ML14072A010
| ML14072A010 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/07/2014 |
| From: | Mark D. Sartain Virginia Electric & Power Co (VEPCO) |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| Download: ML14072A010 (6) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 March 7, 2014 United States Nuclear Regulatory Commission Serial No. 13-611A Attention: Document Control Desk SPS-LIC/CGL R1 Washington, D.C. 20555 Docket No. 50-280 License No. DPR-32 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
SURRY POWER STATION UNIT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELIEF REQUEST SI-15-SPT SYSTEM PRESSURE TEST OF BOTTOM OF REACTOR VESSEL By a letter dated November 26, 2013 (Serial No.13-611), Virginia Electric and Power Company (Dominion) submitted the Unit 1 Fifth 10-year Interval Inservice Inspection (ISI) Program and Alternative Requests, including Relief Request (RR) S1-15-SPT-01, "System Pressure Test of Bottom of Reactor Vessel."
On February 7, 2014, the NRC requested additional information regarding RR SI-15-SPT-01. The response to the NRC's request for additional information is provided in the attachment.
If you have further questions or require additional information, please contact Mr. Gary D. Miller at (804) 273-2771.
Sincerely, Mark D. Sartain Vice President - Nuclear Engineering
Attachment:
Response to Request for Additional Information, Relief Request S1 SPT System Pressure Test of Bottom of Reactor Vessel Commitments made by this letter: None
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Serial No. 13-611A Docket No. 50-280 Page 2 of 2 cc: U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue NE, Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector Surry Power Station Ms. M. C. Barillas, NRC Project Manager - Surry U. S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 G9A 11555 Rockville Pike Rockville, Maryland 20852 Dr. V. Sreenivas, NRC Project Manager - North Anna U. S. Nuclear Regulatory Commission One White Flint North, Mail Stop 08 G9A 11555 Rockville Pike Rockville, Maryland 20852 Mr. R. A. Smith Authorized Nuclear Inspector Surry Power Station
Serial No.13-611 A Docket No. 50-280 Attachment Response to Request for Additional Information Relief Request SI-15-SPT System Pressure Test of Bottom of Reactor Vessel Virginia Electric and Power Company (Dominion)
Surry Power Station Unit 1
Serial No. 13-611A Docket No. 50-280 Attachment Page 1 of 3 Response to Request for Additional Information Relief Request Sl-15-SPT System Pressure Test of Bottom of Reactor Vessel Surry Power Station Unit I By letter dated November 26, 2013 (Agencywide Documents Access and Management Systems (ADAMS) Accession No. ML13336A142), Dominion submitted a request for alternative (RFA) Sl-15-SPT-01 to a certain requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section Xl.
RFA Sl-15-SPT-01 relates to the inservice inspection (ISI) requirement of IWB-5220 during the system leakage test of the bottom of the reactor vessel every refueling outage.
On February 7, 2014, the NRC staff requested additional information to complete their review. The NRC questions and the Dominion responses are provided herein.
- 1. Discuss how the VT-2 visual examination is performed for the bottom of the reactor vessel? (e.g., Is this a bare metal visual of the reactor vessel bottom shell? If not, how is a potential leak from the insulated area on the bottom of the reactor vessel detected? Are there any vessel bottom head penetrations that are also subject to the VT-2 examinations under this request?).
Dominion Response: A visual examination (VT-2) is performed with the insulation in place for the bottom of the reactor vessel and is conducted in accordance with ASME Section Xl, IWA-5242(a), (b), and (c). This examination is conducted as soon as conditions allow entry into the in-core area during reactor shutdown. Adherence to the code requirements, which are also incorporated into site specific examination procedures, would identify leakage from the bottom of the reactor vessel. If leakage is detected, the appropriate actions in accordance with IWA-5250(a) and (b) would be taken. There are bottom mounted instrumentation penetrations that are included in the examination boundary.
- 2. Section 5 of the request stated, in part, that
"...Dominion will continue to monitor leakage on the bottom of the reactor vessel with other surveillance requirements and alarms..."
Describe what the "othersurveillance requirements and alarms" are. Discuss the leakage detection capabilities at the plant, or any measure(s) taken, to monitor and identify leakage in an unlikely event of a through wall leak on the bottom head of the reactorvessel during normal operation.
Dominion Response: Surry Technical Specification (TS) 3.1.C permits 1 gpm unidentified and 10 gpm identified Reactor Coolant System (RCS) operational leakage; the following discussion is provided in the TS 3.1.C Bases:
Serial No. 13-611A Docket No. 50-280 Attachment Page 2 of 3
" Unidentified LEAKAGE - One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO [limiting condition for operation] could result in continued degradation of the RCPB [reactor coolant pressure boundary], if the LEAKAGE is from the pressure boundary.
- Identified LEAKAGE - Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System.
Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Violation of this LCO could result in continued degradation of a component or system.
Leak rate calculations for unidentified and identified leakage are conducted daily.
Leakage originating from the bottom of the reactor vessel would be collected in the in-core sump room. The in-core sump room sump pump discharges to the containment sump. The in-core sump room sump and the containment sump level switches provide alarm indication and pump control signals. The high level alarm from the in-core sump and containment sump would alert control room operators of a high level condition. In addition, the daily leak rate calculations would identify an increase in unidentified leakage. The sensitivity to changes in leak rates is reflected in the procedure(s) that govern their performance. Leak rates are evaluated and trended on a daily basis.
Leakage exceeding action levels prompts entry into procedure(s) for identifying the source(s) of the increased leakage.
The containment atmosphere is monitored by gaseous and particulate radioactivity monitors with indication provided in the Main Control Room. Additionally, the containment atmosphere is sampled and analyzed for particulate, iodine, and noble gas radioactivity levels on a monthly basis with the results being trended for any adverse indications.
- 3. For justifications that the structural integrity and leak tightness of the bottom of reactorvessel be reasonably ensured without performing the requiredASME Code system leakage test, discuss whether there has been any plant-specific, fleet, and industry operating experience (OE) regarding potential degradation of the bottom head of the reactorvessel due to known degradationmechanisms that would lead to leakage.
Dominion Response: There has been no Surry plant-specific or Dominion fleet OE regarding potential degradation of the bottom head of the reactor vessel due to known
Serial No. 13-611A Docket No. 50-280 Attachment Page 3 of 3 degradation mechanisms that would lead to leakage. Surry conducted voluntary volumetric examinations of the bottom mounted nozzles (BMNs) of Unit 1 in November 2004 and of Unit 2 in May 2005 with no issues identified. The Code Case N-722-1 bare metal examinations that are conducted every other refueling outage have also identified no issues.
The following is a list of primary water stress corrosion cracking (PWSCC) experience in BMNs (from EPRI Materials Reliability Program Document MRP-372).
" Takahama Unit 1: A possible indication of PWSCC (< 1 mm) in the base metal of BMN No. 48 was detected through eddy current testing (ET) from the nozzle inner diameter (ID). The ET was performed prior to planned water jet peening of the nozzle ID. There have been no subsequent reports of detected indications in BMNs at Takahama Unit 1 or any other Japanese utility.
" South Texas Project Unit 1: Leakage from BMNs No. 1 and 46 was identified during walkdowns to support the utility's Boric Acid Corrosion Control Program.
Non-destructive examination (NDE) did not reveal any other indications of PWSCC beyond the cracking in BMNs No. 1 and 46.
" Gravelines Unit 1: Indications that are interpreted as PWSCC were detected via ultrasonic testing (UT) from the inside of the nozzles as part of the third 10-year inservice inspection (ISI). There was no evidence of boric acid deposits on the bottom head during the second and third ISIs. lectricit6 de France (EDF) is expanding the requirement for internal UT inspection of BMNs to all reactor vessels in the French fleet.
" Palo Verde Unit 3: Leakage was observed at BMN No. 3 during Code Case N-722-1 examinations. Phased Array UT revealed no indications of wastage. The remaining nozzles showed no unacceptable indications.
The above indications or leakage was discovered when the systems were not at normal operating temperature or pressure. The proposed alternative will continue to perform the ASME Code system leakage test; however, it will be performed in a more habitable environment. Code Case N-722-1, B15.80, Reactor Pressure Vessel (RPV) bottom-mounted instrument penetrations requires all penetrations to be examined every other refueling outage. At Surry, the required direct visual examination (VE) of the bare metal surface is performed with the insulation removed.
- 4. The licensee asked relief from the IWB-5220 "System Leakage Test" requirements.
The NRC staff notes that IWB-5220 contains two separate paragraphs,IWB-5221 and IWB-5222. Confirm that the request for altemative will be to both of the IWB-5221 and IWB-5222 requirements. If yes, provide alternative to each paragraphof IWB-5222 (i.e., IWB-5222(a) and IWB-5222(b)).
Dominion Response: The request for alternative is for IWB-5221. The boundaries are unchanged, thus an alternative to IWB-5222 is not needed.