NL-16-1056, Proposed Alternative HNP-ISI-ALT-05-03 Version 1.0 Regarding Reactor Pressure Vessel Flange Seal Leak-off Piping Testing
| ML16188A313 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 07/06/2016 |
| From: | Pierce C Southern Co, Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| HNP-ISI-ALT-05-03, NL-16-1056 | |
| Download: ML16188A313 (12) | |
Text
Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, AL 35242 Tel 205.992.7872 SOUTHERN << \
Fax 205.992.7601 NUCLEAR A SOUTHERN COMPANY JUl 0 6 2016 Docket Nos.: 50-321 NL-16-1 056 50-366 U.S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant Proposed Alternative HNP-ISI-ALT-05-03 Version 1.0 Regarding Reactor Pressure Vessel Flange Seal Leak-off Piping Testing Ladies and Gentlemen:
In accordance with the provisions of 10 CFR 50.55a(a)(3)(ii), Southern Nuclear Operating Company (SNC) hereby requests NRC approval of an American Society of Mechanical Engineers (ASME)Section XI code alternative for the leakage examination of the Class 1 reactor vessel flange leak-off piping for both Units 1 and 2 of the Edwin I. Hatch Nuclear Plant (HNP). This System Leakage Test of Class 1 pressure retaining components is cited in Table IWB-2500-1, Examination Category B-P, Item No. B 15.1 0, and is required to be performed during each refueling outage. Table IWB-2500-1, subparagraph IWB-5221 (a) indicates that system leakage tests shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power. The requested alternative (HNP-ISI-ALT-05-03, Version 1.0) is necessitated by the impracticality of conformance to these specifications of the subject Code.
The proposed alternative examination for Class 1 reactor vessel flange leak-off piping pressure test at HNP is predicated on the ASME Code Case N-805 which has yet to be approved by the NRC. However, the NRC has previously approved the methods described in ASME Code Case N-805 in response to similar requests from operators of commercial reactors in lieu of testing at pressures corresponding to 100% rated reactor power.
The Enclosure of this letter provides a detailed discussion of the need and justification for the use of an alternate approach for the Class 1 leakage testing of the subject vessel flange leak-off piping in lieu of the methods described in IWB-5221 (a) of the Code. An approval of the use of the proposed alternative is requested by February 1, 2017.
This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.
U.S. Nuclear Regulatory Commission NL-16-1056 Page 2 Res2cfl.s~
C. R. Pierce Regulatory Affairs Director CRP/RMJ
Enclosure:
Alternative HNP-ISI-ALT-05-03 Version 1.0 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Best, Executive Vice President & Chief Nuclear Officer Mr. D. R. Vineyard, Vice President- Hatch Mr. M. D. Meier, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Mr. G. L. Johnson, Regulatory Affairs Manager- Hatch RTYPE: CHA02.004 U.S. Nuclear Regulatory Commission Ms. C. Haney, Regional Administrator Mr. M. D. Orenak, NRR Project Manager- Hatch Mr. D. H. Hardage, Senior Resident Inspector- Hatch
Edwin I. Hatch Nuclear Plant Proposed Alternative HNP-ISI-ALT-05-03 Version 1.0 Regarding Reactor Pressure Vessel Flange Seal Leak-off Piping Testing Enclosure Alternative HNP-ISI-ALT-05-03 Version 1.0
Enclosure to NL-16-1056 Alternative HNP-ISI-ALT-05-03 Version 1.0 Plant Site-Unit:
Edwin I. Hatch Nuclear Plant (HNP) - Units 1 and 2 Interval/Interval Dates:
5th lSI Interval, January 1, 2016 through March 1, 2026 Requested Date for Approval and Basis:
Approval is requested by February 1, 2017, to permit performance of the proposed alternative pressure test during the 24th Refueling Outage of HNP Unit 2. This proposed alternative would also apply to all the remaining refueling outages during the Fifth lSI Interval for HNP Units 1 and 2.
ASME Code Components Affected:
The ASME Code components that are affected are the Unit 1 and Unit 2 Reactor Pressure Vessel Flange Seal Leak-off Piping. The piping configuration between units is slightly different. A description of the piping follows.
Unit 1 NPS 1" and 3/8" Reactor Pressure Vessel Flange Seal Leak-off Piping The 1" piping is A-312 TP 304 stainless steel, schedule 80. Design pressure is 600 psig at 850°F.
The 3/8" tubing is A-213 GR TP 304 or 316. Wall thickness is 0.065". Design pressure is 600 psig at 850°F.
Design pressure at 563°F is 1118 psig.
Unit 2 NPS 1", 3/4"and 3/8" Reactor Pressure Vessel Flange Seal Leak-off Piping The 1" piping is SA-1 06 GR 8 carbon steel, schedule 160. Design pressure is 900 psig at 850°F.
The 3/4" piping is SA-106 GR 8, schedule 160. Design pressure is 900 psig at 850°F.
The 3/8" tubing is SA-213 GR TP 304 or 316. Wall thickness is 0.065". Design pressure is 900 psig at 850°F.
Design pressure at 575°F is 1250 psig.
Applicable Code Edition and Addenda
ASME Section XI, 2001 Edition through the 2003 Addenda from January 1, 2016-November 30, 2017.
ASME Section XI, 2007 Edition through the 2008 Addenda from December 1, 2017 and is currently scheduled to end March 1, 2026.
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Enclosure to NL-16-1 056 Alternative HNP-ISI-ALT-05-03 Version 1.0 Applicable Code Requirements:
ASME Section XI, 2001 Edition 2003 Addenda System Leakage Test of Class 1 pressure retaining components per Table IW8-2500-1, Examination Category 8-P, Item No. 815.10 to occur each refueling outage. As referenced in Table IW8-2500-1, subparagraph IW8-5221 (a) indicates that system leakage tests shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.
ASME Section XI, 2007 Edition 2008 Addenda System Leakage Test of Class 1 pressure retaining components per Table IW8-2500-1, Examination Category 8-P, Item No. 815.10, is to occur each refueling outage, and Item No. 815.20 is to occur once per interval. As referenced in Table IW8-2500-1, subparagraph IW8-5221 (a) indicates that system leakage tests shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.
Background and Reason for Request:
Southern Nuclear has determined, through the use of industry operating experience that HNP is susceptible to an issue identified in the industry regarding compliance with ASME Section XI for examination of the reactor vessel flange leak-off piping.
HNP Units 1 and 2 both have reactor vessel flange leak-off piping that is used to detect leakage past the inner seal-ring. The reactor pressure vessel flanges are sealed with two concentric metal seal-rings designed to permit no detectable leakage through the inner or outer seal at any operating condition, including heating to operating pressure and temperature at a maximum rate of 100°F/h and cold hydrostatic pressure testing at the pressure specified in the ASME Code. To detect a lack of seal integrity, a 1-inch vent tap is located between the two seal-rings. A monitor line is attached to the tap to provide an indication of leakage from the inner seal-ring seal. This monitor connects to a pressure switch that will alarm to the main control room if pressure increases to 600 psig. In the event through wall leakage was present on this line, it is possible that pressure would not reach the 600 psig required to actuate the alarm. The performance of VT-2 testing would give reasonable assurance any external degradation or corrosion would be identified long before any through wall leakage could occur. The piping on both units is completely contained within the drywell. Leakage beyond this subject piping could be identified via increases in drywell pressure and temperature, as well as increases in the drywell leakage collection sump systems. A typical configuration of the vessel flange leak-off piping in question is shown in Attachment 1.
IW8-5221 (a) indicates that the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power. IW8-5222(a) and IW8-5222(b) indicate that the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor startup with E-2
Enclosure to NL-16-1 056 Alternative HNP-ISI-ALT-05-03 Version 1.0 IWB-5222(a) noting that the visual examination shall, however, extend to and include the second closed valve at the boundary extremity.
HNP Unit 1 and 2 configurations are such that the reactor vessel flange leak off lines are not capable of being pressure tested at normal reactor operating pressure, 1045 psig, unless the inner 0-ring seal fails or is intentionally failed.
Pressurizing the lines externally would apply pressure on the 0-rings in a direction opposite its design. This could move the 0-ring from its normal position against the channel wall of the reactor vessel flange potentially affecting the 0-ring leak tightness and requiring maintenance. The only other viable option would be a design change to perform a Code compliant system leakage test. The option to intentionally fail the inner seal of the vessel flange to establish normal operating pressure and temperature on the leak-off piping is not considered a viable option due to the increased dose that would result from the need to replace the inner seal. The dose required to fail the inner 0-ring, tension and de-tension the RPV head is estimated at 5 Rem. Additionally, development and implementation of a modification that would facilitate installation of a means to isolate the piping (e.g., threaded plug for testing prior to RPV head removal) is not considered viable due to the potential introduction of foreign material into the vessel during implementation and during each subsequent installation of the isolation device. This operation would subject individuals performing these tasks to considerable dose (estimated at 2 Rem) and heat stress without a commensurate increase in public health and safety.
Accordingly, these options are not considered to be viable.
Proposed Alternative and Basis for Use:
In accordance with the provisions of 10 CFR 50.55a(a)(z)(2), HNP proposes to examine the Class 1 portion of the leak detection system consisting of the accessible portions of the RPV head flange 0-ring leak-off piping during each refueling outage. The leak-off piping shall be examined using the VT-2 visual examination method and will be performed by certified VT-2 examiners. The test shall be conducted at ambient conditions after the refueling cavity has been residing at its normal refueling water level of approximately 22 feet 4 inches above the reactor vessel flange for at least four hours when the piping is subjected to the static head pressure that exists when the reactor cavity is filled and the closure head is removed. A static pressure of approximately 9.7 psig is expected to be experienced at the reactor pressure vessel head flange. This method is consistent with what has been done in recent years under previously approved Alternative HNP-ISI-ALT-18 which was used for the subject piping in the 41h lnservice Inspection Interval.
The Class 1 portion of piping originating from the reactor vessel flange is required to be examined. The Unit 1 Class 1 boundary stops on the downstream side of 1B21-F071 and 1B21-F063 and includes pressure switch 1B21-N002. The Unit 2 Class 1 boundary does not contain any boundary valves and includes pressure switch 1B21-N002. An excerpt from the Unit 1 and Unit 2 Nuclear Boiler System P&ID is contained in Attachments 2 and 3, respectively. The reactor vessel leak-off piping subject to the proposed alternate examination method is highlighted on the drawings. There are no E-3
Enclosure to NL-16-1 056 Alternative HNP-ISI-ALT-05-03 Version 1.0 sections of Class 2 piping associated with the reactor vessel flange leak-off line.
Both the Unit 1 and Unit 2 piping are contained completely within the drywell.
For segments of the line that are inaccessible for direct VT-2 visual inspection, examination will include inspection of the surrounding areas below the line for evidence of leakage as permitted by IWA-5241 (b) of both the ASME Section XI Code, 2001 Edition 2003 Addenda and 2007 Edition through 2008 Addenda. A small portion of the piping, specifically the connection between the piping and the vessel is covered by insulation. A depiction of the piping covered by the insulation is shown in Attachment 4. The remaining portion of the piping is not insulated.
In lieu of the requirements of IWB-5222(b), a VT-2 visual examination of the accessible areas will be performed each refueling outage on the piping subject to the static pressure from the head of water when the reactor cavity is filled for at least four hours. The reactor vessel flange seal leak-off piping is essentially a leakage collection/detection system and would only function as a Class 1 pressure boundary in the event of a failure of the 0-rings, which isolates the flange, seal leak-off piping from reactor coolant system operating pressure. Any significant leakage due to a failed 0-ring would be expected to clearly exhibit water accumulation that would be discernible during the proposed alternate VT-2 visual examination that will be performed. The static head developed with the flange seal leak-off piping filled with water will allow detection of any gross indications in the leak-off piping.
If the inner 0-ring should leak during the operating cycle, it will be identified through the alarm of a pressure switch in the main control room. Upon receiving an alarm, operator actions will involve monitoring: {1) drywell floor drain leakage per site procedures; (2) drywell dome area temperature on the remote shutdown panel; (3) drywell temperature per site procedures; (4) drywell pressure through multiple indications; and (5) drywell fission product monitors.
If any monitoring actions indicate outer seal failure, operators are directed per the annunciator response procedure to Technical Specification 3.4.4, RCS Operational Leakage. Similarly, should the inner 0-ring leak during the operating cycle and a thru wall leak of the reactor vessel flange leak-off piping exist, leakage will be detected in the same manner described above for leakage resulting from outer seal failure. In addition, there is no site-specific history of degradation associated with either unit's vessel flange leak-off piping.
Since there is reasonable assurance that the proposed alternate examination will detect gross indications of leakage should any exist from this piping and that the examinations will occur every refueling outage, SNC requests authorization to use the proposed alternative pursuant to 10 CFR 50.55a (a)(z)(2) on the basis that compliance with the specified requirement would result in hardship or difficulty without a compensating increase in the level of quality and safety.
Duration of Proposed Alternative:
The alternative is requested for the current Fifth lnservice Inspection Interval for both HNP Unit 1 and Unit 2, which began January 1, 2016 and will tentatively end on March 1, 2026 E-4
Enclosure to NL-16-1056 Alternative HNP-ISI-ALT-05-03 Version 1.0 Precedents:
- 1. Arkansas Nuclear One, Unit 2, Fourth Inspection Interval Alternative, Request for Relief from American Society of Mechanical Engineers (ASME) Code, Section XI- Request for Relief AN02-ISI-015, approved by the NRC in a letter dated June 27, 2013 (ADAMS Accession No. ML13161A241)
- 2. Callaway Plant, Unit 1, Third Inspection Interval Alternative, Proposed Alternative to ASME Section XI Requirements for Leakage Testing of Reactor Pressure Vessel Head Flange Leak-off Lines (Relief Request 13R-14, approved by the NRC in a letter dated August 13, 2013 (ADAMS Accession No. ML13221A091)
- 3. Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Third Inspection Interval Alternative, Request for Relief from the American Society of Mechanical Engineers (ASME) Code,Section XI, Reactor Vessel Head Flange Seal Leak Detection Piping- Relief Request No.
49, approved by the NRC in a letter dated April 4, 2013 (ADAMS Accession No. ML13085A254) 4 . Diablo Canyon, Units 1 and 2, Third Inspection Interval Alternative, Request for Approval of an Alternative to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI Pressure Test Requirements for Class 1 Reactor Vessel Flange Leak-off Lines, approved by the NRC in a letter dated September 12, 2013 (ADAMS Accession No. ML13192A354)
- 5. Dresden, Units 2 and 3, Fifth Inspection Interval Alternative, Request for Relief for Exemption from Pressure Testing Reactor Pressure Vessel Head Flange Seal Leak Detection System, approved by the NRC in a letter dated September 30, 2013 (ADAMS Accession No ML13258A003)
- 6. Vermont Yankee, Fourth Inspection Interval Alternative, Alternative to System Leakage Test for the Reactor Pressure Vessel Head Flange Leak off Lines, approved by the NRC in a letter dated March 1, 2013 (ADAMS Accession No. ML13055A009)
- 7. Edwin I. Hatch Nuclear Plant, Fourth Inspection Interval Alternative, HNP-ISI-ALT-18 Version 1.0 Alternative to System Leakage Test for the Reactor Pressure Vessel Head Flange Leak off Lines, approved by the NRC in a letter dated February 12, 2014 (ADAMS Accession No. ML14038A192)
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Enclosure to NL-16-1 056 Alternative HNP-ISI-ALT-05-03 Version 1.0
References:
- 1. ASME Code Case N-805, Alternative to Class 1 Extended Boundary End of Interval or Class 2 System Leakage Testing of Reactor Vessel Flange 0- ring Leak Detection System was issued to the 2010 Edition of the ASME Section XI Code and is listed in Supplement 6 for Code Cases. However, Code Case N-805 has not been approved by the NRC and is not identified in Regulatory Guide 1.147, lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1.
Status:
Pending NRC approval.
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Enclosure to NL-16-1 056 Alternative HNP-181-ALT-05-03 Version 1.0 ATTACHMENT t Slmplifted Schematic
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Enclosure to NL-16-1 056 Alternative HNP-ISI-ALT-05-03 Version 1.0 ATIACHMENT2 Excerpllrom Hatch Unit 1 P&ID H16062
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