ML14239A495

Wolf Creek, Response to Request for Additional Information Regarding Request for Alternative 13R-1 1

ML14239A495
Person / Time
Site:	Wolf Creek
Issue date:	08/21/2014
From:	Reasoner C Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ET 14-0026
Download: ML14239A495 (7)

Text

W AF E CRE K'NUCLEAR OPERATING CORPORATION Cleveland ReasonerVice President Engineering August 21, 2014ET 14-0026U. S. Nuclear Regulatory Commission ATTN: Document Control DeskWashington, DC 20555

Reference:

1)Letter ET 14-0019, dated June 26, 2014, from C. 0. Reasoner, WCNOC, to USNRC2) Letter dated July 25, 2014, from C. F. Lyon, USNRC, to A.C. Heflin,WCNOC

Subject:

Docket No. 50-482: Response to Request for Additional Information Regarding Request for Alternative 13R-1 1Gentlemen:

Reference 1 provided Wolf Creek Nuclear Operating Corporation's (WCNOC) 10 CFR 50.55aRequest Number 13R-1 1 for the Third Ten-Year Interval of WCNOC's Inservice Inspection (ISI)Program.

Request 13R-1 1 requests relief from the pressure test requirements of the AmericanSociety of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, IWC-5220.Reference 2 provided a Nuclear Regulatory Commission (NRC) request for additional information (RAI) regarding 10 CFR 50.55a Request Number 13R-11. The attachment providesWCNOC's response to the questions in the RAI.This letter contains no commitments.

If you have any questions concerning this matter, pleasecontact me at (620) 364-4171, or Mr. Steven R. Koenig at (620) 364-4041.

Sincerely, Cleveland ReasonerCOR/rltAttachment cc: M. L. Dapas (NRC), w/aC. F. Lyon (NRC), w/aN. F. O'Keefe (NRC), w/aSenior Resident Inspector (NRC), w/aP.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831An Equal Opportunity Employer M/F/HC/VET Attachment to ET 14-0026Page 1 of 6Response to Request for Additional Information Regarding Request for Alternative 13R-11 to ASME Code Requirements Reference 1 provided Wolf Creek Nuclear Operating Corporation's (WCNOC) 10 CFR 50.55aRequest Number 13R-1 1 for the Third Ten-Year Interval of WCNOC's Inservice Inspection (ISI)Program.

Request 13R-1 1 requests relief from the pressure test requirements of the AmericanSociety of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code, Section X1,IWC-5220.

Reference 2 provided a Nuclear Regulatory Commission (NRC) request foradditional information (RAI) regarding 10 CFR 50.55a Request Number 13R-11. The specificNRC questions are provided below in italics followed by the WCNOC response.

1. Please discuss whether or not the subject piping can be pressure tested in accordance with the ASME Code, Section Xl, IWA-5000 and IWC-5000 before removing the RPVhead in the beginning of a refueling outage. If yes, the proposed relief request would notbe needed. If no, please discuss the hardship and potential personnel radiation

exposure, including an estimate for person-roentgen equivalent man (rem) exposurewith consideration of as low as reasonably achievable (ALARA) for the system leakagetest performed in accordance with IWC-5221 before removing the RPV head in thebeginning of a refueling outage.Response:

The following response will first describe alternative test options beyond thosealready discussed in the WCNOC Request 13R-1 1, which could possibly be used to pressuretest the leak-off piping to a higher pressure although still not at the reactor coolant system(RCS) pressure.

The response will describe the related hardships associated with thesealternative test options and finally describe how the Request 13R-1 1 test method would beeffective in detecting potential piping leakage.As outlined in Request 13R-1 1, with the reactor vessel closure head installed and the RCSdepressurized, any pressurization of the annulus between the inner and outer O-rings createspressure forces in a direction opposite to the design function of the inner O-ring, which couldresult in degradation or damage. Thus, that test option is not considered feasible.

Performing an alternative higher pressure test has two limitations.

First, performing analternative higher pressure test by pressurizing the leak-off piping at the beginning of therefueling outage while the reactor vessel closure head is still installed would require that RCSpressure always be higher than the O-ring annulus test pressure.

Second, an alternative higherpressure test would not test the open-ended portion of line 075-BCB-3/4, the portion of the leak-off piping that originates outside of the outer O-ring and extends to normally closed valveBBV0079.

The only test that does pressurize this portion of line 075-BCB-3/4 is the proposedtest in Request 13R-11.Alternative Higher Pressure Test Options:It could be possible to connect a portable hydro pump skid at the blind flange connection downstream of valve BBV0081 to pressurize leak-off lines BB-076-BCB and BB-077-BCB throughBBV0081 from the reactor vessel to the Code boundary isolation valve BBHV8032.

Thisalternative higher pressure test could be performed at either of two operating plateaus:

normaloperating pressure and temperature in Mode 3 or at approximately 350 psia in Mode 5.Immediately after reactor shutdown for a refueling outage, the plant is stabilized in Mode 3,followed closely by RCS cool down. To perform a leak-off line alternative test at or near fullRCS pressure would require holding the plant at normal operating pressure and temperature for Attachment to ET 14-0026Page 2 of 6a minimum of six extra hours beyond the normal refueling outage sequence, while a portablehydro pump skid is installed and operated to pressurize and hold lines BB-076-BCB and BB-077-BCB at just below RCS operating pressure for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to performance of the VT-2 examinations.

At this time, the entire containment would have limited access asradiation protection personnel would not have sufficient time to survey and de-post the variousareas required to setup and perform the test. Once a cool down starts, the plant enters Mode 5approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> later. The RCS is stabilized at approximately 350 psia to keep thereactor coolant pumps (RCPs) running for peroxide injection

/ crud removal (to minimizepersonnel dose during the refueling outage as a whole) and to cool down the RCS metal mass.This 350 psia plateau is maintained for about 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The portable hydro pump skid could beinstalled and operated at this lower pressure plateau.

However, during the peroxide injection Icrud removal process, which also occurs at this pressure level, the radiation levels increasesuch that radiation protection personnel prohibit entry into the areas required to setup andperform the test. As a result, performance of a leak-off line alternative test at this lower RCSpressure would require postponing peroxide injection

/ crud removal process for a minimum ofsix extra hours while a portable hydro pump skid is installed and operated to pressurize andhold lines BB-076-BCB and BB-077-BCB at just below current RCS system pressure for aminimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to performance of the VT-2 examinations.

Alternative Higher Pressure Test Method Hardships:

The performance of an alternative higher pressure test would impose the following hardships:

• The hydro pump test skid pressure and water quality must be adequately controlled, such that the reactor vessel O-rings are not over-pressurized or compromised.

There isadditional risk of O-ring damage and potential RCS leakage resulting from the alternative test method, as compared to the proposed Request 13R-1 1 test method." If an operational transient were to occur during testing such that the RCS pressure wereto decrease below the O-ring annulus test pressure, inner O-ring loading would reverseresulting in a condition that has not been fully analyzed.

• Control Room operators would have to hold the RCS at an elevated pressure andtemperature for a longer period (i.e., Mode 3 pressure test option) or postpone peroxideinjection

/ crud removal process (i.e., Mode 5 pressure test option),

either of which wouldimpact critical path time at the start of the refueling outage.* Additional personnel dose would be accrued to transport,

install, test and remove thehydro pump test skid at a point in the refueling outage that would require continuous radiation protection personnel coverage (i.e., Mode 3 pressure test option).* Since testing the reactor pressure vessel (RPV) leak-off piping prior to removing thereactor vessel closure head would involve the use of a hydro pump test skid, this couldexpose personnel stationed near pressurized vent or drain valves to unnecessary personal safety hazards in the event of a leak from or break of a non-class test pressureskid connection (i.e., Mode 3 pressure test option).10 CFR 50. 55a Request 13R- 11 Test Method:The above described alternative higher pressure test options are not consistent with the ASMESection Xl Code intent for conduct of pressure tests in accordance with IWA-5000 and IWC-5000. In fact, ASME Code Case N-805, "Alternative to Class 1 Extended Boundary End ofInterval or Class 2 System Leakage Testing of Reactor Vessel Head Flange O-ring LeakDetection System,"

was approved by the ASME Section Xl Standards Committee and isincluded in the 2010 Edition, Supplement 6, of the ASME BPV Nuclear Code Case book. WhileCode Case N-805 has not yet been endorsed by the NRC in Regulatory Guide 1.147, "Inservice Attachment to ET 14-0026Page 3 of 6Inspection Code Case Acceptability, ASME Section Xl, Division 1," Request 13R-1 1 is consistent with the Code Case methods and is, therefore, an appropriate test approach.

The testing methodology proposed would be performed when the refueling pool is filled. Duringa refueling outage, the refueling pool is normally kept filled for several days while fuelassemblies are off-loaded and other outage activities are performed.

The leak-off piping visualexamination, as proposed in Request 13R-1 1, would be performed at some time after therefueling pool has been flooded for an extended period. As the visual examination is looking forthrough-wall flaws, the proposed test is considered a reasonable and sufficient method, even forsmall leak rates that result from the proposed low test pressure, since the piping is exposed tothe pressure for a number of days. Additionally, as noted above, this test method is the onlyfeasible test that tests line 075-BCB-3/4, the portion of the leak-off piping that originates outsideof the outer O-ring (open ended) and extends to normally closed valve BBV0079.Summary:The test methodology of Request 13R-1 1 is appropriate for detecting through-wall leakage.Performance of either of the alternative higher pressure tests at the beginning of the refueling outage, prior to removing the reactor vessel closure head, is not justified based upon theadditional operational risk and hardships.

For the reason stated above, either of the alternative higher pressure tests would constitute a hardship without a compensating increase in the levelof quality and safety.2. In Section 4.0 of the Attachment to the letter dated June 26, 2014, the licensee stated, inpart,The installation of the plugs and subsequent use would incur additional radiological dose due to additional time for personnel at the reactor vesselflange.Please provide an estimate for amount of personnel radiation exposure (i.e., person-rem exposure) for the installation of the plugs and subsequent removal.Response:

Installation of pipe plugs has not been performed during the operational life of theplant. The condition of the threads in these two connections is unknown.

Therefore, anaccurate estimate of personnel radiation exposure for the installation of the plugs andsubsequent removal cannot be made. However, based on past refueling outage dose surveys,the dose rates at the reactor vessel flange are 3 to 4 REM/hour.

This information is provided toindicate the high radiation exposure for personnel attempting to install and remove these pipeplugs if such were attempted.

The concerns documented in Request 13R-11 along with thepotentially high dose received while attempting to install plugs, makes this option a significant hardship and an option that has no assurance of successfully testing the leak-off piping.3. Please explain if any segment(s) of the RPV flange leak-off piping is insulated andinaccessible for the ASME Code-required VT-2 visual examination.

If yes, pleasediscuss how the licensee will perform the VT -2 visual examination of the insulated andinaccessible segment(s) of the subject piping.Response:

The length of the inner O-ring leak-off line BB-076-BCB is approximately 69 feet.The added length of line BB-077-BCB is approximately 6 feet. Line BB-077-BCB is insulated and accessible.

Line BB-076-BCB is insulated and accessible from BB-077-BCB to thesecondary shield wall, approximately 8 feet. The portion of line BB-076-BCB between the Attachment to ET 14-0026Page 4 of 6secondary shield wall and the primary shield wall is not insulated and is accessible, approximately 28 feet. The remaining portion of BB-076-BCB runs through a primary shield wallreactor vessel main loop nozzle penetration and inside the reactor vessel main loop nozzlegallery between the inside of the primary shield wall and the reactor vessel. Portions of thispiping run behind the primary shield wall mirror insulation on the inside of the primary shieldwall. The entire piping in the nozzle gallery between the inside of the primary shield wall andthe reactor vessel is inaccessible for direct VT-2 examination with the refueling cavity flooded.Line BB-075-BCB runs nearly parallel to BB-076-BCB with similar dimensions and configuration.

The examination of the insulated accessible piping will be in accordance with Section Xlparagraph IWA-5242.

The examination of the non-insulated accessible piping will be inaccordance with Section Xl paragraph IWA-5241 (a). The examination of the inaccessible pipingwill be in accordance with Section Xl paragraph IWA-5241(b) supplemented with subsequent VT-2 visual examination for boric acid residue indicative of leakage from the leak-off pipingwhen the piping can be made accessible later in the refueling outage after drain down of therefueling cavity and access to the reactor vessel nozzle gallery is made available.

Thissupplemental VT-2 visual examination will include opening of the mirror insulation that covers aportion of the inaccessible leak-off piping in the nozzle gallery to allow direct performance of theVT-2 visual examination.

4. Please discuss the construction materials for the subject RPV flange leak-off piping.Discuss operating experience (e.g., plant-specific, fleet, and industry) regarding potential degradation of the subject piping due to any known degradation mechanisms that wouldlead to pipe through-wall leakage.Response:

The RPV flange leak-off piping material is Type 304 Stainless Steel, Piping Class2500#, ASME Section III Class 2.In January 1988, during plant heat up following a refueling outage, WCNOC, identified leakagefrom the inner RPV flange O-ring seal. The plant was subsequently cooled down and the 0-rings were replaced.

No degradation in the RPV flange leak-off piping was identified.

This wasthe only time the RPV leak-off piping at Wolf Creek Generating Station (WCGS) experienced elevated pressure since plant construction.

At Calvert Cliffs in January 1994, cracking was found in the RPV head O-ring leakagemonitoring lines in both Units 1 and 2. Analysis of the failed material revealed that the linescracked from transgranular stress corrosion cracking caused by elevated chloride concentration.

It was concluded that the chlorides were present since construction.

WCNOC has not identified any degradation of the RPV flange leak-off piping during requiredVT-2 examinations at the end of each refueling outage, which includes accessible portions ofthe leak-off piping after pressurization from the static head of the flooded refueling cavity. Inaddition, in Refueling Outage 18 (Spring 2011), the mirror insulation was removed from theleak-off piping inside the reactor vessel main loop nozzle gallery (during walk down and workplanning for reactor vessel nozzle mitigation) with no evidence of leak-off piping leakageidentified.

Attachment to ET 14-0026Page 5 of 65. Please discuss the reactor coolant system leakage detection capabilities at the plant, orany measure(s) taken, to monitor and identify leakage in an unlikely event of a throughwall leak in the RPV flange seal leak-off line piping concurrent with leak or failure of theRPV flange inner seal during normal operation.

Response:

During the operating cycle if the inner O-ring should leak it will be identified by anincrease in temperature of the leak-off line above ambient temperature.

This leak detection piping has temperature indication and a high temperature alarm in the Control Room, which ismonitored by the Control Room Operators.

This piping also acts as a leak-off line to collectleakage which is routed to the Reactor Coolant Drain Tank, whose level is indicated andalarmed in the Control Room. If inner O-ring leakage is identified, an operator is directed toenter containment and place the outer O-ring leakage detection in service by closing valveBBV0080 and opening valve BBV0079.

The Control Room Operators then monitor outer O-ringseal for leakage.

If outer O-ring seal leakage is identified, the Control Room Operators willinitiate an orderly plant shutdown.

An unlikely event of a through wall leak in the RPV flange seal leak-off line piping concurrent with leak or failure of the RPV flange inner seal during normal operation would result inunidentified RCS leakage controlled by Technical Specification 3.4.13. Leakage detection systems have been designed to aid Control Room Operators in differentiating between possiblesources of detected leakage within the containment and identifying the physical location of theleak. The RCS leakage detection system consists of the sump level and flow monitoring system, the containment air particulate monitoring system, the containment cooler condensate measuring system, and the containment humidity monitoring system. The sump level and flowmonitoring system indicates leakage by monitoring increases in sump level. The containment cooler condensate measuring system and the containment humidity measuring system detectleakage from the release of steam or water to the containment atmosphere.

The air particulate gas monitoring system detects leakage from the release of radioactive materials to thecontainment atmosphere.

The containment atmosphere particulate monitoring system provides the primary means ofremotely determining the presence of reactor coolant leakage within the containment.

Leakageto the containment atmosphere from the RCS would cause a change in the containment airborne radioactivity which would be detected by the air particulate monitors.

Increases incontainment airborne activity levels detected by either of the monitors indicate the reactorcoolant pressure boundary as the source of leakage.Since a leak in the flange leak-off piping concurrent with the inner O-ring leakage is a leak in theprimary system, it would result in reactor coolant flowing into the containment normal orinstrument tunnel sumps, which would be indicated by a level increase in the sumps. Indication of increasing sump level is transmitted from the sump to the control room level indicator bymeans of a sump level transmitter.

The system provides measurements of low leakages bymonitoring level increase versus time. Increases in the frequency of a particular containment sump pump operation or increases in the level in a particular sump facilitate localization of thesource to components whose leakage would drain to that sump. If operator actions to place theouter O-ring seal leak-off line in service did not isolate the leak, operators would perform anRCS leak rate determination to identify the type of leak and quantify the leakage and takeactions as required by Technical Specifications 3.4.13.

Attachment to ET 14-0026Page 6 of 6

References:

1. WCNOC Letter ET 14-0019, "10 CFR 50.55a Request 13R-11 for the Third Inservice Inspection Program Interval,"

June 26, 2014. ADAMS Accession No. ML14182A091

2. Letter from C. F. Lyon, USNRC, to A. C. Heflin, WCNOC, "Wolf Creek Generating Station-Request for Additional Information Re: Request for Alternative 13R-1 1 (TAC No.MF4304),"

July 25, 2014. ADAMS Accession No. ML14197A336