ML14072A009
| ML14072A009 | |
| 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 | |
| 13-639A | |
| Download: ML14072A009 (9) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 March 7, 2014 United States Nuclear Regulatory Commission Serial No.
13-639A Attention: Document Control Desk SPS-LIC/CGL R1 Washington, D.C. 20555 Docket Nos.
50-280/281 License No.
DPR-32/37 VIRGINIA ELECTRIC AND POWER COMPANY (DOMINION)
SURRY POWER STATION UNITS I AND 2 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING REVISED RELIEF REQUESTS SPT-003 AND SPT-002 ALTERNATIVE TESTING REQUIREMENTS FOR SMALL DIAMETER REACTOR COOLANT SYSTEM PRESSURE BOUNDARY CONNECTIONS By a letter dated December 9, 2013 (Serial No.13-639), Virginia Electric and Power Company (Dominion) submitted revised Relief Requests (RRs) SPT-003 and SPT-002, "Alternative Requirements to ASME Code Requirements for Small Diameter Reactor Coolant System Pressure Boundary Connections," for Surry Units 1 and 2, respectively.
The revisions to RRs SPT-003 and SPT-002 specifically added the Units 1 and 2 Reactor Vessel Head Vent lines, respectively.
On February 7,
- 2014, the NRC requested additional information regarding revised RRs SPT-003 and SPT-002. 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 Regarding Revised Relief Requests SPT-003 and SPT-002 - Alternative Testing Requirements for Small Diameter Reactor Coolant System Pressure Boundary Connections Commitments made by this letter: None
-. 9-7
Serial No. 13-639A Docket Nos. 50-280/281 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. 13-639A Docket Nos. 50-280/281 Attachment Response to Request for Additional Information Regarding Revised Relief Requests SPT-003 and SPT-002 - Alternative Testing Requirements for Small Diameter Reactor Coolant System Pressure Boundary Connections Virginia Electric and Power Company (Dominion)
Surry Power Station Units I and 2
Serial No. 13-639A Docket Nos. 50-280/281 Attachment Page 1 of 6 Response to Request for Additional Information Regarding Revised Relief Requests SPT-003 and SPT-002 - Alternative Testing Requirements for Small Diameter Reactor Coolant System Pressure Boundary Connections Surry Power Station Units 1 and 2 By letter dated December 9, 2013 (Agencywide Documents Access and Management Systems (ADAMS) Accession No. ML13350A109),
Dominion submitted revised requests for alternative (RFAs) SPT-003 and SPT-002 to include the reactor vessel head vent (RVHV) lines. The alternative requests are associated with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Section Xl inservice inspection (ISI) requirement specified in IWB-5222(b) for system leakage testing conducted at or near the end of the inspection interval.
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. The licensee submitted RFAs SPT-003 and SPT-002 pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(a)(3)(i). In accordance with 10 CFR 50.55a(a)(3)(i), the NRC staff authorizes a proposed alternative to the ASME Code requirement only if the alternative is equivalent in quality to the ASME Code requirement. The staff finds that the proposed alternative boundary for the system leakage test would not achieve the equivalent of quality of the ASME Code requirement specified in IWB-5222(b). The NRC staff finds that submitting the proposed alternative with a hardship justification pursuant to 10 CFR 50.55a(a)(3)(ii) would be appropriate and acceptable. Please revise RFAs SPT-003 and SPT-002 accordingly or justify why the proposed alternative satisfies 10 CFR 50.55a(a)(3)(i).
Dominion Response: Dominion letter Serial No. (SN)13-639, dated December 9, 2013, provided a revision to Relief Requests (RRs) SPT-003 and SPT-002 to add the Units 1 and 2 RVHV lines, respectively. Revision of the original RRs SPT-003 and SPT-002, including the basis of request for relief and alternate provisions, are not being modified by the proposed addition. As noted in letter SN 13-639, the original versions of RRs SPT-003 and SPT-002 were approved by the NRC for Units 1 and 2 Fourth ISI Intervals by letters dated August 11, 2004 (TAC Nos. MB7762, MB7763, and MB9528) and September 9, 2004 (TAC Nos. MB7762, MB7763, and MB9528), respectively.
However, in response to this question, and as discussed in the Response to Question 3 below, a hardship justification pursuant to 10 CFR 50.55a(a)(3)(ii) is proposed for the RVHV lines.
Serial No. 13-639A Docket Nos. 50-280/281 Attachment Page 2 of 6
- 2.Section III of Attachments I and 2 of the requests state, in part, that
"...This code case [N-7981 allows the use of IWB-5222(a) for defining the pressure retaining boundary as an acceptable alternative to the extended pressure test boundaries of IWB-5222(b)..."
Item (1),Section IV, Attachments I and 2 of the requests state, in part, that
"... The RCS vent, drain, instrumentation, and sample connections will be visually examined for leakage and any evidence of past leakage, with the isolation valves in the normally closed position each refueling outage during the ASME Section Xl Class 1 System Leakage Test (IWB-5220)..."
The NRC staff notes that IWB-5220 contains two subparagraphs, IWB-5221 and IWB-5222. It is not clear to the NRC staff which subparagraph(s) the proposed alternative will be following. Please clarify whether the proposed alternative is to use the boundary specified in ASME Code Case N-798, or IWB-5222(a), in lieu of the requirement of IWB-5222(b).
Dominion Response: In letter SN 13-639, ASME Code Case N-798 was introduced with respect to the RVHV lines. Code Case N-798 was approved by the ASME Committee on December 20, 2010, subsequent to NRC approval of the original versions of RRs SPT-003 and SPT-002 on August 11, 2004 and September 9, 2004, respectively. For the RVHV lines, the boundaries of IWB-5222(a) shall apply as permitted by the Code Case.
- 3. Discuss in detail the hardships or unusual difficulties associated with compliance to IWB-5222(b) requirement when performing pressure tests. (Examples of hardship or unusual difficulty may include having to enter multiple technical specification (TS) limiting conditions for operations (LCO), radiation dose and as low as is reasonably achievable (ALARA) concerns, creating excessive plant personnel hazards, causing reactor shut down, and defeat of the double isolation barrier.) The Sequoyah request for alternative 11-SPT-1 (ADAMS Accession No. ML13178A280) may provide additional examples.
Dominion Response:
Surry Technical Specification (TS) 3.1.7, Reactor Vessel Head Vents, requires that "At least two Reactor Vessel Head Vent paths consisting of two isolation valves in series powered from emergency buses shall be OPERABLE and closed whenever RCS temperature and pressure are > 350'F and 450 psig." Thus, TS 3.1.7 does not permit leaving open or opening of the normally closed valve to accomplish the ASME Code IWB-5220 required System Leakage Test after reaching system test temperature and pressure.
Extending the pressure retaining boundary during the system pressure test to the reactor vessel head vent pressure retaining components beyond the first normally closed valve would require a number of off-
Serial No. 13-639A Docket Nos. 50-280/281 Attachment Page 3 of 6 normal temporary system configurations, such as the temporary installation of testing equipment to achieve test pressures at system segments beyond the first isolation valve.
While the RCS is being brought to normal operating temperature and pressure (approximately 2235 psig and 547 'F) in accordance with the plant TSs, the reactor vessel head vent piping segments are isolated from the RCS. Consequently, the head vent piping segments would be at pressures and temperatures that are less than full RCS pressure and temperature. The plant design configuration complies with TS 3.1.7 and the RCS boundary requirements for double isolation but cannot satisfy the code test requirement for nominal operating pressure associated with 100 percent rated reactor power (i.e., full RCS pressure and temperature between the piping segments).
Furthermore, the use of temporary hoses/test rig/hydrostatic pump to raise the piping segment volumes to full test pressure and temperature constitutes a personnel safety hazard and could adversely affect plant safety. Temporary hoses are not qualified to meet all aspects of the plant design (e.g., pressure, temperature, ASME Code, seismic, dead load). The failure of unqualified temporary hoses during full pressure testing could result in personnel injury, as well as the loss of the reactor coolant pressure boundary and reactor coolant inventory. In addition, establishment of and restoration from such temporary configurations could take a considerable amount of time to complete, result in an unwarranted increase in worker radiation exposure, contaminate test equipment, and potentially delay normal plant startup following the refueling outage. Additionally, this temporary configuration would not test the end cap or end flange connection that would be reinstalled following removal of the test connection. Thus, pressure testing of the RVHV piping segments would present a hardship or unusual difficulty in establishing a system configuration that would meet ASME Code IWB-5222(b) boundary requirements without a compensating increase in the level of quality and safety.
In addition, as stated in the bases of the original relief request, the revised RRs SPT-003 and SPT-002 provide an acceptable level of safety and quality with respect to the RVHV lines based on the following:
- 1) ASME Section Xl Code, 1998 Edition with addenda up to and including the 2000 Addenda, paragraph IWA-4540, provides the requirements for hydrostatic pressure testing of piping and components after repairs by welding to the pressure boundary.
IWA-4540(b)(6) excludes component connections, piping, and associated valves that are 1 inch nominal pipe size and smaller from the hydrostatic test. Visual examination of these < 1 inch diameter RCS vent/drain/sampling connections once each 10-year interval is unwarranted considering that a repair weld on the same connections is exempted by the ASME Xl Code.
- 2) The non-isolable portion of the RCS vent and drain connections will be pressurized and visually examinedas required. Only the isolable portion of these small diameter vent and drain connections will not be pressurized.
Serial No. 13-639A Docket Nos. 50-280/281 Attachment Page 4 of 6
- 3) These piping connections are typically socket welds that received a surface examination after installation.
- 4) The piping and valves are nominally heavy wall. These piping components and associated piping are towards the free end of a cantilever configuration (stub end isolated by either a valve or a flange). There is no brace or support for this portion of the pipe. Consequently, this portion does not experience any thermal loading.
- 5) This portion of the line is isolated during normal operation and does not experience pressure loading unless there is a leak at the first isolation valve.
- 6) The valves do not have an extension operator, so the rotational accelerations at the valve do not produce significant stress.
- 7) The stresses towards the free end of the cantilever due to other types of loading are only a small fraction of the applicable Code allowable.
- 4. Clarify whether the proposed alternative is for the piping segments identified in both tables on page 3 of Attachments I and 2 of the requests.
Dominion Response: The revision to RRs SPT-003 and SPT-002 was proposed only to include the Units 1 and 2 RVHV lines, respectively. These lines were added in the second table on page 3 of Attachments 1 and 2 of the revised RRs SPT-003 and SPT-002, respectively.
- 5. The NRC staff notes that the combined piping segments identified in both tables on page 3 of Attachments I and 2 of the requests contain exactly 19 and 15 segments, respectively. Clarify why Section I of Attachments I and 2 of the requests state, approximately 20 connections.
Dominion Response:
The statement "Approximately 20, small diameter (< inch),
Class 1, reactor coolant system (RCS) pressure boundary vent, drain, sample, and instrumentation connections." is wording contained in the original RRs SPT-003 and SPT-002. The original RRs were transmitted to the NRC by Unit 1 letter SN 02-642, dated December 12, 2002, and by Unit 2 letter SN 03-428, dated August 25, 2003. The original RRs did not include the listing of lines for which the relief was being requested.
Subsequently in responses to NRC requests for additional information (RAI), the listings of lines for which the relief was being requested were provided for 17 lines in the Unit 1 letter SN 02-642A, dated December 5, 2003, and for 13 lines by the Unit 2 letter SN 03-428A, dated May 5, 2004. The wording in the original RRs was not revised to reflect the specific number of lines identified in the RAI responses. Likewise, when RRs SPT-003 and SPT-002 were revised to include the RVHV lines, this original wording was not revised.
Serial No. 13-639A Docket Nos. 50-280/281 Attachment Page 5 of 6
- 6. Provide the name of system(s) for each piping segment listed in both tables on page 3 of Attachments 1 and 2 of the requests.
Dominion Response: In the tables on page 3 of Attachments I and 2 of RRs SPT-003 and SPT-002, respectively, the piping segments with RC in the line numbers are part of the Reactor Coolant System, the piping segments with CH in the line numbers are part of the Chemical and Volume Control System (CVCS), and the piping segments with SI in the line numbers are part of the Safety Injection System.
- 7. Provide materials of construction for each piping segment listed in both tables on page 3 of Attachments I and 2 of the requests.
Dominion Response: For the lines listed in the tables on page 3 of Attachments 1 and 2 of RRs SPT-003 and SPT-002, respectively, the material of construction is stainless steel. Specifically, the piping segments listed in Attachments 1 and 2 associated with SI piping segments are ASTM A269 austenitic stainless steel; all other piping segments are ASME SA376TP316 / ASTM A-376-TP-316 austenitic stainless steel, except for the SPT-002 line between 2-RC-36, 2-RC-186, and 2-RC-FNG-545A (Segment 6) which is a combination of both via a reducer.
- 8. Confirm that the code of record for the fourth 10-year ISI interval at Surry, Units 1 and 2, is the 1998 Edition through 2000 Addenda of the ASME Code.
Dominion Response: As reflected in Section II of Attachments 1 and 2 of RRs SPT-003 and SPT-002, respectively, the Code of Record for the fourth 10-year ISI interval at Surry Units 1 and 2 is the 1998 Edition through 2000 Addenda of the ASME Code.
- 9. For justifications that the structural integrity and leak tightness of the piping and associated components (listed in both tables on page 3 of Attachments 1 and 2 of the requests) will be reasonably ensured without required extension in pressure retaining boundary during system leakage test, discuss operating experience (i.e.,
plant specific, fleet, and industry) regarding potential degradation of the subject piping and components due to known degradation mechanisms (e.g., fatigue and stress corrosion cracking) that would lead to leakage.
Dominion Response: Operating experience reviews regarding potential degradation of the piping and associated components that are listed in the tables on page 3 of Attachments 1 and 2 of SPT-003 and SPT-002, respectively, were performed. No Surry plant specific occurrences and no Dominion fleet occurrences were identified involving stress corrosion cracking or fatigue in socket or butt welds in the subject piping and components.
Our review of the Sequoyah RAI response (ML13273A26, dated September 25, 2013) and St. Lucie RAI response (ML13192A326, dated July 3, 2013)
Serial No. 13-639A Docket Nos. 50-280/281 Attachment Page 6 of 6 identifies that their survey or queries in the industry identified no occurrences involving stress corrosion cracking or fatigue in socket or butt welds in the subject piping and components; the plants included in these reviews were Sequoyah, Watts Bar, St. Lucie, Turkey Point, Point Beach, and D. C. Cook.
In addition, the March 26, 2013 Quad Cities Unit 1 reactor pressure vessel head vent leakage event was reviewed. The apparent cause of the Quad Cities leak was poor weld quality as indicated by localized porosity and /or slag inclusion in the small bore carbon steel piping socket weld with ongoing possible corrosion and/or fatigue over time. The Quad Cities Unit 1 operating experience was determined to not be applicable to Surry based on design differences (i.e., Quad Cities is a boiling water reactor, and the leaking line was a Class 2 carbon steel pipe and fitting).