ML14055A271
| ML14055A271 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 02/21/2014 |
| From: | Church C Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML14055A271 (7) | |
Text
Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 February 21 ,2014 10 CFR 50.55a ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 1 Facility Operating License No. NPF-90 NRC Docket No. 50-390 Subject. Response to Request for Additional lnformation Regarding American Society of Mechanica! Engineers (ASME) Boiler and Pressure Vessel (B&PV) Section Xl Code of Record for Second 10-Year lnservice Interval 2001 Edition
References:
Boiler and Pressure Vessel (B&PV) Section Xl Code of Record for Second 10 Year Inservice lnterval 2001 Edition," dated January 10, 2014
- 2. Electronic Mailfrom NRC to TVA, "Watts Bar Nuclear Station, Unit 1 -
Request for Additional lnformation Related Request for Alternative Number ISPT-O2, Regarding System Leakage Test of Reactor Pressure Vessel Head Flange Seal Leak-Off Piping," dated February 11,2014 By letter dated January 10,2014 (Reference 1), the Tennessee Valley Authority (TVA) submitted a request for alternative regarding system leakage test of reactor pressure vessel head flange seal leak-off piping. By Electronic Mail from NRC to TVA dated February 11,2014 (Reference 2), TVA received a Request for Additional lnformation (RAl) regarding the proposed request for alternative. The NRC requested the response by February 19,2014. On February 19,2014, Andrew Hon, WBN NRC Project Manager, approved a revised due date of February 21,2014.
The enclosure to this letter contains TVA's response to the Reference 2 RAl.
U.S. Nuclear Regulatory Commission Page2 February 21,2014 There are no regulatory commitments in this letter. Please address any questions regarding this request to the Licensing Director, Gordon Arent al (423) 365-2004.
k(k Resoectfullv.
Site Vice President Watts Bar Nuclear Plant
Enclosure:
Response to NRC Request forAdditional lnformation cc (Enclosure):
NRC RegionalAdministrator - Region ll NRC Resident lnspector - Watts Bar Nuclear Plant Director, Division of Radiological Health - Tennessee State Department of Environment and Conservation
ENCLOSURE TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT UNIT 1 RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION NRG Question 1:
The NRC requires deviations from the ASME Code, Secfrbn Xl, in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(ii) be pre-approved. RFA ISPT-l2 was submitted forthe second 1O-year lSl interualwhich commenced on May 27, 2007, and will end on May 26, 2016, for Watts Bar, Unit 1. The RPV flange leak-off piping is subject to system leakage test every inspection period in accordance with the ASME Code, Secfircn Xl, IWC-2500, Table IWC-2500-1, Examination Category B-P, ltem No. C7.10. Discusswhetherthe system leakage fesf was performed in the previous inspection periods (i.e., the first two periods) of the second 10-year lSl interual in accordance with IWC-5221 as specified in Table IWC-2500-1. lt the answer is No, drscuss the reason(s) for not obtaining the NRC prior authorization for deviation from the ASME Code requirements for the sysfem leakage testing in the previous inspection period(s). lf the answerr.s Yes, dr.scuss why the required sysfem leakage test in accordance with the ASME Code cannot be performed in the upcoming inspection period (the third period) of the second 1O-year lSl interual.
TVA Response:
Watts Bar Nuclear Plant, Unit 1 WBN1) performed a)fT-2 examination for Code credit of the subject piping during the Mode 3 walkdown at Reactor Coolant System (RCS) Normal Operating Pressu relNormal Operati ng Tem perature (NOP/NOT) d uri ng the WBN1R9 (Fall 2009) and WBN1R1 1 (Fall 2012) refueling outages. Additionally, the same areas were examined for credit of the Class 1 piping during the WBN1R8 (Spring 2008) and WBN1R10 (Spring 2010) refueling outages.
Historically, TVA believed the following to be true:
- 1. The requirements of IWC-5221 ("...test conducted at the system pressure obtained while the system, or portion of the system, is in service performing its normal operating function...") were met.
- 2. The normal operating function consisted of a standby function with only the static head of any residual water from the refueling outage contained within this piping segment.
- 3. Full RCS pressure within this piping segment is an abnormal operating condition that would occur only in the event of Vessel O-ring failure.
This interpretation of IWC-5221 would result in not performing the leakage test at full RCS pressure in this piping.
IWA-5243 (200112003; WBN1 Code of Record) addresses leakage of vessel flange gaskets and specifies that the VT-2 examination shal! be conducted by verifying that the leakage collection system is operative.
Upon review of the non-cited violation (NCV) issued at Palo Verde where the interpretation of IWC-5221 resulted in the determination that the applicable pressure for this piping while performing its normal operating function was full RCS pressure, TVA initiated an investigation
into its applicability to WBN1. The Palo Verde NCV addressed the use of IWA-5243 and a VT-2 examination while in Mode 3. The previous WBN1 examination methodology, while not the same, was of a similar nature. TVA identified the Palo Verde issue approximately 30 days following the conclusion of the WBN1R11 refueling outage and return of the Unit to normal power operation.
TVA placed this issue in the Conective Action Program, performed benchmarking, and maintained contact with other utilities with regard to determining the optimal course of action for resolution. As a proactive resolution for this issue, TVA submitted a Request forAlternative with a requested approval date to support examination at the next available opportunity, i.e., the upcoming WBNl R1 2 refueling outage.
NRC Question 2:
Dr.scuss whetherthe subject piping can be pressure tested in accordance with the ASME Code in the beginning of a refueling outage before removing the RPV head. lf yes, the proposed relief request would not be needed. lf no, discuss fhe hardship and potential personnel radiation exposure including an estimate for person-roentgen equivalent man (rem) exposure with consideration of as low as reasonably achievable (ALARA) forthe sysfem leakage test pefiormed in accordance with IWC-5221 before removing the RPV head.
TVA Response:
It is not possible to pressure test the subject piping in accordance with the NRC's interpretation of the ASME Code with the Reactor Pressure Vessel (RPV) head installed. Attempting to perform this test would most likely result in failure of the O-rings which would prevent achieving the required test pressure. Additionally, no barrier (additional O-ring) exists to ensure the test pressure could be achieved for testing of the monitoring tube for leakage past the outer O-ring.
Therefore, a pressure test performed in accordance with IWC-5221would require the RPV head to be removed and isolation points (plugs) installed in the RPV flange.
Performance of this pressure test in accordance with the ASME Code would result in substantial hardship and personne! radiation exposure. These hardships include:
Potential foreign material exclusion (FME) issues with installation and removal of small diameter, threaded plugs in the reactor vessel flange to provide the test boundary as this activity would be located at the edge of the open reactor vessel.
Subjecting site personnelto significant radiological dose to install and remove the test plugs which would be contrary to ALARA considerations. TVA estimates that the dose rates for these tasks would be approximately 1.0 to 1.5 R/hr.
lnstallation and remova! of shielding for this task would most likely not be effective and would also result in the potentialfor FME introduction and increased dose.
NRC Question 3:
On page 2 of Enclosure 1 of RFA ISPT-12, the licensee stated, in paft, " ... Purposely failing the inner O-ring to peiorm the ASME Code required test would require purchasing a new sef of O-rings, additionaltime, and radiation exposure to de-tension the RPV head, installthe new O-rings, and then reset and re-tension the RPV head." Provide an estimate for radiation exposure (i.e.,
person-rem exposure if the test in accordance with lWG5221 is pertormed after the RPV head and new O-rings are installed).
TVA Response:
TVA estimates that the dose rates for these tasks would be approximately 1.0 to 1.5 R/hr with an estimated duration of approximately one hour each to install and remove the test plugs to allow for pressure testing of the monitoring tubes. This action is contrary to ALARA practices.
NRC Question 4:
Provide materials of construction for the RPV flange leak-off piprng. Dr.scuss operating expeience (e.9., plant-specific, fleet, and industry) regarding potential degradation of the subject leak-off piping and ifs assocrated welded connections due to any known degradation mechanisms fhaf would lead to leakage.
TVA Response:
Materials of Construction:
o Piping and nipples are schedule 160, 5A376, TP304 material.
Fittings are 6000#, SA1 82, F304.
Tubing is 0.065" wall, 5A269 or SA213, type 316.
o Valves are 1500#, SA182, F316 with manufacturer's hydro test at 5400 psig.
o Design conditions are 2485 psig at 650'F.
o The reactor operating pressure is 2235 psig.
Potential Deqradation:
All piping and components were hydro tested to 3290 psig during construction. A portion of the piping between the isolation valves was retested to 3150 psig. All components involved in the installation of the RPV flange leak-off piping were purchased and installed to ASME B&PV Section lll requirements. All components are constructed of stainless steel which is not subject to degradation from boric acid. All piping and components within this segment, with the exception of the outer O-ring monitoring tube from the RPV flange to the closed manual isolation valve, are aligned to the Reactor Coolant Drain Tank (RCDT). The RCDT is normally maintained between 0.5 and 2.0 psig. The design pressure for the tank is 25 psig. Nitrogen is provided to the tank as a cover gas during normal operation. Because this piping is designed for borated water, but is normally dry, exposed to nitrogen, and at essentially atmospheric pressure, there are no degradation mechanisms. Only a severe O-ring failure would be able to pressurize the leakoff line prior to isolation. The piping is routed along a wall approximately 20 feet above the floor which prevents inadvertent damage occurring from other work activities.
TVA is not aware of any operating experience/history of reactor pressure vessel flange O-ring leakage either within TVA or the industry.
NRC Question 5:
What is the proposed test pressure (i.e., the numerical value for the static head pressure developed from water level above fhe vesse/ flange when the reactor cavity is flooded) for the proposed sysfem leakage testing of the subject leak-off line piping?
TVA Resoonse:
The taps in the vessel flange are located at elevation 723'-O 5/8". The normal water level for refueling operations is between elevation 749'-1.5" to 749'-2" pet plant operating procedures GO-7 and GO-10. This provides a static head of 25.5 ft or 11 psig.
NRC Question 6:
Dr.scuss leakage detection capabilities at the plant, or any measure(s) taken, to monitor and identrfy leakage in an unlikely event of a through wall leak in the RPV flange seal leak-off line piping concunent wrth bak or failure of the RPV flange inner seal during normal operation.
TVA Response:
The detection of RCS leakage at WBN1 conforms to the requirements of Regulatory Guide 1.45, "Guidance on Monitoring and Responding to Reactor Coolant System Leakage." The instrumentation can detect a 1 gallon per minute (gpm) leak within one hour. lnstrumentation credited for RG 1.45 conformance is containment floor and equipment drain sump level monitors, containment air particulate monitors, and containment radioactive gas monitors. ln addition, there are tank level monitors, containment temperature and humidity monitors, and for large leaks containment pressure monitors. These instruments would detect through wall leakage in the leakoff line in conjunction with O-ring leakage.
WBN1 monitors for RPV Flange O-ring leakage by use of a temperature element (1-TE-68-22) located downstream of the Class 2 air-operated header isolation valve (1-FCV-68-22). Once temperature of the leakage monitoring header reaches 140'F increasing, annunciator window 88A (Reactor Vessel Flange Leakoff Temperature Hi) alarms in the Main Control Room. The basis for the 140"F setpoint was to establish a limit based on ambient temperature near the temperature element plus 20'F.
Plant procedure ARI-88-94 (Reactor Coolant System) directs the Operations staff to conduct monitoring activities to validate the alarm and to determine if leakage to the Reactor Coolant Drain Tank (RCDT) is indicated. When leakage to the RCDT is indicated, the required actions are to initiate performance of 1-A0!-6 (Small Reactor Coolant System Leak) and to close the leakage monitoring air-operated header isolation valve. When permitted by Radiological Protection, further actions are taken to determine whether one or both O-rings are leaking.
Plant procedure 1-A0I-6 directs the Operations staff to check indicated temperature on the leakage monitoring header temperature indicator. When the indicated temperature exceeds 120"F, the Main Control Room staff is directed to isolate the leakage monitoring header by use of the air-operated valve and to evaluate the need for a Unit shutdown. Plant Technical Specifications directs WBNl to take action to reduce the leakage and/or initiate shutdown when leakage exceeds limits.
Plant procedure 0-Tl-68.017 (Reactor Building Post Shutdown Walkdown) requires walkdowns of the Reactor Building following shutdown for refueling outages and forced outage of sufficient duration to practically allow for performance of the walkdown. These areas are walked down at the beginning of each refueling outage. The Keyway under the vessel and the Excess Letdown Heat Exchange Room are excluded from normalforced outage walkdowns. Performance of this procedure would identify any leakage from the Vessel Flange leak-off lines from the biological shield wallto the end of the Class 2 piping boundary. A keyway entry would be required to check for leakage that may have occurred in this piping from the Vessel flange to the biological shield wall. ln the event that vessel flange leakage was detected within the leak-off lines and the Unit was not shutdown, the only piping accessible for visual exam would be outside the Polar Crane Wall and located in the Raceway. The effort to locate leakage inside the Polar Crane Wallwould require aftempts to insert a camera through penetrations to detect through-wall leakage that may exist coincident with leakage past the vessel O-rings.