are subject to the requirements of Subsection IWL.

ASME Section XI Code of Record In accordance with 10CFR50.55a(g), Entergy is required to update the ASME Section XI (the Code) Containment lSI Program once every ten years. The updated Containment lSI Program is required to comply with the latest edition and addenda of the Code incorporated by reference in 10CFR50.55a one year prior to the start of the interval per 10CFR50.55a(g)(4)(ii). The prior interval dates for GGNS were June 2, 1997 through May 31, 2008 [2nd Interval].

Based on "lock in" date of May 31, 2007, the 2001 Edition with the 2003 Addenda of ASME Section XI is the version of Section XI that Entergy must meet for the current interval [3rd Interval].

ASME Section XI Code Cases Per 10CFR50.55a(g), ASME Code Cases that have been determined to be suitable for use in lSI Program Plans by the NRC are listed in RG 1.147 "Inservice Inspection Code Case Acceptability-ASME Section XI, Division 1". The use of Code Cases (other than those listed in Regulatory Guide 1.147) may be authorized by the Director of the Office of Nuclear Reactor Regulation upon request pursuant to 10CFR50.55a(a)(3). The ASME Section XI Code Cases incorporated into the Containment lSI Program Plan are listed in Table 3.3.7-1. At the time this Containment lSI Program Plan was originally issued, Revision 15 of Regulatory Guide 1.147 was the latest revision. At the time this Containment lSI Program Plan was issued, Revision 16 of Regulatory Guide 1.147 was the latest revision.

TABLE 3.3.8-1 CODE CASES INCORPORATED INTO THE CISI PROGRAM Code Case Title/

RG 1.147 Revision/

Regulatory Guide 1.147 Conditions For Request for Alternative Use N-532-4 Alternative Requirements to Repair and 15 Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission N-624 Successive Inspections 14 N-739 Alternative Qualification Requirements for 16 Personnel Performing Class CC Concrete and Posttensioning System Visual Examinations N-753 Vision Tests 16

GNRO-20 16/00062 Requests For Relief and Requests For Alternatives In cases where the ASME Section XI requirements for inservice inspection are considered impractical, requests for relief may be submitted in accordance with 10CFR50.55a(g)(5)(iii).

In cases where alternatives to the ASME Section XI requirements exist that would provide an acceptable level of quality and safety, a Request for Alternative may be submitted to the NRC in accordance with 10CFR50.55a(a)(3)(i).

In cases where compliance with the specified requirements of ASME Section XI would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, a Request for Alternative may be submitted to the NRC, as allowed by 10CFR50.55a(a)(3)(ii).

Per 10CFR50.55a paragraphs (a)(3) and (g)(6)(i), the Director of the Office of Nuclear Reactor Regulation will evaluate Requests for Relief and Requests for Alternatives per Paragraph (g)(5) and II ***may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility", There are no currently pending Requests for Relief and Requests for Alternatives for the current interval are included.

General Requirements-Subsection IWE and IWL Examinations The examinations conducted under the Containment Inservice Inspection Program are performed to meet the requirements of ASME Section XI Subsections IWE and IWL as modified by 10CFR50.55a. The following modifications apply to the 2001 edition through 2003 addenda.

In accordance with 10CFR50.55a(b)(2)(viii),

Examination of concrete containments, Licensees applying Subsection IWL, 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section, shall apply paragraphs (b)(2)(viii)(E) through (b)(2)(viii)(G) of this section.

In accordance with 10CFR50.55a(b)(2)(vii)(E) for Class CC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the lSI Summary Report required by IWA-6000:

(1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2) An evaluation of each area, and the result of the evaluation, and; (3) A description of necessary corrective actions.

In accordance with 10CFR50.55a(b)(2)(vii)(F)

Personnel that examine containment concrete surfaces and tendon hardware, wires, or strands must meet the qualification provisions in IWA-2300. The "owner-defined" personnel qualification provisions in IWL-

GNRO-2016/00062 2310(d) are not approved for use.

An alternative to the requirements of 10CFR50.55a(b)(2)(vii)(F) is included in Code Case N-739. Entergy has requested the use of these alternative requirements in Request for Alternative CEP-CISI-001.

10CFR50.55a(b)(2)(vii)(G) applies to Corrosion protection material used in post tensioning systems. This requirement is not applicable to GGNS because the containment design does not use a post tensioning system.

In accordance with 10CFR50.55a(b)(2)(ix), Examination of metal containments and the liners of concrete containment, Licensees applying Subsection IWE, 1998 Edition through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section, shall satisfy the requirements of paragraphs (b)(2)(ix)(A), (b)(2)(ix)(B), and (b)(2)(ix)(F) through (b)(2)(ix)(I) of this section.

In accordance with 10CFR50.55a(b)(2)(ix)(A) For Class MC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the lSI Summary Report as required by IWA-6000:

(1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2) An evaluation of each area, and the result of the evaluation, and; (3) A description of necessary corrective actions.

In accordance with 10CFR50.55a(b)(2)(ix)(B) When performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2210-1 may be extended and the minimum illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination.

In accordance with 10CFR50.55a(b)(2)(ix)(F), VT-1 and VT-3 examinations must be conducted in accordance with IWA-2200.

Personnel conducting examinations in accordance with the VT-1 or VT-3 examination method shall be qualified in accordance with IWA-2300. The "owner-defined" personnel qualification provisions in IWE-2330(a) for personnel that conduct VT-1 and VT-3 examinations are not approved for use.

In accordance with 10CFR50.55a(b)(2)(ix)(G), the VT-3 examination method must be used to conduct the examinations in Items E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 examination method must be used to conduct the examination in Item E4.11 of Table IWE-2500-1. An examination of the pressure-retaining bolted connections in Item E1.11 of Table IWE-2500-1 using the VT-3 examination method must be conducted once each interval. The "owner-defined" visual examination provisions in IWE-2310(a) are not approved for use for VT-1 and VT-3 examinations.

In accordance with 10CFR50.55a(b)(2)(ix)(H) Containment bolted connections that are disassembled during the scheduled performance of the examinations in Item E1.11 of Table IWE-2500-1 must be examined using the VT-3 examination method. Flaws or degradation identified during the performance of a VT-3 examination must be examined in

GNRO-2016/00062 accordance with the VT-1 examination method. The criteria in the material specification or IWB-3517.1 must be used to evaluate containment bolting flaws or degradation. As an alternative to performing VT-3 examinations of containment bolted connections that are disassembled during the scheduled performance of Item E1.11, VT-3 examinations of containment bolted connections may be conducted whenever containment bolted connections are disassembled for any reason.

In accordance with 10CFR50.55a(b)(2)(ix)(l) the ultrasonic examination acceptance standard specified in IWE-3511.3 for Class MC pressure-retaining components must also be applied to metallic liners of Class CC pressure-retaining components.

General Visual Examinations General Visual Examinations for Subsection IWE shall be performed in accordance with Program Section, "General Visual Examinations of Class MC Components".

The requirements of IWA-2210 are not applicable to Subsection IWE General Visual examinations.

General Visual Examinations for Subsection IWL shall be performed in accordance with Program Section, "General and Detailed Visual Examinations of Concrete Containments".

The requirements of IWA-2210 are not applicable to Subsection IWL visual examinations.

Detailed Visual Examinations for Subsection IWL shall be performed in accordance with Program Section, "General and Detailed Visual Examinations of Concrete Containments".

The requirements of IWA-2210 are not applicable to Subsection IWL visual examinations.

VT-1 and VT-3 Visual Examinations VT-1 Visual Examinations for Subsection IWE shall be performed in accordance with Program Section, "VT-1 Examination".

In accordance with 10CFR50.55a(b)(2)(ix)(G), the VT-1 examination method must be used to conduct the examination in Item E4.11 of Table IWE-2500-1. The "owner-defined" visual examination provisions in IWE-2310(a) are not approved for use for VT-1 examinations.

VT-3 Visual Examinations for Subsection IWE shall be performed in accordance with Program Section, "VT-3 Examination".

In accordance with 10CFR50.55a(b)(2)(ix)(G) the VT-3 examination method must be used to conduct the examinations in Items E1.12 and E1.20 of Table IWE-2500-1. An examination of the pressure-retaining bolted connections in Item E1.11 of Table IWE-2500-1 using the VT-3 examination method must be conducted once each interval. The "owner-defined" visual examination provisions in IWE-2310(a) are not approved for use for VT-3 examinations.

Remote VT-1 and VT-3 Visual Examinations When access or other conditions prevent direct examination, remote visual examination can be substituted for direct examination provided that the requirements of IWA-2210(c) are met. IWA-2210(c) requires that the selected test characters of IWA-2210(b) can be resolved as a part of the remote examination procedure demonstration. Remote visual

GNRO-2016/00062 examination aids include but are not limited to

mirrors, telescopes, periscopes, borescopes, fiberoptics, and Closed Circuit Television (CCTV) systems with or without permanent recording capabilities.

Per 10CFR5.55a(b)(2)(ix)(B), when performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2210-1 may be extended and the minimum illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination.

Volumetric Examinations Volumetric examinations to determine wall thinning shall use an ultrasonic thickness measurement method specified in ASME Section V, T-544.

One-foot square grids shall be used for the ultrasonic thickness measurement unless smaller grids are specified by the Responsible Individual. The location of the minimum wall thickness shall be recorded and/or marked such that periodic reexamination of the location can be performed.

In accordance with 10CFR50.55a(b)(2)(ix)(I), the ultrasonic examination acceptance standard specified in IWE-3511.3 for Class MC pressure-retaining components must also be applied to metallic liners of Class CC pressure-retaining components.

Alternative Examinations In accordance with IWE-2500(a) and IWA-2240, alternative examination methods may be used provided the ANI! is satisfied that the results are demonstrated to be equivalent or superior to the results of the method specified by Subsection IWE. The 1997 addenda of IWA 2240 must be used in accordance with 10CFR50.55a(b)(2)(xix) which states:

10CFR50.55a(b)(2)(xix) Substitution of alternative methods. The provisions for the substitution of alternative examination methods, a combination of methods, or newly developed techniques in the 1997 Addenda of IWA-2240 must be applied.

The provisions in IWA-2240, 1998 Edition through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section, are not approved for use.

The provisions in IWA-4520(c), 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section, allowing the substitution of alternative examination methods, a combination of methods, or newly developed techniques for the methods specified in the Construction Code are not approved for use.

Approved alternative examinations shall be documented by incorporation into this program section by a Program Change Notice (PCN).

Inspection Intervals-Subsection IWE Per IWA-2430 of ASME Section XI, the inservice examinations required by Subsection IWE shall be completed during each of the inspection intervals for the service lifetime of the power unit. The inspections shall be performed in accordance with Inspection Program B of IWA-2432.

GNRO-2016/00062 Per IWA-2430(d), for components inspected under Program B, each of the inspection intervals may be extended or decreased by as much as one year. Adjustments shall not cause successive intervals to be altered by more than one year from the original pattern of intervals.

Per IWA-2430(e), in addition to the interval adjustment allowed per IWA-2430(d), for power units that are out of service continuously for 6 months or more, the inspection interval during which the outage occurred may be extended for a period equivalent to the outage and the original pattern of intervals extended accordingly for successive intervals.

Inspection Schedule-Subsection IWE Per IWA-2420, inspection plans shall be prepared for the first inservice interval and subsequent inspection intervals.

Per IWA-2420(b), an implementation schedule for performance of examinations and tests shall be prepared for each inspection plan.

Subarticle IWE-2400 includes the requirements for the scheduling of examination and tests for Class MC components and Metallic Liners of Class CC components.

Specific scheduling criteria are included in IWE-2412 for plants, which are employing Inspection Program B. This paragraph references Table IWE-2412-1, which includes minimum and maximum percentages of examinations required to be completed by each inspection period. Table 3.3.7-1-1 summarizes this information. Figure 3.3.7-1 provides interval, period and approximate outage dates.

Per IWE-2412(a), the following examinations listed in Table IWE-2500-1 as deferrable to the end of the inspection interval are not required to meet the criteria in Table IWE-2412-1:

a) Category E-A, Item No E1.12 - Wetted Surface of Submerged Areas b) Category E-A, Item No E1.20- BWR Vent System Accessible Surface Areas TABLE 3.3.8-2 IWE COMPONENT SCHEDULING INSPECTION INSPECTION MINIMUM MAXIMUM INTERVAL PERIOD EXAMINATIONS EXAMINATIONS (CALENDAR COMPLETED, %

CREDITED, %

YEARS OF PLANT SERVICE WITHIN THE INTERVAL) 3rd 3

16 50 3rd 7

50' 75 3rd 10 100 100 1 If the first period completion percentage for any examination category exceeds 34%, at least 16% of the required examinations shall be performed in the second period.

GNRO-2016/00062 Figure 3.3.8-1: 3rd Interval Schedule RF 16 F 2008 RF 17 SP 2010 RF 18 SP 2012 RF 19 SP 2014 RF 20 SP 2016 1/2008

~

~

1/2009 1/2~0 1/2011 1/20~

IWllnspections 1/2013 1/20~

1/2015 1/201)6 1/2017 1/2018 IWllnpsections 1st Period 30 Months 5/31/2008 Start of 3rd interval 12/2010 2nd Period 48 Months 12/2014 3rd Period 30 Months 6/2017 End of 3rd Interval

GNRO-2016/00062 Inspection Periods-Subsection IWL Subarticle IWL-2400 includes the requirements for the scheduling of examination and tests for Class CC Concrete components. Table 3.3.7-3 summarizes the Subsection IWL Examination periods for GGNS.

Concrete examinations shall be conducted every 5 years (a period) as described in IWL-2410(a) and (c). For the purposes of this program section, an IWL inspection period shall be defined as the window of time allowed by IWL-2410 for the completion of one set of IWL examinations.

Concrete surface areas affected by repair/replacement activities shall be examined in accordance with IWL-2410(d).

TABLE 3.3.8-3 Proiected IWL Examination Periods Period Start Date End Date 30 Year 1/2/2011 1/2/2013 35 Year 1/2/2016 1/2/2018 40 Year 1/2/2021 1/2/2023 45 Year 1/2/2026 1/2/2028 50 Year 1/2/2031 1/2/2033 55 Year 1/2/2036 1/2/2038 60 Year 1/2/2041 1/2/2043 Subsection IWE Component Selection Criteria Class MC components are selected for examination per the requirements of the 2001 Edition with the 2003 Addenda of Section XI, Table IWE-2412-1. IWE-2420(a) states that the sequence of component and component support examinations established during the first interval shall be repeated during the successive intervals to the extent practical. Entergy will select and examine a majority of the Class MC components and component supports in accordance with this criteria. However, Entergy has adopted Code Case N-624 as approved in Regulatory Guide 1.147, which states that the sequence of examinations may be modified provided the percentage of requirements of Table IWE-2412-1 are met. Code Case N-624 allows the sequence of examinations established during the first interval to be modified by factors such as scaffolding erection, radiological concerns, insulation removal or other considerations.

The Containment Inspection Engineer selects examinations for a given interval and for a given period in accordance with the requirements of 10CFR50.55a and ASME Section XI.

Examination selections are documented in the IDDEAL database. The implementing organization determines the appropriate schedule for completion of the examinations within a given period.

Examination Category E-A Selection

GNRO-2016/00062 In accordance with Table IWE-2500-1, Examination Category E-A, Item E1.11, a General Visual examination of the accessible surface areas shall be conducted from BOTH the inside and outside surfaces, as accessible, each period. While the majority of the GGNS containment liner is inaccessible from the outside due to the design of the containment, outer surface areas that are accessible will be included in the examinations. The General Visual examination includes all accessible interior and exterior surfaces of Class MC components, parts and appurtenances, and metallic and shell penetration liners of Class CC components as discussed in Table IWE-2500-1, Note (1 )(a), (b) and (c).

Table IWE-2500-1 Examination Category E-A, Item E1.11 Note (1 )(d) covers pressure retaining bolted connections including bolts, studs, nuts, bushings, washers, threads in based material and flange ligaments between fastener holes.

A General Visual examination of 100%

of these components and parts, as accessible, shall be performed once each period.

In order to facilitate record keeping of bolted examinations, the bolted connections are be designated as Category E-A, Item Number E1.11b examinations in the IDDEAL database.

In accordance with 10CFR50.55a(b)(2)(ix)(H) Containment bolted connections that are disassembled during the scheduled performance of the examinations in Item E1.11 of Table IWE-2500-1 must be examined using the VT-3 examination method. Flaws or degradation identified during the performance of a VT-3 examination must be examined in accordance with the VT-1 examination method.

The criteria in the material specification or IWB-3517.1 must be used to evaluate containment bolting flaws or degradation. As an alternative to performing VT-3 examinations of containment bolted connections that are disassembled during the scheduled performance of Item E1.11, VT-3 examinations of containment bolted connections may be conducted whenever containment bolted connections are disassembled for any reason.

Table IWE-2500-1 Examination Category E-A, Item E1.11 Note (3) covers moisture barriers. The design of GGNS does not include moisture barriers. As a result no moisture barrier examinations are required. Table IWE-2500-1 Examination Category E-A, Item E1.12 refers to wetted surfaces of submerged areas. A VT-3 Visual examination of the submerged surfaces is required once each interval. 10CFR50.55a(b)(2)(ix)(G) requires that the VT-3 examination method be used instead of the General Visual method specified in IWE.

Table IWE-2500-1, Examination Category E-A, Item E1.20 refers to the vent systems of BWR Mark I containments and is not applicable to GGNS.

Examination Category E-C Selection In accordance with IWE-2500 (b)(1), Category E-C, Item No. E4.11 surfaces, surfaces that are accessible for visual examination, shall be examined visually. In lieu of the detailed visual method specified in IWE, the VT-1 Method shall be used in accordance 10CFR50.55a(b)(2)(ix)(G).

GNRO-2016/00062 Category E-C, Item No. E4.12 surfaces, surfaces that are not accessible for visual examination, shall be examined for wall thinning using an ultrasonic thickness measurement technique.

Subsection IWL Component Selection Criteria Class CC components were selected for examination per the requirements of the 2001 Edition with the 2003 Addenda of Section XI, IWL-2410, IWL-2420 and Table IWL-2500-

1. The IDDEAL database presents the selection of IWL surface areas for examination.

Examination Category L-A Selection In accordance with Table IWL-2500-1, Examination Category L-A Item L1.11, a General Visual examination of the accessible concrete surface areas shall be conducted once every five years.

In accordance with Table IWL-2500-1, Examination Category L-A, Item L1.12, a Detailed Visual examination of the suspect concrete surface areas shall be conducted once every five years.

Examination Category L-B Selection Once each IWL inspection period, the components of the unbounded Post-Tensioning System are examined in accordance with Table IWL-2500-1 Category L-B as modified by 10CFR50.55a. Since the containment structure at GGNS does not have a post-tensioning system; Category L-B examinations do not apply to GGNS.

IWE Augmented Examination of Containment Surface Areas - Examination Category E-C Containment surface areas likely to experience accelerated degradation and aging per IWE-1240 require examination in accordance with IWE-2500(b) and Table IWE-2500-1, Examination Category E-C. In accordance with IWE-2420(b), examinations accepted by evaluation per IWE-3000 shall be examined under Category E-C in the next inspection period. These areas shall be listed in the Iddeal database as Category E-C Item Number E4.11 and/or E4.12 as follows.

• For surfaces accessible for visual examination, the requirements of Examination Category E-C, Item No. E4.11, as modified by 10CFR50.55a(b)(2)(ix)(G), shall be implemented. Where access allows, the VT-1 examination shall be performed from both sides of the surface.

• For surfaces where the side requiring augmented examination is not accessible for visual examination, ultrasonic thickness measurements shall be performed in accordance with Examination Category E-C, Item No. E4.12, and in accordance with IWE-2520(b)(3) and IWE-2520(b)(4).

The examination(s) must be performed once per period until the areas examined remain essentially unchanged for the next inspection period. In accordance with' Table IWE-2500-1, Examination Category E-C, Note 2 and IWE-2420(c), if the areas examined

GNRO-2016/00062 remain essentially unchanged, they no longer require Examination Category E-C examination.

10CFR50.55a Limitations and Modifications The requirements of Table IWE-2500-1 Examination Category E-C, are modified by 10CFR50.55a as follows:

Per 10CFR50.55a(b)(2)(ix)(G), the VT-1 examination method must be used to conduct the examination in Item E4.11 of Table IWE-2500-1.

The "owner-defined" visual examination provisions in IWE-2310(a) are not approved for use for VT-1 examinations.

Per 10CFR50.55a(b)(2)(ix)(I),

the ultrasonic examination acceptance standard specified in IWE-3511.3 for Class MC pressure-retaining components must be applied to metallic liners of Class CC pressure-retaining components.

Identification of IWE Augmented Examination of Containment Surface Areas -

Examination Category E-C Whenever a plant has an area(s) requiring examination under Category E-C, the area shall be identified in the IDDEAL database. Areas requiring Category E-C examination are listed in Table 3.3.7-4.

Table 3.3.8-4 Areas Requiring Category E-C Examination Component Description Exam Requirements ID EC-01 Corroded liner plate area under Ultrasonic Containment Penetration 38.

EC-02 Corroded plate around Containment Ultrasonic Penetration 38.

EC-03 Corroded liner plate in expansion foam Ultrasonic area at AZ 45 decrees.

EC-04 Corroded liner plate in expansion foam Ultrasonic area at AZ 240 decrees.

EC-05 Corroded liner plate in expansion foam Ultrasonic area at AZ 270 deQrees.

1-FP-02F-2 Suppression Pool liner, Mech. Damage, Ultrasonic dent indication.

1-FP-08A-1 Suppression Pool liner, Mech. Damage, Ultrasonic other indication.

1-FP-08A-1 Suppression Pool liner, Mech. Damage, Ultrasonic dent indication.

1-FP-04D-4 Suppression Pool liner, Mech. Damage, Ultrasonic other indication.

1-WP-01C-3 Suppression Pool liner, Mech. Damage, Ultrasonic other indication.

GNRO-2016/00062 Tracking of Augmented Examination Areas IWE requires augmented examination of surface areas likely to experience accelerated degradation and aging as described in IWE-1241, or surface areas accepted by evaluation as specified in IWE-2420.

When it is determined that a given surface area requires augmented examination, the area shall be added to the IDDEAL database. Areas added to the augmented examination table due to the provisions of IWE-2420(b) may be removed from the IDDEAL database when the provisions of IWE-2420(c) have been met.

Augmented examination areas added to the IDDEAL database table due to the provisions of IWE-1241(a) or (b) may be removed from the IDDEAL database only after determination that the area is no longer likely to experience accelerated degradation or aging as described in IWE-1241 (a) or (b) as applicable.

3.3.9 RF20 Summary of IWE Examinations Suppression Pool Liner Inspection The indications documented on QAR-RF20-01 for the suppression pool liner are acceptable by examination in accordance with the requirements of ASME Section XI Subsection IWE based on the following:

QAR-RF20-01 documents five areas with degradation. For area 1, the area containing the worst case thickness loss, additional UT thickness measurements were obtained during RF20 and reported in BOP-UT-16-011. While additional UT thickness was not available for the remaining areas, extensive UT thickness measurements of the suppression pool liner were obtained in 2007 to support evaluation of indications detected in 2007. These measurements are documented in ER-GG-2007-0073-000 and C-G-0130.0 Supplement 3. The minimum wall thickness recorded for the suppression pool liner was 0.272 11 with the majority of readings between 0.278 11 and 0.295 11 in thickness. For the purposes of this review, the thickness is assumed to be the minimum of 0.272 11 for all areas except area 1 in the table below. For area 1 the minimum thickness reported in BOP-UT-16-011 is used for the general area plate thickness.

Table 3.3.9-1 VT-3 of GGNS Suppression Pool in RF20 Area Recorded Plate UT Thickness Min.

Plate Thinning Thickness 1

0.055 11 0.315 11 0.260 11 2

<0.030 11 0.272 11

>0.242 11 3

<0.012 11 0.272 11

>0.260 11 4

<0.011 11 0.272 11

>0.261 11 5

<0.009 11 0.272 11

>0.263 11 For areas 1, 3, 4. and 5, the minimum plate thickness remains above the nominal plate thickness of 0.250 11

GNRO-2016/00062 For area 2, the minimum plate thickness is above the minimum thickness of 0.225" established in C-G-0130.0 Supplement 3.

All the noted indications are small localized rounded indications which do not result in a plate thickness below the allowable plate thicknesses in C-G-0130.0 Supplement 3. As a result, these areas are accepted by examination in accordance with ASME IWE-3122.1 (2001 Edition with 2003 Addenda).

Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

Suppression Pool Underwater Surfaces 1-FP-02F-2 Dent Indication. Random mech Report to Approved damage affecting substrate-30 Engineering Count-0.25in. dia. per Indication, In. x In. area. Metal loss 55 mils.

1-FP-08A-1 Other Indication. Random mech Report to Approved damage affecting substrate-1 Engineering Count-0.1251 n.x 21n. per Indication, In. x In. area. Metal loss <11mils 1-FP-08A-1 Dent Indication. Random mech Report to Approved damage affecting substrate-1 Engineering Count-0.251 n. dia. per Indication, In. x In. area. Metal loss <30mlls 1-FP-040-4 Other indication. Random mech Report to Approved damage affecting substrate-Engineering 0.06251 n. x 101n.per Indication, In. x in. area affected, 4 Per SqFt Metal loss <9mlls 1-WP-01C-3 Other Indication. Isolated mech Report to Approved damage affecting substrate-Engineering 0.25in dia. per Indication, 8ln.x 31n. area affected, 9 Per SqFt Metal loss <12mlls Containment Building Dome ISI-VT-16-074 Indications were previously Accepted Accepted by identified and evaluated per examination QIPN0411-000-2006. No chanQes ISI-VT-16-075 Indications were previously No changes Accepted by identified and evaluated per examination QIPN0411-000-2006.

ISI-VT-16-102 Indications were previously No changes Indications identified and evaluated per were not QIPN0411-000-2006.

observed.

ISI-VT-16-103 Indications were previously No changes Indications identified and evaluated per were not QIPN0411-000-2006.

observed.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

Containment Building Liner ISI-VT-16-060 Uncoated surface, 3 places, light Items were No additional rust on liner weld, no evidence of previously degradation pitting.

identified and noted during evaluated this Uncoated surface at liner weld.

acceptable.

examination.

Light rust with no evidence of pitting.

ISI-VT-16-061 Uncoated surface with light rust Items were No additional There is no evidence of pitting.

previously degradation identified and noted during Weld on liner is not coated. Rust evaluated this with no evidence of pitting.

acceptable.

examination.

Uncoated surface area approx.

3 10 11 x 1 10 11

• Has light rust with no evidence of pitting. Coating appears to have been removed by mech means.

Beam attachments are not accessible for visual inspection.

This applies only to bottom section.

ISI-VT-16-062 Uncoated surface on weld Items were No additional seams. Light corrosion with no previously degradation evidence of pitting.

identified and noted during evaluated this Uncoated surface with light acceptable.

examination.

corrosion. No evidence of pitting.

Uncoated surfaces with light to medium rust. This is mainly around the welded connections and surrounding area in the water box. There is no evidence of pitting.

Uncoated surfaces with light to medium rust (3 areas). There is no evidence of pitting.

Uncoated surface with light corrosion on embed plate. No evidence of pittinq.

ISI-VT-16-063 Coatings removed in this area.

Items were No additional

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

This area is where suppression previously degradation pool instrument cable was identified and noted during attached to liner. There are areas evaluated this of light rust that have no acceptable.

examination.

evidence of pitting.

ISI-VT-16-065 Coating removed at welds.

Items were No additional Welds have light rust.

previously degradation identified and noted during Coatings removed in these evaluated this areas. Zinc is in place.

acceptable.

examination.

Coatings removed in this area.

This area is where suppression pool instrument cable was attached to liner. Areas have light rust.

There is no evidence of pitting.

ISI-VT-16-067 Uncoated surface on liner weld.

Items were No additional Light rust with no evidence of previously degradation pitting.

identified and noted during evaluated this Uncoated surface. End of weld acceptable.

examination.

has light corrosion with no evidence of pitting.

ISI-VT-16-018 Uncoated surface with light rust.

Items were No additional There is no evidence of pitting.

previously degradation identified and noted during Uncoated surface on top half of evaluated this penetration. Zinc coating is acceptable.

examination.

intact. There is no evidence of pitting.

Uncoated surface where suppression pool instruments are installed. Light to medium rust in some areas with no evidence of pitting.

Uncoated surface with light rust.

No evidence of pitting.

ISI-VT-16-019 Uncoated surface. Zinc still in Items were No additional place. Surface is 1%" approx.

previously degradation identified and noted during Uncoated surface. Welds have evaluated this minor rust. No evidence of acceptable.

examination.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

pitting.

Visible rust streaks. The source is inaccessible for inspection.

Uncoated surface. Light rust at weld on embed plate. No evidence of pitting.

181-VT-16-020 Uncoated surface with no rust.

Items were No additional Area is aprox. 1/8" x 112".

previously degradation identified and noted during Uncoated surface with no rust.

evaluated this Area is aprox. 112" x 112".

acceptable.

examination.

Uncoated surface with no rust.

Area is aprox. 1112" x 112".

There is no indication of pitting on items shown above.

181-VT-16-021 Uncoated surface approx. %"

Items were No additional (chipped paint).

previously degradation identified and noted during Uncoated surface approx. 1Y2" X evaluated this 3-4" (chipped paint).

acceptable.

examination.

There is no indication of pitting in removal areas.

181-VT-16-022 Coatings removed in this area.

Items were No additional This area is where suppression previously degradation pool instrument cable was identified and noted during attached to liner. Rust in area.

evaluated this acceptable.

examination.

This area is used for storage during refuel outages. There are multiple areas where coatings have been removed (chipped, scratches). Light rust in area.

There is no indication of pitting in areas of liqht rust.

181-VT-16-023 Coatings removed in this area.

Items were No additional This area is where suppression previously degradation pool instrument cable was identified and noted during installed. Light rust in areas.

evaluated this acceptable.

examination.

Visible rust streaks. The source

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

is inaccessible for inspection.

Coating removed from 3" area.

Primer is still intact.

There is no indication of pitting in areas of light rust.

181-VT-16-024 Uncoated surfaces with light and Items were No additional medium rust where suppression previously degradation pool instrumentation cables are identified and noted during attached to liner. No evidence of evaluated this pitting.

acceptable.

examination.

181-VT-16-025 Uncoated surfaces with light and Items were No additional medium rust where suppression previously degradation pool instrumentation cables are identified and noted during installed. Rust in some areas evaluated this with no evidence of pitting.

acceptable.

examination.

Uncoated surface at end cap.

Light rust with no evidence of pitting.

181-VT-16-026 Uncoated surfaces around Items were No additional removal area on embed plate.

previously degradation There is light rust with no identified and noted during evidence of pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust on penetration cover. There is no evidence of pitting.

Uncoated surface with light rust.

Areas are aprox. 1". There is no evidence of pitting.

Uncoated surface with light rust.

These are small areas with no evidence of pitting (3 places).

Uncoated surface with light rust.

Area is aprox. 2" with no evidence of pitting.

Uncoated surface on weld. There is light to medium rust with no evidence of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

181-VT-16-027 Uncoated surface with light to Items were No additional medium rust. This area extends previously degradation to below the seam weld into the identified and noted during expansion joint at 16felev. This evaluated this area is inaccessible for acceptable.

examination.

examination. There is no indication of pitting.

Uncoated surface with light rust (2 places). This is no indication of pitting.

Uncoated surface with light rust.

This is no indication of pitting.

Uncoated surface with no rust or pitting. Paint in an area of approx. 4" x 2.5" has peeled.

Zinc is still intact.

Uncoated surface with no rust or pitting. Paint in an area of approx. 1.5" x 2.5" has peeled.

Zinc is still intact.

Uncoated surface with light rust.

Area is approx. 2" in length.

There is no indication of pitting.

Uncoated surface with no rust or pitting. There are scattered areas of chipped paint. Zinc is still intact.

Uncoated surface on portions of the weld. There is no rust or pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

ISI-VT-16-028 Uncoated surface on plate to Items were No additional liner weld. There is light rust with previously degradation no indication of pitting.

identified and noted during evaluated this Uncoated surface on seam weld.

acceptable.

examination.

There is light rust with no indication of pitting.

Item 3. Uncoated surface with no rust and no pitting. Paint has chipped away in an area of approx. 1 11 x 2 11

• Zinc coating is still intact.

Uncoated surface with no rust or pitting. Paint has flaked away in this area. Zinc coating is still intact.

Uncoated surface with no rust or pitting. There are numerous areas of chipped paint. Zinc coating is still intact.

Uncoated surface with light rust on the weld. There is no indication of pitting.

Uncoated surface on embed plate. There is light rust with no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

Uncoated surface on seam weld.

There is light rust with no indication of pitting.

Uncoated surface on the liner at the expansion joint. This does not extend behind the expansion joint. There is no evidence of pitting.

Uncoated surface with no rust or pitting. Paint has flaked away in an area approx. 1 11 x 1.5 11

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

181-VT-16-029 Uncoated surface with light rust.

Items were No additional There is no indication of pitting (2 previously degradation places).

identified and noted during evaluated this Light rust on weld. There is no acceptable.

examination.

indication of pittinq.

181-VT-16-030 Uncoated surface with light rust.

Items were No additional There is no evidence of pitting.

previously degradation identified and noted during Uncoated surface with light rust evaluated this around penetrations. There is not acceptable.

examination.

evidence of pitting.

Uncoated surface with medium to light rust (7 places). There is no evidence of pitting.

Uncoated surface with light rust.

Area is approx. 12" x 2". There is no evidence of pitting.

Uncoated surface with light rust.

Area is approx. 14" x 2". There is no evidence of pitting.

Uncoated surface with light rust.

There is no evidence of pitting.

Uncoated surface with light rust.

Area is approx. 4" x 12". There is no evidence of pitting.

Uncoated surface with light rust.

Area is approx. 2" diameter.

There is no evidence of pitting.

Uncoated surface with light rust.

Area is approx. 3" diameter.

There is no evidence of pitting.

Uncoated surface with light rust.

There is no evidence of pitting.

Uncoated surface with light rust at weld. There is no evidence of pittino.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

ISI-VT-16-031 Uncoated surface with light rust Items were No additional behind I-beam (3 places).

previously degradation identified and noted during Uncoated surface with light rust evaluated this at welds.

acceptable.

examination.

Uncoated surface with light rust approx. 3" sq.

Uncoated surface with light rust approx. 10" x 4".

Uncoated surface with light rust approx. 4" x 6".

Uncoated surface with light rust the full length of the plate. Paint has chipped and or peeled.

Bare metal with light rust. Areas are approx. 3" diameter (4 places).

Bare metal with light rust approx.

3" diameter.

Inaccessible areas behind vertical trays.

No evidence of pitting in areas identified as having light rust.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

ISI-VT-16-032 Uncoated surface at embed to Items were No additional liner weld. There is light rust with previously degradation no indication of pitting.

identified and noted during evaluated this Uncoated surface at eye bolt acceptable.

examination.

welds. There is light rust with no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 24" x 2". There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 3" x 4". There is no indication of pitting.

Uncoated surface with no rust and no pitting. Location is scattered with areas of chipped paint. Zinc is intact.

Uncoated surface with light rust on attachment. There is no indication of pitting.

There is a gouge approx. 3/4" x 1/4" x 1/32" deep. There is light rust with no indication of pitting.

There is a scratch approx. 3/8" x 1/4". There is light rust with no indication of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

ISI-VT-16-033 Uncoated surface with light rust.

Items were No additional Area is approx. 3 11 x 1 II. There is previously degradation no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

on weld. There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 10 11 X 4 11

• There is no indication of pitting.

Uncoated surface with light rust on penetration covers. There is no indication of pitting.

Uncoated surface with light rust on attachment welds (8 places).

There is no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

ISI-VT-16-034 Uncoated surface with light rust Items were No additional on penetration surface. There is previously degradation no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

on attachment welds. There is no indication of pitting.

Uncoated surface with light rust on penetration cover and attachment welds. There is no indication of pitting.

Uncoated surface with light rust on strain gauge (2 places). There is no indication of pitting.

Uncoated surface with light rust (3 places). Area is approx. 2 11 diameter. There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 3 1/2 11 X 1 1/2 11 There is no indication of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

Uncoated surface with light rust.

Area is approx. 4" x 2". There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 3" x 2". There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 5" x 4". There is no indication of pitting.

Uncoated surface with light rust (3 places). There is no indication of pitting.

Uncoated surface with light rust (10 places). There is no indication of pitting.

Uncoated surface. There is numerous areas of chipped and peeled paint. There is light rust in some areas with no indication of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

ISI-VT-16-035 Uncoated surface with light rust.

Items were No additional Area is approx. 4 11 x 3 11 (4 places).

previously degradation There is no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

on the penetration covers (3 places). There is no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

Discernable bulge with a diameter of 8 11

• This bulge extends outward approx. 1/2 11 There is no flaking or peeling of coatings in this area. UT thickness were performed in the area on and around the bulge.

There is no reduction of base material from that recorded on NDE Report BOP-UT-07-008.

Uncoated surface with light rust.

Area is approx. 1/2 11 x 4 11

• There is no indication of pitting.

Uncoated surface with light rust or pitting. Paint has been chipped but the zinc coating is intact.

Uncoated surface with light rust.

Area is approx. 4 11 diameter.

There is no indication of pitting.

Uncoated surface with light rust on the seam weld (2 places).

There is no indication of pitting.

Uncoated surface on the sway strut attachment weld. There is no indication of pitting.

Paint has flaked in area approx.

1/2 11 x 2 11

• Zinc coating is intact.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

181-VT-16-036 Uncoated surface at T.8.

Items were No additional attachment welds. There is light previously degradation rust with no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

on penetration cover. There is no indication of pitting.

Uncoated surface with light rust on penetration cover and pipe.

There is no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

181-VT-16-037 Uncoated surface at T.8.

Items were No additional attachment welds. There is light previously degradation rust with no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

on penetration cover. There is no indication of pitting.

Uncoated surface with light rust on penetration cover and pipe.

There is no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component 10 Indication Description Disposition Comments I Report No.

ISI-VT-16-037 Uncoated surface with light rust Items were No additional and no pitting on penetration previously degradation covers (5 pieces).

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

and no pitting on the attachment weld.

Uncoated surface with light rust and no pitting on the gauge mount.

Uncoated surface with light rust.

No indication of pitting.

Uncoated surface with light rust and no pitting on removal areas (3 places).

Uncoated surface with light rust and no pitting at seam weld. Area is approx. 6 11 in length.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component 10 Indication Description Disposition Comments I Report No.

181-VT-16-038 Uncoated surface with light rust Items were No additional where unistrut attaches to the previously degradation plate (6 places). No indication of identified and noted during pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust.

Area is approx. 10" x 4" (3 places). There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 6" x 4" (2 places).

There is no indication of pitting.

Uncoated surface with light rust on the penetration cover and pipe (2 places). There is no indication of pitting.

Uncoated surface at the strain gauge weld attachments (6 places). There is no indication of pitting.

Uncoated surface with light rust.

No indication of pitting.

Uncoated surface with light rust at the attachment weld. No indication of pitting.

Uncoated surface with light rust.

Area is approx. 1'1 x 3" (2 pieces).

No indication of pitting.

Uncoated surfaces approx. 6" x 3" (2 places). There is light rust with no indication of pitting.

Uncoated surfaces approx. 6" x 4" (2 places). There is light rust with no indication of pittinq.

181-VT-16-039 Uncoated surface on attachment Items were No additional weld at strain gauges (4 places).

previously degradation There is light rust with no

.identified and noted during indication of pitting.

evaluated this acceptable.

examination.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

Uncoated surface with light rust.

Area is approx. 14" x 8". There is no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

Uncoated surface with four areas with arc strikes. No indication of cracking or pitting.

Uncoated surface 360 0 around the equipment hatch flange. Light rust with no indication of pitting.

Uncoated surface with light rust.

Area is approx. 18" x 6". There is no indication of pitting.

ISI-VT-16-040 Uncoated surface on pipe behind Items were No additional penetration cover (2 places).

previously degradation There is light rust with no identified and noted during indication of pitting.

evaluated this acceptable.

examination.

Uncoated surface on attachment weld at strain gauge. There is light rust with no indication of pitting.

Uncoated surface on attachment weld. There is light rust with no indication of pitting.

ISI-VT-16-041 Uncoated surface at strain gauge Items were No additional weld attachment. There is light previously degradation rust with no indication of pitting.

identified and noted during evaluated this Scattered areas of chipped paint.

acceptable.

examination.

No rust. Zinc coating is intact.

Angle removal area with light rust. No indication of pitting.

ISI-VT-16-042 Uncoated surface with light rust Items were No additional (3 places). There is no indication previously degradation of pitting.

identified and noted during evaluated this Uncoated surface with strain acceptable.

examination.

gauge. There is medium rust with

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

no indication of pitting.

Uncoated surface at attachment weld. There is light rust with no indication of pitting.

Uncoated surface with light rust (4 places). There is no indication of pitting.

Uncoated surface with light rust (7 places). There is no indication of pitting.

Uncoated surface with no rust.

Area is an ark strike with no indication of cracks or pitting.

Uncoated surface with light rust.

There is no indication of pitting.

Gouge with a reduction of 0.060" from nominal wall. Nominal wall is 0.275 ". Reduction was caused from grinding.

Location of gouge is at 194 1 EL.

Gouge with a reduction of 0.046" from nominal wall. Nominal wall is 0.268 ". Reduction was caused from lug removal at location 186 1-5" EL.

Gouge with a reduction of 0.0625" from nominal wall.

Nominal wall is 0.268". Reduction was caused from lug removal at location 18T-5" EL.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

ISI-VT-16-043 Uncoated surface with light rust.

Items were No additional There is no indication of pitting.

previously degradation identified and noted during Uncoated surface with light rust evaluated this (10 places). There is no acceptable.

examination.

indication of pitting.

Uncoated surface with light rust.

Area is approx. 4" x 2".

Uncoated surface with no rust or pitting. There are numerous areas of chipped paint. Zinc coating is intact.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

ISI-VT-16-044 Gouge with reduction of 0.078" Items were No additional from nominal wall. Nominal wall previously degradation is 0.282". Area is approx. 1/2" identified and noted during diameter. Location of gouge is evaluated this 189 1-1 0 II at AZ-117°.

acceptable.

examination.

Uncoated surface with chipped paint in numerous areas. There is no rust and the zinc coating is intact.

Uncoated surface with light rust.

There is no indication of pitting.

Uncoated surface due to arc strikes (5 places). There is no indication of pitting.

Uncoated surface with light rust.

Area approx. 1" diameter. There is no indication of pitting.

Uncoated surface with medium rust (3 places). There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 2" diameter.

There is no indication of pitting.

Uncoated surface with light rust (5 places). There is no indication of pitting.

Uncoated surface with medium rust in an area approx. 12" square. There is no indication of pitting.

Uncoated surface with light rust in an area approx. 1" diameter.

There is no indication of pitting.

Uncoated surface with light rust (3 places). There is no indication of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component 10 Indication Description Disposition Comments I Report No.

ISI-VT-16-047 Uncoated surface at attachment Items were No additional with light rust (2 places). There is previously degradation no indication of pitting.

identified and noted during evaluated this Uncoated surface at penetration acceptable.

examination.

weld. There is light rust with no indication of pitting.

Uncoated surface with light rust on penetration (2 places). There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 8" x 5". There is no indication of pitting.

0, Uncoated surface with light rust.

Area is approx.12" x 5", There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 2" diameter (3 places). There is no indication of pitting.

Uncoated surface with light rust at strain gauge. There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 12" x 10". There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 12" x 10". There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 14" x 18". There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 12" x 4". There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 12" x 6". There is no indication of pitting.

Uncoated surface with light rust (9 places). There is no indication of oittino.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

181-VT-16-049 Uncoated surface with light rust Items were No additional and unistrut attachment (2 previously degradation places). There is no indication of identified and noted during pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust at strain gauge (2 places). There is no indication of pitting.

Uncoated surface with light rust at attachment weld. There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 3" diameter.

There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 4" diameter.

There is no indication of pitting.

Uncoated surface with medium rust at strain gauge. There is no indication of pitting.

181-VT-16-050 Uncoated surface with light rust Items were No additional at attachment weld. There is no previously degradation indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust.

acceptable.

examination.

Area isapprox. 2" diameter.

There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 1/2" diameter.

There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 1 1/2" x 1 ".

There is no indication of pitting.

181-VT-16-051 Uncoated surface with light rust Items were No additional at attachment weld (2 places).

previously degradation There is no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

(6 places). There is no indication of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

Uncoated surface with light rust.

Area is approx. 2 11 diameter.

There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 8 11 x 6 11

• There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 12 11 x 12 11

• There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 3 11 x 2 11

• There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 6 11 x 4 11

• There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 6 11 x 2 11

• There is no indication of pitting.

Uncoated surface with light rust (2 places). There is no indication of pitting.

Uncoated surface with light rust (2 places). There is no indication of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

181-VT-16-052 Uncoated surface with light rust.

Items were No additional Where structural steel welds to previously degradation embed plate. There is no identified and noted during indication of pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust.

Area is approx. 1/12 11 x 1/2 11 There is no indication of pitting.

Uncoated surface with no rust or pitting. Numerous areas of chipped paint. Zinc is intact.

Comments:

There are a few previously identified gouges in the lower liner plate. In 8/2003 a UT was performed to verify the thickness of the liner plate. The average thickness of the liner plate in this area is.291'1. Gouges do not exceed.037 11 181-VT-16-053 Uncoated surface with light rust Items were No additional where structural steel welds to previously degradation embed plate. There is no identified and noted during indication of pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust.

Area is approx. 1 1/2 11 X 1/2 11 There is no indication of pitting.

Uncoated surface with no rust or pitting. Numerous areas of chipped paint. Zinc is intact.

Uncoated surface with light rust (

2 places 4 11 x 8 11

). There is no indication of pitting.

Comments:

There are a few previously identified gouges in the lower liner plate. In 08/2003 a UT was performed to verify the thickness of the liner plate. The average thickness of the liner plate in this area is.297 11

• Gouges do not exceed 0.37 11

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

ISI-VT-16-054 Uncoated surface with light rust.

Items were No additional Area is approx. 1 11 diameter.

previously degradation There is no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust.

acceptable.

examination.

Area is approx. 2 11 x 6 11

• There is no indication of pitting.

Uncoated surface with medium rust at unistrut weld. There is no indication of pitting.

Uncoated surface with light rust.

There is no indication of pitting.

Uncoated surface with light rust on seam weld. Area is approx. 1 11 long. There is no indication of pitting.

Uncoated surface with no rust or pitting. Zinc is intact.

Uncoated surface with light rust.

Small areas of chipped paint.

There is no indication of pitting.

Uncoated surface with no rust or pitting. Zinc is intact.

Uncoated surface with no rust or pitting. Numerous areas of chipped paint. Zinc is intact.

Uncoated surface repaired area.

There is no indication of rust.

Comments:

There a few previously identified gouges in the lower liner plate.

The average thickness of the liner plate in this area is.291'1.

Gouqes do not exceed 0.37 11 ISI-VT-16-055 Uncoated surface with medium Items were No additional rust where structural steel welds previously degradation to embed plate (2 places). There identified and noted during is no indication of pitting.

evaluated this

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

acceptable.

examination.

Uncoated surface with light rust where structural steel weld to embed. There is no indication of pitting.

Uncoated surface with no rust or pitting. Numerous areas of chipped paint. Zinc is intact.

Comments:

There are a few previously identified gouges in the lower liner plate. On 08/2003 a UT was performed to verify the thickness of the liner plate. The average thickness of the liner plate in this area is.29T'. Gouges do not exceed 0.3T'.

181-VT-16-056 Uncoated surface with light rust Items were No additional at strain gauge (2 places). There previously degradation is no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

on embed plate (2 places). There is no indication of pitting.

Uncoated surface with no rust or pitting.

Uncoated surface with no rust or pitting.

There are numerous areas of chipped paint.

Uncoated surface behind T.8.

There is no rust or pitting.

Uncoated surface with bare metal (1 place - 1 1/2" X 3") with no rust or pitting.

Uncoated surface (bare metal) with no rust or pitting behind pipe support.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component 10 Indication Description Disposition Comments I Report No.

ISI-VT-16-057 Uncoated surface with light rust Items were No additional on embed plate. Area is approx.

previously degradation 3" x 1 ". There is no indication of identified and noted during pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust at T.S. to embed welds. There is no indication of pitting.

Uncoated surface with light rust.

Coating is cracked and peeling.

There is no indication of pitting.

Uncoated surface where paint is flaking and peeling. There is no pitting. Zinc is intact.

Uncoated surface where paint is peeling. There is no rust or pittinq. Zinc is intact.

ISI-VT-16-058 Uncoated surface with medium Items were No additional rust on embed plate weld to liner.

previously degradation Area is approx. 12" long. There is identified and noted during no indication of pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust.

Area is approx. 1" x 1/2". There is no indication of pitting.

Uncoated surface with light rust where structural steel welds to embed. There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 1" diameter.

There is no indication of pitting.

Uncoated surface with no rust or pitting. Numerous areas have chipped paint. Zinc coating is intact.

Uncoated surface with no rust or pitting. 'Numerous areas have chipped paint. Zinc coating is intact.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

Uncoated surface (1/2 11 x 12 11

)

with no rust or pitting. Zinc coating is intact.

ISI-VT-16-059 Uncoated surface with light rust Items were No additional where T.S. is welded to embed.

previously degradation There is no indication of pitting.

identified and noted during evaluated this Uncoated surface with light rust acceptable.

examination.

where structural steel welds to embed. There is no indication of pitting.

Uncoated surface with light rust where structural steel welds to embed. There is no indication of pitting.

Uncoated surface with light rust on embed plate. There is no indication of pitting.

Uncoated surface with light rust (5 places). There is no indication of pitting.

Uncoated surface with light rust (2 places). Areas are approx. 5" x 9". There is no indication of pitting.

Uncoated surface with no rust or pitting. There are numerous areas where paint has peeled or chipped. Zinc in intact.

Containment Dome Liner

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

181-VT-15-005 Uncoated surface with light rust Items were No additional on seam weld. Area is approx. 1 previously degradation 011 long. There is no indication of identified and noted during pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust Area is approx. 3 11 diameter.

There is no indication of pitting.

Uncoated surface with light rust Area is approx. 6 19 x 4 11

• There is no indication of pitting.

Uncoated surface with light rust Area is approx. 4 11 x 12 11

• There is no indication of pittinq.

181-VT-15-002 Uncoated surface with light rust Items were No additional Area is approx. 2 11 diameter (2 previously degradation places). There is no evidence of identified and noted during pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust Area is approx.11/2 11 diameter (2 places). There is no evidence of pitting.

Uncoated surface with light rust Area is approx.tz" x 3 11

• There is no evidence of pitting.

Uncoated surface with light rust Area is approx. 4 11 x 12 11

• There is no evidence of pitting.

Uncoated surface with light rust Area is approx. 2 11 diameter.

There is no evidence of pitting.

Uncoated surface with light rust.

Area is approx. 6 11 x 4 11

• There is no evidence of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component ID Indication Description Disposition Comments I Report No.

181-VT-15-003 Uncoated surface with light rust.

Items were No additional Area is approx. 2" diameter (2 previously degradation places). There is no indication of identified and noted during pitting.

evaluated this acceptable.

examination.

Uncoated surface with medium rust Area is approx. 4" x 1" diameter. There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 3" diameter.

There is no indication of pitting.

Uncoated surface with medium rust Area is approx. 6" x 1" diameter. There is no indication of pitting.

Uncoated surface with medium rust Area is approx. 1" diameter.

There is no indication of pittinq.

181-VT-15-004 Uncoated surface with light rust Items were No additional at plate to stiffener to embed previously degradation plate. There is no evidence of identified and noted during pitting.

evaluated this acceptable.

examination.

Uncoated surface with light rust at clevis to liner weld. Area is approx. 6" x 6" diameter. There is no evidence of pitting.

Uncoated surface with light rust.

Area is approx. 11/2" x 3/4".

There is no indication of pitting.

Uncoated surface with light rust.

Area is approx. 2" diameter.

There is no indication of pittinq, Lower Personnel Airlock

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component 10 Indication Description Disposition Comments I Report No.

181-VT-16-068 Reference Page 3 Items were No additional Uncoated surface with light rust.

previously degradation Area around angles are not identified and noted during painted. There is no indication of evaluated this pitting. (4 places) acceptable.

examination.

Uncoated surface with light rust.

There are numerous areas of chipped paint. There is no indication of pitting.

Uncoated surface with light rust.

Instrument tubing brackets are not painted. There is no indication of pitting.

Reference Page 4 Uncoated surface with light rust.

Paint remove from wear. There is no indication of pitting.

Uncoated surface with light rust.

There are several areas of chipped paint. There is no indication of pitting.

Uncoated surface with light rust on flange. There is no indication of pitting.

Uncoated surface with light rust.

Paint in these areas are chipped with no indication of pitting.

Uncoated surface with light rust on side of the door. There is no indication of pitting.

Uncoated surface (1 II x 4 11

) with light rust. There is no indication of pitting.

Uncoated surface with light rust on seal clamp, there is no indication of pitting.

Reference Page 5 Uncoated surface with light rust.

Area is approx. 6 11x6 11

• There is no indication of pitting.

Uncoated surface with liqht rust

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component 10 Indication Description Disposition Comments I Report No.

181-VT-16-068 Reference Page 6 Items were No additional Uncoated surface with light rust previously degradation on seal clamp. There is no identified and noted during indication of pitting.

evaluated this acceptable.

examination.

Reference Page 7 Uncoated surface with light rust on unistrut bracket. There is no indication of pitting.

Uncoated surface with light rust on instrument tubing bracket.

There is no indication of pitting.

Uncoated surface with light rust on seal clamp. There is no indication of oittinq.

Upper Personnel Airlock 181-VT-16-069 Reference Page 3 Items were No additional Uncoated surface with light rust previously degradation on angle clips. There is no identified and noted during indication of pitting. (4 places) evaluated this acceptable.

examination.

Uncoated surface with light rust on the left side facing containment. There is no indication of pitting.

Uncoated surface with light rust on instrument tubing brackets.

There is no indication of pitting.

Reference Page 4 Item 3. Uncoated surface with light rust. Paint removed from wear. There is no indication of pitting.

Uncoated surface with light rust on seal clamp. There is no indication of pitting.

Reference Page 5 Uncoated surface with light rust on instrument bracket. There is no indication of pitting.

GNRO-2016/00062 Table 3.3.9-2 RF20 Containment Visual Inspection Component 10 Indication Description Disposition Comments I Report No.

Uncoated surface with light rust on T.S. weld. There is no indication of pitting.

Uncoated surface with light rust.

Area is approx 5 1x5 1

• There is no indication of pitting.

Uncoated surface with light rust on seal clamp. There is no indication of pitting.

Reference Page 6 Uncoated surface. Paint removed from wear. There is no rust of pitting.

Uncoated surface with light rust on seal clamp. There is no indication of pitting.

Reference Page 6 Uncoated surface due to chipped paint. There is numerous areas.

There is no rust or pitting.

Uncoated surface with light to medium rust. There is no indication of pitting.

Uncoated surface with light rust on seal clamp. There is no indication of pitting.

3.4 NRC Information Notices (INs) 3.4.1 NRC Information Notice 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

GNRO-2016/00062 problem.

GGNS has only one bellows that may be subject to the failure mechanism described in this IN.

This is the expansion bellows (1 G41G515) associated with the horizontal fuel transfer tube (Containment Penetration No.4). GGNS conducted several tests to verify the adequacy of the local leak rate testing for this bellows and determined the following:

The bellows have been tested locally every refueling outage until the bellows were placed on an extended test frequency (currently 5 years).

The acceptance criteria are very low for this penetration (50 seem) and the tests have always demonstrated zero leakage.

During refueling outage 5 (1992), a visual inspection of the exterior surface of the bellows was done while under LLRT test pressure of 11.5 psig. No indications were found of cracks or gouges and the bellows were described as being in good condition.

Tests were done to verify that air could pass through each of the bellows halves from one test connection to the other and that there were no obstructions to the flow.

During refueling outage 6 (1993), tests were done to confirm that the bellows annulus was vented to the containment atmosphere. This ensured that the annulus was being subjected to ILRT test pressure (about 12 psig). This was the fourth ILRT with all results being well below the acceptance limits. In addition, a visual inspection using liquid leak detection fluid was done of the exterior of the bellows while attempting to pressurize the bellows with air.

This testing provides a high degree of confidence that the test methods currently being used are adequate to detect leakage across the bellows assembly. It is also worthwhile to note that the bellows are not subjected to large or rapid temperature changes or other operationally induced stresses.

3.4.2 Information Notice (IN) 2010-12, "Containment Liner Corrosion" The NRC issued this IN to inform addressees of issues concerning the degradation of the containment liner that could affect the leak-tightness of the containment structure.

IN 2010-12 described the degradation as follows:

Concrete reactor containments are typically lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions.

.., Operating experience shows that containment liner corrosion is often the result of liner plates being in contact with objects and materials that are lodged between or embedded in the containment concrete. Liner locations that are in contact with objects made of an organic material are susceptible to accelerated corrosion because organic materials can trap water that combined with oxygen will promote carbon steel corrosion. Organic materials can also cause a localized low pH area when they decompose. Organic materials located inside containment can come in contact with the containment liner and cause accelerated corrosion. However, corrosion that originates between the liner plate

GNRO-2016/00062 and concrete is a greater concern because visual examinations typically identify the corrosion only after it has significantly degraded the liner.

Based on the Operating Experience (OE) Evaluation, Grand Gulf is susceptible to the corrosion on the liner plates but Grand Gulf currently has barriers in place to minimize the likelihood of this event. Grand Gulf currently performs liner exams every inspection period and concrete exams every 5 years in accordance with CEP-CISI-102.

3.4.3 Information Notice (IN) 2014-07, "Degradation of Leak Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" The NRC issued this IN to inform addressees of issues 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 of the leak-chase channel.

IN 2014-07 describes operating experience that is concerned about the omission of code-required exams that were masked by other components and were therefore not included in the IWE database.

GGNS is not at risk as the leak chase system of the containment is included in the Containment Inservice Inspection (CISI) program.

No new actions were required to address this IN.

3.4.4 Regulatory Issue Summary 2016-07 "Containment Shell Or Liner Moisture Barrier Inspection" The U.S. Nuclear Regulatory Commission (NRC) is issuing this regulatory issue summary (RIS) to reiterate the NRC staff's position in regard to inservice inspection requirements for moisture barrier materials, as discussed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (hereinafter "the ASME Code"),Section XI, "Ru'les for Inservice Inspection of Nuclear Power Plant Components,"

Subsection IWE.

Section XI of ASME Code, Item E1.11, in Table IWE-2500-1 (E-A), requires general visual examination of 100 percent of accessible surface areas during each inspection period, while Item E1.30 in the same table requires general visual examination of 100 percent of accessible moisture barriers during each inspection period. Note 4 (Note 3 in editions before 2013) for Item E1.30 under the "Parts Examined" column states,

GNRO-2016/00062 "Examination shall include 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."

GGNS does not have a moisture barrier. GGNS is not at risk as the leak chase system of the containment is included in the Containment Inservice Inspection (CISI) program.

No new actions were required to address this RIS.

3.5 Plant-Specific Confirmatory Analysis The purpose of this analysis is to provide risk insights about extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) and Drywell Bypass Test (DWBT) interval by 18 months for Grand Gulf Nuclear Station. The extended test interval is a one cycle 18-month increase over the currently approved 10-year test interval. This translates to an extended test interval of 11.5 years. The extension would allow for substantial cost savings as the ILRT could be deferred to the next scheduled refueling outage for the Grand Gulf Nuclear Station (GGNS). The risk assessment follows the guidelines from NEI 94-01, Revision 3-A, the methodology used in EPRI TR-104285, 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, risk insights in support of a request for a plant's licensing basis as outlined in Regulatory Guide (RG) 1.174, 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, and the methodology used in EPRI 1018243, Revision 2-A of EPRI 1009325.

The findings of the GGNS risk assessment confirm the general findings of previous studies that the risk impact associated with extending the ILRT interval from 10 years to 11.5 years is "small." The GGNS plant-specific results for extending ILRT interval from the current 10 years to 11.5 years are summarized below:

Since the ILRT does not impact Core Damage Frequency (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 three in 10 years to one in 11.5 years is very conservatively estimated to be "very small."

An additional assessment of the impact from external events was also performed. In this sensitivity case, the change in the total GGNS LERF (including external events) was conservatively estimated to be "small."

The change in Type A test frequency to one per 11.5 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.0059 person-rem/year.

EPRI Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of

< 1.0 person-rem per year, or < 10/0 of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals.

Moreover, the risk impact when compared to other severe accident risks is "negligible."

GNRO-2016/00062 The increase in the conditional containment failure from the three in 10 year interval to one in 11.5 year interval is approximately 0.62%. EPRI Report No.

1009325, Revision 2-A, states that increases in CCFP of < 1.5 percentage points is very small. Therefore, this increase is judged to be livery small."

Therefore, increasing the ILRT interval to 11.5 years is considered to be insignificant since it represents a "small" change to the GGNS risk profile.

The NRC, in NUREG-1493, has previously concluded that:

Reducing the frequency of Type A tests (ILRTs) from three per 10 years to one per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Type B 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 one 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 GGNS confirm these general findings on a plant-specific basis considering the severe accidents evaluated for GGNS, the GGNS containment failure modes, and the local population surrounding GGNS.

The insights from this risk analysis support the deterministic analysis showing that there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner of this license request.

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.

The requirements to perform testing of the primary reactor containment are set forth in 10 CFR 50.54(0) and 10 CFR 50, Appendix J.

Both of these sections address criteria established in 10 CFR 50, Appendix A in General Design Criteria (GOC):

GOC 50 (Containment Design Basis); GOC 51 (Fracture Prevention of Containment Pressure Boundary); GOC 52 (Capability for Containment Leakage Rate Testing); and, GOC 53 (Provisions for Containment Testing and Inspection). A discussion of the GGNS conformance with these GOC is provided in the GGNS Updated Final Safety Analysis Report (UFSAR) Chapter 3.1.

Entergy has determined that the proposed change does not require any additional exemptions or relief from regulatory requirements and does not affect conformance with any GOC as described in the UFSAR. However, this change does propose a one cycle extension of the frequency for performance of the Type A Integrated Leakage Rate Test (ILRT) and the Orywell Bypass Leakage Rate Test (OWBT). The requirement to perform a drywell bypass leakage rate test is derived from 10 CFR 50.36.

GNRO-2016/00062 10 CFR 50.54(0) 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 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.

10 CFR 50.36(c)(3), "Surveillance requirements," states, in part, that TS shall include the "requirements relating to test, calibration or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met."

This proposed change makes an administrative change to TS 5.5.12 and SR 3.6.5.1.1, to add the date-related information for the next Type A test performance, along with the date related information for the next DWBT. Therefore, this 10 CFR 50.36 requirement continues to be met by this change.

10 CFR 50.36(c)(5), "Administrative controls," requires that "provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner" will be included in the TS.

10 CFR 50, Appendix J, Option B,Section V.B, "Implementation" requires that the implementation document used to develop a performance-based leakage testinq program be included by general reference in the TS. The Appendix J Testing Program is included in the Administrative Controls section of the GGNS TS, as TS 5.5.12, "10 CFR 50, Appendix J, Testing Program." This LAR does not remove this administrative control requirement, but simply revises the administrative controls TS 5.5.12, to include extending the frequency for performing the Type A ILRT from 10 years to 11.5 Years. In addition, the TS revision will revise SR 3.6.5.1.1 to incorporate the extension of the 10 year (120 month) DWBT to 11.5 Years. Therefore, this 10 CFR 50.36 requirement continues to be met by this change.

4.2 Precedent This license amendment request is similar in nature to the following license amendments previously approved by the NRC to extend the Type A test frequency:

December 29, 1994 (ML011080782), for Nine Mile Point Nuclear Station Unit 1, June 2,2003 (ML031320686), for Vermont Yankee Nuclear Power Station, July 20, 2009 (ML091540158) for Arkansas Nuclear One, Unit No.2, August 23, 2010 (ML102090137) for Palisades Nuclear Plant.

GNRO-2016/00062 October 1, 2012 (ML12250A339) for Oconee Nuclear Station Unit 1.

August 5,2013 (ML13193A329) for Oconee Nuclear Station Units 2 and 3.

September 26, 2016 (ML16236A053) for McGuire Nuclear Station Units 1 and 2.

4.3 Significant Hazards Consideration Entergy Operations, Inc. (Entergy) 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 Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years.

The proposed extension does 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.

Type Band C testing ensures that individual containment isolation valves are essentially leak tight.

In addition, aggregate Type Band C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The assessment of the leaktightness of the drywell will continue to be performed at least once each operating cycle. The proposed amendment will not change the leakage rate acceptance requirements.

As such, the containment will continue to perform its design function as a barrier to fission product releases. In addition, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment and the assessment of the leaktightness of the drywell 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.

Therefore, this proposed extension does not involve a significant increase in the probability of an accident previously evaluated.

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 Technical Specifications (TS) involves the extension of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years.

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.

GNRO-2016/00062 The proposed change does not involve a physical change to the plant (Le., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled.

.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 the Technical Specifications (TS) involves the extension of the Grand Gulf Nuclear Station, Unit 1 (GGNS) Type A integrated leakage rate test and the drywell bypass leakage rate test intervals to 11.5 years.

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 10 CFR 50, Appendix J, Testing Program for containment leak rate testing 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 the extension of the interval for only the Type A containment leakage rate test and the drywell bypass leakage rate test for GGNS.

The proposed surveillance interval extension is bounded by the 15-year Type A test interval currently authorized within NEI 94-01, Revision 3-A. 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 the proposed changes to the Type A test interval. In addition to the scheduled performance of DWBT GGNS will continue to monitor the drywell for significant leakage during operation.

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

Based on the above, Entergy 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.

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.

5.0 ENVIRONMENTAL CONSIDERATION

GNRO-2016/00062 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

1.

Regulatory Guide 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.

NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, July 1995.

5.

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

6.

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

7.

Letter from M. J. Maxin (NRC) to J. C. Butler (NEI), Final Safety Evaluation for NEI Topical Report (TR) 94-01, Revision 2, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated June 25,2008; and EPRI Report No.1 009325, Revision 2, August 2007, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (TAC No. MC9663)

(ML081140105).

8.

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 (ML121030286).

9.

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

10. Letter from P. W. O'Connor (NRC) to C. R. Hutchinson (EOI), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Exemption from the Requirements of 10 CFR Part 50, Appendix J, (TAC No. 87209)," dated April 26, 1995 (GNRI-95/00087)

(ML021480397).

GNRO-2016/00062

11. Letter from J. N. Donohew (NRC) to C. R. Hutchinson (EOI), "Issuance of Amendment No. 126 to Facility Operating License No. NPF Grand Gulf Nuclear Station, Unit 1 (TAC No. M94176)," dated August 1, 1996 (GNRI-96/00162)

(ML021480466 & ML021490103).

12. Letter from J. N. Donohew (NRC) to J. J. Hagan (GGNS-EOI), "Issuance of Amendment No. 128 to Facility Operating License No. NPF Grand Gulf Nuclear Station, Unit 1 (TAC No. M95338)," dated 'October 18, 1996 (GNRI-96/00212)

(ML021490101).

13. Letter from J. N. Donohew (NRC) to J. J. Hagan (GGNS-EOI), "Issuance of Amendment No. 135 to Facility Operating License No. NPF Grand Gulf Nuclear Station, Unit 1 (TAC No. M99879)," dated April 6, 1998 (GNRI-98/00028)

(ML021490221 ).

14. Letter from J. N. Donohew (NRC) to W. A. Eaton (GGNS-EOI), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment No. 145, Re: Full-Scope Implementation of an Alternative Accident Source Term (TAC No. MA8065)," dated March 14, 2001 (GNRI-2001/00032) (ML010780172).

15. Letter from B. Vaidya (NRC) to G. A. Williams (GGNS-EOI), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment No. 164, Re: One-Time Extension of the Integrated Leak Rate Test and Drywell Bypass Test Interval (TAC No. MB8940),"

dated January 28, 2004 (GNRI-2004/00013) (ML040300152).

16. Letter from B Vaidya (NRC) to G. A. Williams (GGNS-EOI), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment No. 168, Re: Containment Air Lock Leak Rate Test Acceptance Criteria (TAC No. MC5539)," dated July 12, 2005 (GNRI-2005/00057) (ML051540277).

17. Letter from B Vaidya (NRC) to W. R. Brian (GGNS-EOI), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment No. 176, Re: Change to Technical Specifications to Allow Certain Types of Relief Valves to be Used as Isolation Devices (TAC No. MD4676)," dated August 24, 2007 (GNRI-2007/001 01)

(ML072140501 ).

18. Letter from A. B. Wang (NRC) to GGNS-EOI, "Grand Gulf Nuclear Station, Unit 1 -

Issuance of Amendment [No. 191] Re: Extended Power Uprate (TAC No. ME4679),"

dated July 18,2012 (GNRI-2012/00153) (ML121210020).

19. Letter from A. B. Wang (NRC) to C. R. Hutchinson (EOI), "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment No. 197, Re: Revise Technical Specification Surveillance Requirement Frequencies from 18-to 24-Month Fuel Cycle Intervals (TAC No. ME9764)," dated December 26,2013 (GNRI-2013/00187)

(ML13343A109).

20. Letter from J. Kim (NRC) to Entergy Operations (GGNS), "Grand Gulf Nuclear Station, Unit 1 - Issuance Of Amendment Re: Revision Of Technical Specifications For Containment Leak Rate Testing (CAC No. MF631 0)," dated February 17, 2016 (GNRI-2016/00020)

GNRO-2016/00062

21. ML102090137, Letter from M. Chawla (NRC) to Vice President, Operations (Entergy) dated August 23, 2010. Palisades Nuclear Plant - Issuance of Amendment Re: One-Time Extension to the Integrated Leak Rate Test Interval (TAC NO. ME2122).

22. ML13193A329, Letter from J. Boska (NRC) to S. Batson (Duke) dated August 5, 2013. Oconee Nuclear Station, Units 2 And 3, Issuance of Amendments Regarding Extension of the Reactor Building Integrated Leak Rate Test (TAC NOS. ME9777 AND ME9778).

23. ML12250A339, Letter from J. Boska (NRC) to P. Gillespie (Duke) dated October 1, 2012. Oconee Nuclear Station, Unit 1, Issuance of Amendment Regarding Extension of the Reactor Building Integrated Leak Rate Test (TAC NO. ME8407).

24. ML011080782, Letter from D. Brinkman (NRC) to R. Sylvia (Niagara Mohawk) dated December 29, 1994. Issuance of Amendment For Nine Mile Point Nuclear Station Unit No.1 (TAC NO. M90278).

24. ML031320686, Letter from R. Pulsifer (NRC) to J. Thayer (Entergy) dated June 2, 2003. Vermont Yankee Nuclear Power Station - Issuance of Amendment Re: One-Time Extension of Appendix J Type A Integrated Leakage Rate Test Interval (TAC NO. MB6507).

25. ML091540158, Letter from N. Kalyanam (NRC) to Vice President, Operations (Entergy), dated July 20, 2009. Arkansas Nuclear One, Unit NO.2 - Issuance of Amendment Re: One-Time Extension to 10-Year Frequency of Integrated Leak Rate Test (TAC NO. MD9502).

26. ML16236A053, Letter from G. Miller (NRC) to S. Capps (McGuire), dated September 26, 2016. McGuire Nuclear Station, Units 1 and 2 - Issuance of Amendment Re:

One-Time Extension of Appendix J Type A Integrated Leakage Rate Test Interval (CAC Nos. MF7407 and MF7408).

27. Letter from A. Wang (NRC) to Vice President, Operations (Entergy), dated August 31, 2015. Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment Re:

Maximum Extended Load Line Limit Analysis Plus (TAC No. MF2798).

to GNRO-2016/00062 Proposed Technical Specification Changes (Marked-Up)

Drywell 3.6.5.1 SURVEILLANCE REQUIREMENTS SR 3.6. 5. 1. 1 SURVEILLANCE Verify bypass leakage is less than or equal to the bypass leakage limit.

However, during the first unit startup following drywell bypass leak rate testing performed in accordance with this SR, the acceptance criterion is leakage 5 10% of the bypass leakage limit.

FREQUENCY 24 months following 2 consecutive tests with bypass leakage greater than the bypass leakage limit until 2 consecutive tests are less than or equal to the bypass leakage limit 48 months following a test with bypass leakage greater than the bypass leakage 1i mi t

---

NOTE-----

SR 3.0.2 is not applicable for extensions> 12 months.

120 mo~

~ (continued)

, except that the next drywell bypass leak rate test performed after the October 19, 2008 test shall be performed no later than the plant restart after the End of Cycle 22 Refueling Outage.

GRAND GULF 3.6-53 Amendment No. ~, 164. ~

WWit~ (orrection letter of 9/16/96

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 5.5.12 Technical Specifications (TS)

Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1.

A change in the TS incorporated in the license; or 2.

A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain prOV1Slons to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes that do not meet the criteria of either Specification 5.5.1l.b.1 or Specification 5.5.11.b.2 above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

10 CFR 50, Appendix J, Testing Program This program establishes the leakage rate testing program of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be implemented in accordance with the Safety Evaluation issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087) as modified by the Safety Evaluation issued for Amendment No.

135 to the Operating License~~

For Type B and Type C local leakage rate testing, this program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," date July 2012.

Consistent with standard scheduling practices for Technical Specifications required surveillances, intervals for the reco~~ended surveillance frequency for Type A testing may be extended by up to 25 percent of the test interval, not to exceed 15 months.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 12.1 psig.

I except that the next Type A test performed after the October 19, 2008 Type A test shall be performed no later than the plant restart after the End of Cycle 1_

22 Refueling Outage.

GRAND GULF 5.0-16 Amendment No.~, ~ +/-9+, ~~

to GNRO-2016/00062 Revised Technical Specification Pages (Clean)

SURVEILLANCE REQUIREMENTS SURVEILLANCE Drywell 3.6.5.1 FREQUENCY SR 3.6.5.1.1 Verify bypass leakage is less than or equal to the bypass leakage limit However, during the first unit startup following drywell bypass leak rate testing performed in accordance with this SR, the acceptance criterion is leakage

~ 10% of the bypass leakage limit 24 months following 2 consecutive tests with bypass leakage greater than the bypass leakage limit until 2 consecutive tests are less than or equal to the bypass leakage limit 48 months following a test with bypass leakage greater than the bypass leakage limit

---

NOTE-----

SR 3.0.2 is not applicable for extensions> 12 months.

120 months, except that the next drywell bypass leak rate test performed after the October 19, 2008 test shall be performed no later than the plant restart after the end of Cycle 22 Refueling Outage.

(continued)

GRAND GULF 3.6-53 Amendment No. ~*, 164.~,_

• \\Vi~h COffee~ioA Letter of 9116/96

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.11 5.5.12 Technical Specifications (TS)

Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a.

Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b.

Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1.

A change in the TS incorporated in the license; or 2.

A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d.

Proposed changes that do not meet the criteria of either Specification 5.5.11.b.l or Specification 5.5.11.b.2 above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

10 CFR 50, Appendix J, Testing Program This program establishes the leakage rate testing program of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions.

This program shall be implemented in accordance with the Safety Evaluation issued by the Office of Nuclear Reactor Regulation dated April 26, 1995 (GNRI-95/00087) as modified by the Safety Evaluation issued for Amendment No.

135 to the Operating License, except that the next Type A test performed after the October 19, 2008 test shall be performed no later than the plant restart after the end of Cycle 22 Refueling Outage.

For Type B and Type C local leakage rate testing, this program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012.

Consistent with standard scheduling practices for Technical Specifications required surveillances, intervals for the recommended surveillance frequency for Type A testing may be extended by up to 25 percent of the test interval, not to exceed 15 months.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 12.1 psig.

GRAND GULF 5.0-16 Amendment No.~, ~ +/-9+/-,

205,209,