ET 17-0007, CFR 50.55a Request 14R-05 for the Fourth Inservice Inspection Program Interval

CFR 50.55a Request 14R-05 for the Fourth Inservice Inspection Program Interval

ML17111A865
Person / Time
Site:	Wolf Creek
Issue date:	04/13/2017
From:	Mccoy J Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ET 17-0007
Download: ML17111A865 (8)

Text

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W8LFCREEK -'NUCLEAR OPERATING CORPORATION Jaime H. McCoy Vice President Engineering April 13, 2017 ET 17-0007 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: 10 CFR 50.55a Request 14R-05 for the Fourth lnservice Inspection Program Interval To Whom It May Concern:

Pursuant to 10 CFR 50.55a(z)(2), Wolf Creek Nuclear Operating Corporation (WCNOC) hereby requests Nuclear Regulatory Commission (NRC) approval of 10 CFR 50.55a Request Number 14R-05 for the Fourth Ten-Year Interval of WCNOC's lnservice Inspection (ISi) Program. The Attachment provides 10 CFR 50.55a Request 14R-05, which requests relief from the pressure test requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, IWC-5220.

Performance of the proposed alternative pressure testing of this Request will be performed in Refueling Outage 22, which is scheduled to begin March of 2018. Therefore approval is requested prior to this date.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4156, or Cynthia R. Hafenstine (620) 364-4204.

Sincerely, 1~°;1/'JJ~

Jaime H. McCoy JHM/rlt Attachment cc: K. M. Kennedy (NRC), w/a B. K. Singal (NRC), w/a N. H. Taylor (NRC), w/a Senior Resident Inspector (NRC), w/a P.O. Box 411 I Burlington, KS 66839 /Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET

Attachment to ET 17-0007 Page 1 of 7 Wolf Creek Nuclear Operating Corporation 10 CFR 50.55a Request 14R-05 Request for Relief from the Pressure Test Requirements of ASME Section XI IWC-5220

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Attachment to ET 17-0007 Page 2 of 7 10 CFR 50.55a Request 14R-05 Request for Relief from the Pressure Test Requirements of ASME Section XI IWC-5220 - -

Proposed Alternative in Accordance with 10 CFR 50.55a(z)(2)

Hardship Without Compensating Increase in Quality and Safety

1.0 ASME Code Components Affected

The affected components are the Wolf Creek Generating Station (WCGS) piping and components in the reactor vessel flange leak-off lines connected to the reactor vessel running to isolation valve BBHV8032. The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)Section XI (Reference 1) examination category and item number covering the required examination and testing are IWC-2500-1, Examination Category C-H, Item No. C7.10. The piping and components in the reactor vessel flange leak-off lines are ASME Code Class 2 and are constructed of Type 304 stainless steel.

A marked-up drawing that shows the configuration of piping is included as Figure 1 on Page 7 of this Request. The components referred to in this Request are also identified in Figure 1.

2.0 Applicable Code Edition and Addenda ASME Code Section XI, 2007 Edition through 2008 Addenda 3.0 Applicable Code Requirement ASME Code Section XI, IWC-5221, states "The system leakage test shall be conducted at a system pressure obtained while the system, or portion of the system, is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability (e.g., to demonstrate system safety function or satisfy technical specification surveillance requirements)."

ASME Code Section XI, IWC-5222(a), states "The pressure retaining boundary includes only those portions of the system required to operate or support the safety function up to and including the first normally closed valve (including a safety or relief valve) or valve capable of automatic closure when the safety function is required."

4.0 Reason for Request Nuclear Regulatory Commission (NRC) Information Notice 2014-02, Failure to Properly Pressure Test Reactor Vessel Flange Leak-Off Lines (Reference 2), was issued on February 25, 2014. It provided examples of instances where licensees did not perform, or inadequately performed, system pressure tests of reactor vessel flange leak-off lines.

During the Third Ten-Year Interval of Wolf Creek Nuclear Operating Corporation's (WCNOC) lnservice Inspection (ISi) Program, a similar 10 CFR50.55a Request, 13R-11,

Attachment to ET 17-0007 Page 3 of 7 was submitted to the NRC as a proposed alternative to the ISi requirements of the ASME Code Section XI, ISi Program for WCGS. This was approved by the NRC by letter transmitted on January 28, 2015 (Refer to Section 7.0, Precedents -TAC No. MF4304).

The following discussion provides the reasons for the requested relief and the need for NRC approval of the proposed alternative testing in accordance with 10 CFR 50.55a(z)(2) because complying with ASME Code Section XI, IWC-5221 requirements would result in a hardship to the station without a compensating increase in quality and safety.

The ASME Code Section XI, 2007 Edition through 2008 Addenda requires that Class 2 pressure boundary piping shall be pressure tested once each inspection period. The reactor vessel flange seal leak detection piping is separated from the reactor* coolant pressure boundary by a metallic 0-ring seal. The pressure openings for the leak detection piping are located on the reactor vessel flange mating surface. The inboard pressure opening is located between the inner and outer 0-rings. The outboard pressure opening is located on the outboard side of the outer 0-ring in the reactor vessel flange. Failure of the inner 0-ring is the only condition under which the reactor vessel flange leak-off line could be pressurized to any significant pressure. Therefore, the reactor vessel flange leak-off line is not expected to be pressurized during the system pressure test following a refueling outage or during normal operation. The leak-off line located outboard of the outer 0-ring is outside of both 0-ring sealing boundaries. The isolation valve for the outboard leak-off line, BBV0079, is normally closed. The isolation valve for the inboard leak-off line between the two 0-rings, BBV0080, is normally open during plant operations.

The configuration of this piping poses personnel and equipment safety concerns if pressure testing is performed at Reactor Coolant System (RCS) operating pressure:

• Plugs would need to be installed _in the reactor vessel flange face to act as a pressure boundary for each test, and then removed after the test.

• The installation of the plugs and subsequent use would incur additional radiological dose due to additional time for personnel at the reactor vessel flange.

• The plugs would also present a foreign material exclusion issue for the handling of a very small diameter plug over the reactor vessel that would be required to be installed to complete a leakage test at pressure.

• The use of an alternative test rig to test those isolated portions of piping to full RCS operating pressure would have to include application of a compatible pressurized medium. This would result in personnel stationed near pressurized vent or drain valves, exposing them to unnecessary personal safety hazards in the event of a leak from the non-class test pressure rig connections. A break at any connection of the rig under such conditions (temporary non-code connections under RCS test pressure) would pose a substantial personnel safety hazard.

The configuration also precludes pressurizing, the line externally with the reactor vessel closure head installed. The closure head contains two concentric grooves that hold the inner and outer 0-rings. The 0-rings are held in place by a series of retainer clips that are housed in recessed cavities in the flange face. If a pressure test were to be performed with the closure head installed, the 0-ring would be pressurized in a direction opposite to its design function. This test pressure would result in a net inward force on the inner 0-

Attachment to ET 17-0007 Page 4 of 7 ring that would tend to push it into the recessed cavity that houses the retainer clips. The thin 0-ring material could be damaged by the inward force.

• Purposely failing or not installing the inner 0-ring in order to perform a pressure test would require a new 0-ring set to be installed each time the test is conducted. This would result in additional time needed during the outage and additional radiation exposure to personnel involved with the removal and reinstallation of the reactor vessel closure head.

5.0 Proposed Alternative and Basis for Use WCNOC proposes to use reduced pressure testing as an alternative for the ASME Code required pressure. The pressure tests and VT-2 examinations performed at the lower pressures (as an alternative to the_pressure requirement in ASME Code Section XI, IWC-5221) provide an acceptable level of assurance of the leakage integrity and operational readiness of the tested piping.

During the operating cycle if the inner 0-ring should leak it will be identified by an increase in temperature of the leak-off line above ambient temperature. This leak detection piping has a temperature indication arid a high temperature alarm in the Control Room, which is monitored by the operator. This piping also acts as a leak-off line to collect leakage which is routed to the Reactor Coolant Drain Tank.

Additionally, the reactor vessel flange leak detection piping would only function as an ASME Code Class 2 pressure boundary if the inner 0-ring fails; thereby, pressurizing the line. If any significant pipe through-wall leakage were to occur in the leak detection piping during this time of pressurization, it would exhibit boric acid accumulation that would be identified by the boron trace residue during the VT-2 visual examination to be performed as proposed in this request.

The boundary of the proposed pressure test is from the reactor vessel flange to 88HV8032 [Refer to Figure 1]. The length of the inner 0-ring leak-off line, 88-076-8C8, is approximately 69 feet. The added length of the line 88-077-8C8 is approximately 6 feet. Line 88-077-8C8 is insulated and accessible. Line 88-076-8C8 is insulated and accessible from 88-077-8C8 to the secondary shield wall, approximately 8 feet. The portion of line 88-076-8C8 between the secondary shield wall and the primary shield wall is not insulated and is accessible, approximately 28 feet. The remaining portion of 88-076-8C8 runs through a primary shield wall reactor vessel main loop nozzle penetration and inside the reactor vessel main loop nozzle gallery between the inside of the primary shield wall and the reactor vessel. Portions of this piping run behind the primary shield wall mirror insulation on the inside of the primary shield wall. The entire piping in the nozzle gallery between the inside of the primary shield wall and the reactor vessel is inaccessible for direct VT-2 examination with the refueling cavity flooded. Line 88-075-8C8 runs nearly parallel to 88-076-8C8 with similar dimensions and configuration.

VT-2 Examination of Accessible Lines The proposed alternative to the System Leakage Test requirements of ASME Code Section XI, Table IWC-2500-1, Examination Category C-H, Item No. C7 .10, and ASME Code Section XI, IWC-5221, is to perform a VT-2 visual examination of the accessible areas of the leak detection system piping while the system is subjected to the static pressure from the head of water when the reactor cavity is filled to its normal refueling water level, 23 feet of water or greater, for at least four hours. The static head developed

Attachment to ET 17-0007 Page 5 of 7 with the leak-off lines filled with water will allow for the detection of pressure boundary leakage.

The examination of the insulated and non-insulated accessible piping will be in accordance with ASME Code Section XI, IWA-5241, as applicable.

VT-2 Examination of Lines Inaccessible with the Refueling Cavitv Flooded The piping located beneath the refueling cavity is inaccessible while the refueling cavity is flooded. It includes line 88-075-BCB and 88-076-BCB which run parallel to each other with similar dimensions and configuration. The inaccessible portion of 88-075-BCB and 88-076-BCB runs through a primary shield wall reactor vessel main loop nozzle penetration and inside the reactor vessel main loop nozzle gallery between the inside of the primary shield wall and the reactor vessel. Portions of this piping run behind the primary shield wall mirror insulation on the inside of the primary shield wall. The entire piping in the nozzle gallery between the inside of the primary shield wall and the reactor vessel is inaccessible for direct VT-2 examination while the refueling cavity is flooded.

The examination of the inaccessible piping will be in accordance with ASME Code Section XI, IWA-5241 (b) supplemented with subsequent VT-2 visual examination for boric acid residue indicative of leakage from the leak-off piping when the piping can be made accessible later in the refueling outage after drain down of the refueling cavity and access to the reactor vessel nozzle gallery is made available. This supplemental VT-2 visual examination will include opening of the mirror insulation that covers a portion of the inaccessible leak-off piping in the nozzle gallery to allow direct performance of the VT-2 visual examination.

WCNOC did not identify any degradation during the pressure test conducted in RF20 (Period 3 of Interval 3), which utilized the same alternative as requested in this 10 CFR 50.55a request. The pressure test in RF20 included removal of the insulation from the leak-off piping inside of the reactor vessel main loop nozzle gallery and examination for evidence of leakage after drain down of the refueling cavity and access was made available to the gallery.

The proposed alternative is based on ASME Code Case N-805, "Alternative to Class1 Extended Boundary End of Interval or Class 2 System Leakage Testing of Reactor Vessel Flange 0-ring Leak Detection System" (Reference 3). This code case was issued to the 2010 Edition of the ASME Code Section XI and is included in the 2010 Edition, Supplement 6, of the ASME BPV Nuclear Code Case book. Code Case N-805* has not been approved by the NRC and is not identified in the current issue of Regulatory Guide 1.147, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1".

6.0 Duration of Proposed Alternative Relief is requested for the Fourth Ten-Year ISi Interval, which ends on September 2, 2025.

Implementation of the alternative testing will be performed during Refueling Outage 22, which is the last refueling outage of the first period of the Fourth Ten-Year ISi Interval. Implementation of this alternative testing will also be performed during Periods 2 and 3 of the Fourth Ten-Year ISi Interval.

Attachment to ET 17-0007 Page 6of7 7.0 Precedents NRC letter from M.T. Markley, USNRC, to R.K. Edington, Arizona Public Service, "Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Request for Relief from the American Society of Mechanical Engineers (ASME) Code,Section XI Requirements Regarding the Reactor Vessel Head Flange Seal Leak Detection Piping (TAC Nos.

MF0447, MF0448, and MF0449)," April 4, 2013. ADAMS Accession No. ML13085A254.

NRC letter from M.T. Markley, USNRC, to Vice President Operations, Entergy Operations, Inc., "Arkansas Nuclear One, Unit 2 - Request for Relief AN02-ISl-015 from American Society of Mechanical Engineers (ASME) Code,Section XI, for Periodic Pressure Sealing Requirements on the Reactor Vessel Flange Leak Detection Piping (TAC No. MF0941),"

June 27, 2013. ADAMS Accession No. ML13161A241.

NRC letter from M.T. Markley, USNRC, to R. Flores, Luminant Generation Company LLC, "Comanche Peak Nuclear Power Plant, Unit 2 - Request for Relief from Pressure Test Requirements on Reactor Pressure Vessel Flange Leak-off Piping for the Second 10-year lnservice Inspection Interval (TAC No. MF2997)," April 4, 2014. ADAMS Accession No. ML14084A291.

NRC letter from E.R. Oesterle, USNRC to AC. Heflin, "Wolf Creek Generating Station -

. Request for Relief No. 13R-11 for the Third 10-Year In-service Inspection Program Interval (TAC No. MF4304)," January 28, 2015. ADAMS Accession No. ML15023A220.

8.0 References

1. ASME Boiler and Pressure Vessel Code,Section XI, 2007 Edition through 2008 Addenda, American Society of Mechanical Engineers, New York.

2. NRC Information Notice 2014-02: "Failure to Properly Pressure Test Reactor Vessel Flange Leak-Off Lines," February 25, 2014.

3. ASME Code Case N-805, "Alternative to Class1 Extended Boundary End of Interval or Class 2 System Leakage Testing of Reactor Vessel Flange 0-ring Leak Detection System," February 25, 2011.

9.0 Figures The boundaries of the portion of ASME Code Class 2 piping and components covered in this Request are indicated in Figure 1 on Page 7, which is a marked-up extraction from the applicable piping and instrumentation drawing.

Attachment to ET 17-0007 Page 7of7

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