NL-16-0717, ISI Program Alternative HNP-ISI-ALT-05-02

ISI Program Alternative HNP-ISI-ALT-05-02

ML16127A191
Person / Time
Site:	Hatch
Issue date:	05/06/2016
From:	Pierce C Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF7692, HNP-ISI-ALT-05-02, NL-16-0717
Download: ML16127A191 (8)

Text

Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.

40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 Tel 205.992.7872 SOUTHERN ._\

Fax 205.992.7601 NUCLEA May 6, 2016 A SOUTHERN COMPANY Docket Nos.: 50-366 NL-16-0717 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant lSI Program Alternative HNP-ISI-ALT-05-02 Ladies and Gentlemen:

It has recently been determined that several safety relief valves (SRVs) removed from Plant Hatch Unit 1 exhibited unsatisfactory post-installation inspection results. Based on these results, Plant Hatch has conservatively decided to replace the Unit 2 SRV main valve bodies with components which have undergone augmented testing following the 2015 Target Rock Part 21 recommendations. This replacement will be performed during a Plant Hatch Unit 2 maintenance shutdown that has been scheduled to begin on May 20, 2016 and will result in the need to perform system leakage testing in accordance with certain provisions of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code).

Southern Nuclear Operating Company (SNC) had previously requested approval of an alternative to the requirements of the ASME Code Section XI, 2001 Edition through 2003 Addenda, Subsection IWB-5221 (a) in our letter NL-11-2498 dated December 13, 2011 . Specifically, SNC requested NRC approval of proposed Alternative HNP-ISI-ALT-15, Version 2, to perform the VT-2 visual examination during a system leakage test of Class 1 components with mechanical joint connections at a pressure lower than the Code-required pressure following repair and replacement activities on several SRVs installed in Plant Hatch Unit 2.

The 2001 Edition through 2003 Addenda of the ASME Code,Section XI requires the performance of a system pressure test in accordance with Section IWB-5220 for ASME Code Class 1 pressure boundary components prior to plant startup following each reactor outage. Paragraph IWB-5221 (a) requires that the system leakage test be conducted at a test pressure not less than the nominal operating pressure associated with 100% rated reactor power, which for Plant Hatch is 1045 psi g. These pressure test requirements are supplemented by 10 CFR50.55a(b)(2)(xxvi) which invokes the requirements of IWA-4540(c) of the 1998 Edition of the ASME Section XI Code for repair/replacement activities of Class 1, 2, and 3 mechanical joint connections.

U.S. Nuclear Regulatory Commission NL-16-0717 Page2 Pursuant to 10 CFR 50.55a(a)(3)(ii), SNC hereby requests approval of an alternative to the requirements of the ASME Code Section XI, 2001 Edition through 2003 Addenda, Subsection IWB-5221 (a). Specifically, SNC requests NRC approval of proposed Alternative HNP-ISI-ALT-05-02, to perform the VT-2 visual examination during a system leakage test of Class 1 components with mechanical joint connections at a pressure lower than the Code-required pressure following repair and replacement activities. This proposed alternative would allow the performance of the VT-2 visual leakage examination following the SRV repair and replacement activities at the lower pressure of greater or equal to 920 psig while employing a one-hour hold-time for non-insulated components and a six hour hold time for insulated components.

10 CFR 50.55a(a)(3)(ii) provides for the use of such alternatives to the requirements of paragraph 50.55a(g) when so authorized by the NRC. Such authorization is contingent upon a demonstration by the applicant that the proposed alternative would provide an acceptable level of quality and safety or that the specified requirements would result in hardship or unusual difficult without a commensurate increase in the level of quality and safety. It can be noted that this reduced reactor system pressure will result in more tolerable ambient temperature conditions for examination personnel and therefore should facilitate a higher quality examination. This alternative is therefore justified since compliance with the cited requirements of the subject code would result in a plant hardship without a commensurate increase in the level of quality and safety of the associated maintenance activity. Further details of the demonstration required by 10 CFR 50.55a(a)(3)(ii) are provided in Attachment 1 to this letter.

The NRC has previously approved several similar relief requests for performing pressure tests at less than nominal operating pressure including SNC's request for approval of lSI Program Alternative HNP-ISI-ALT -15, version 2 transmitted by letter NL-11-2498 dated December 13, 2011.

Expedited approval of the proposed alternative i,s requested on or before May 20, 2016, to allow the use of this proposed alternative for the planned SRV replacer'llents scheduled to be performed during the Plant Hatch Unit 2 May 2016 maintenance shutdown.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

Respectfully submitted, C/f~

C. R. Pierce Regulatory Affairs Director CRP/cdp

U.S. Nuclear Regulatory Commission NL-16-0717 Page2 : Request for lSI Program Alternative HNP-ISI-ALT-05-02 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Best, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President- Hatch Mr. M. D. Meier, Vice President - Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Mr. G. L. Johnson, Regulatory Affairs Manager- Hatch RTYPE: CHA02.004 U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager- Hatch Mr. D. H. Hardage, Senior Resident Inspector- Hatch

Edwin I. Hatch Nuclear Plant Enclosure 1 to NL-16*0717 HNP-ISI*ALT-05*02 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii)

SOUTH ERN NUCLE AR OPERA TING COMPA NY HNP-ISI-ALT-05-02, VERSIO N 1.0 PROPOS ED ALTERN ATIVE IN ACCOR DANCE WITH 10 CFR 50.55a(a)(3)(ii)

Plant Site-Unit: Edwin I. Hatch Nuclear Plant - Unit 2 Interval-Interval Dates: 5th lSI Interval, January 1, 2016 through December 31, 2025 Requested Date for Approval is requested by May 20, 2016 to support testing during a Unit 2 Ap]!l'oval: maintenance shutdown that is currently scheduled to begin on May 20,2016.

ASMECo de Components Class 1 pressure-retaining mechanical joint connections which require a VT-2 Affected: examination for leakage subsequent to repair/replacement activities.

Applicable Code Edition ASME Section XI Code, 2001 Edition through the 2003 Addenda and Addenda:

1. IWA-4540(a) requires a hydrostatic or system leakage test, in accordance with IWA-5000, for repair/replacement activities performed by welding or brazing on a pressure-retaining boundary prior to, or as part of, returning to service.

Applicable Code

2. IWA-5213(b) requires a 10 minute hold time for non-insulated components and Requirements: 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time for insulated components prior to performing the VT-2 leakage test.

3. IWB-5221(a) requires the system leakage test to be conducted at a pressure not less than the nominal pressure associated with 100% rated reactor power.

10CFR50.55a(b)(2)(xxvi) Pressure Testing Class 1, 2, and 3 Mechanical Joints provides supplemental code requirements to those of IWA-4540(a) stated above.

10CFR50.55a(b)(2)(xxvi) invokes the IWA-4540(c) repair/replacement activity provisions ofthe 1998 Edition of Section XI for pressure testing of Class 1, 2, and 3 mechanical joints when using the 2001 Edition through the latest edition and addenda of ASME Section XI. Therefore, even though the lSI Code of Record applicable at Plant Hatch does not require pressure testing and VT-2 examination of mechanical joint connections, the 1998 Edition of Section XI does.

Relief is requested from the test pressure requirement of IWB-5221(a) (i.e., 1045 psig) on the basis of hardship as cited below.

• Replacement of some components installed via mechanical joints (e.g., Safety Reason for Relief Valves (SRVs)) is planned during a maintenance shutdown which is Request: scheduled to begin May 20, 2016. These repair/replacement activities will require a VT-2 leakage examination of the mechanical joint connections during unit startup.

• Nominal operation pressure (i.e., 1045 psig) will not be reached for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 920 psig during the startup sequence:

0 Control Rod Drive withdrawal limitations and the associated gradual increases in reactor power, pressure and temperature.

0 Technical Specification-required Pressure versus Temperature limitations.

0 Main Steam line piping, turbine control and stop valve warming requirements.

0 Main turbine warming requirements.

0 Small increases in pressure over time to provide better seating

SOUTH ERN NUCLE AR OPERA TING COMPA NY HNP-ISI-ALT-05-02, VERSI ON 1.0 PROPO SED ALTER NATIV E IN ACCOR DANCE WITH 10 CFR S0.5Sa(

a)(3)(ii) characteristics ofthe SRVs.

• VT-2 leakage examination inside the drywell (primary containment) represents a hardship at the nominal operating pressure of 1045 psig during start-up because of high ambient and component temperatures.

o Data was retrieved for the two most recent startups (9/2013, 3/20 15) using instrumentation approximately 8ft higher in elevation than the SRVs.

• Ambient temperature was approximately 144 degrees Fahrenheit once reaching 920 psig.

• Data shows ambient temperature increases to approximately 150 degrees Fahrenheit over a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> period while holding pressure steady at 920 psig.

o The local drywell temperature seen once nominal reactor pressure of 1045 psig is attained is approximately 170 degrees Fahrenheit or higher.

• Reactor coolant system nominal operating pressure results in drywell ambient temperatures that require special safety precautions such as ice vests and cool air supply lines for personnel performing the VT-2 examinations.

• These adverse conditions could also compromise the quality of the leakage examination due to the hardship imposed on examination personnel.

• Performance of a cold leakage test (that is, a non-nuclear heat-up such as that required following a refueling outage) subsequent to a maintenance shutdown is judged to be an imprudent course of action for the reasons described below:

o Main Steam Lines are flooded with Main Steam Isolation Valves closed.

o The reactor pressure vessel (RPV) is required to be virtually water solid.

o Extensive valve manipulations, system lineups, and procedural controls are required in order to heat up and pressurize the reactor coolant system to establish the necessary test pressure.

o The additional valve lineups and system reconfigurations necessary to support this test will impose an additional challenge to the affected systems. A normal plant startup would then occur, after completion and subsequent recovery from the cold leakage test.

o Performing a cold leakage test would add approximately 2 days to the shutdown duration.

o Performance of an additional cold leakage test places the unit in a position of reduced margin, unnecessarily approaching the fracture toughness limits defined in the Technical Specification Pressure-Temperature (P-T) curves.

o The scope ofthe VT-2 leakage examination does not include the reactor pressure vessel.

Plant Hatch will perform the required VT-2 leakage examination for any repair/

replacement activities of mechanical joint connections performed during the May 2016 maintenance shutdown at a reactor pressure of~ 920 psig. Similar Proposed to Alternative previous similar submittals, SNC agrees to implement a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> hold time for non-and Basis for insulated components and a 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> hold time for insulated components prior to Use: performing the VT-2 Leakage Test. In addition, if there are unplanned shutdow ns with drywell entries before the next refueling outage (currently scheduled to begin in February 2018), inspections of the affected mechanical joint connections will be performed at reactor system pressure of> 920 psig to look for any evidence of

SOUT HERN NUCL EAR OPER ATING COMP ANY HNP-I SI-ALT 02, VERS ION 1.0 PROPO SED ALTE RNAT IVE IN ACCO RDAN CE WITH 10 CFR 50.55a(a)(3)(ii) leakage.

Disposition of any observed leakage will consider the marginal increase in leakage rates that might occur at the nominal operating pressure associa ted with I 00% rated reactor power (i.e., I 045 psig) and the actual reactor pressur e when the examination was performed.

In addition, drywell monitoring systems would detect leakage that might occur in mechanical joint connections at higher pressures associated with nomina l reactor operation. These systems include drywell air temperature and pressure monitoring and the drywell floor and equipment drain sumps.

Since the reactor coolant system pressure boundary is subjected to a leakage test and visual examination at nominal operating pressure (i.e., I 045 psig) near the end of every refueling outage and monitoring systems detect leakage inside the drywell, a leakage test and visual examination performed at 920 psig for the repair/replacement of mechanical joint connections provide adequate assurance of structural and pressure boundary integrity. Therefore, this proposed alternative should be granted pursuant to IO CFR 50.55a(a)(3)(ii).

This alternative is essentially the same as Alternatives HNP-ISI-ALT

-09, Version 2.0, and HNP-ISI-ALT-I5, Version 2.0, both previously approved by the NRC for Hatch Unit 2 (References 5 and 6, respectively).

Duratio n of The Hatch Unit 2 maintenance shutdown currently scheduled to begin on May 20, Propos ed 20I6, with the alternative continuing in effect through the start Alternative: of the next refueling outage currently scheduled to begin in February 20I8.

1. Entergy Nuclear Northeast, Pilgrim Nuclear Power Station, 4th IO-Yea r Interval lSI Program Relief Request PRR-2, NRC T AC NO. MC8286 dated June 29, 2006.

2. Nuclear Management Company, Monticello Nuclear Generating Plant, 3"' IO-Year Interval lSI Program Relief Request RR-I7, NRC TAC NO.

MC0593 dated March 25, 2004.

3. PSEG Nuclear, LLC, Hope Creek Nuclear Generating Station

, 2nd IO-Year Interval lSI Program Relief Request HC-RR-I2-023, NRC TAC No.

MC2396 dated August 27, 2004.

4. Nebraska Public Power District, Cooper Nuclear Station, 3"' I 0-Year Interval

References:

lSI Program Relief Request PR-I 0, NRC TAC No. MA0677 dated Februa ry 26, I998.

5. Southern Nuclear Operating Company, Inc., Edwin I. Hatch Plant Unit 2, lSI Program Alternative HNP-ISI-ALT-09 Version 2.0, March 29, 20IO (NRC ADAMS Accession No. ML 10089005I ).

6. Southern Nuclear Operating Company, Inc., Edwin I. Hatch Plant Unit 2, lSI Program Alternative HNP-ISI-ALT-15 Version 2.0, December 13, 20II (NRC ADAMS Accession No. ML II3480 294).

7. NextEra Energy Duane Arnold, LLC, Duane Arnold Energy Center, Request for Authorization of Alternative Regarding Pressure Test Requirements, NRC TAC No. ME5I4 3 dated September 6, 20Il.

SOUTH ERN NUCLE AR OPERA TING COMPA NY HNP-ISI-ALT-OS-02, VERSIO N 1.0 PROPO SED ALTERN ATIVE IN ACCOR DANCE Wim 10 CFR SO.SSa(a)(3)(ii)

Status: Awaiting NRC approval.