ML14038A192
| ML14038A192 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 02/12/2014 |
| From: | Robert Pascarelli Plant Licensing Branch II |
| To: | Pierce C Southern Nuclear Operating Co |
| Martin, Robert NRR/DORL 415-1493 | |
| References | |
| TAC MF3228, TAC MF3229 | |
| Download: ML14038A192 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 February 12, 2014 Mr C. R. Pierce Regulatory Affairs Director Southern Nuclear Operating Company, Inc.
Post Office Box 1295, Bin - 038 Birmingham, AL 35201-1295
SUBJECT:
EDWIN I. HATCH NUCLEAR PLANT, UNITS 1 AND 2 (HNP), SAFETY EVALUATION OF RELIEF REQUEST HNP-ISI-ALT-18 FOR THE FOURTH 10-YEAR INSERVICE INSPECTION INTERVAL (TAC NOS. MF3228 AND MF3229)
Dear Mr. Pierce:
By letter to the U.S. Nuclear Regulatory Commission (NRC), dated December 16, 2013, as supplemented January 13, 2014, Southern Nuclear Operating Company, Inc. (SNC, the licensee) submitted a request for alternative HNP-ISI-ALT-18 from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) at the Edwin I. Hatch Nuclear Plant, Units 1 and 2. Specifically, in accordance with Title 1 0 of the Code of Federal Regulations (1 0 CFR) 50.55a(a)(3)(ii), the licensee proposed an alternative to the inservice inspection requirement of IWB-5220 of the system leakage test of the reactor pressure vessel flange leak-off piping.
Based on the review of the information the licensee provided, the NRC staff concludes that the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety and the alternatives provide reasonable assurance of structural integrity and leak tightness of the subject piping. Therefore, the licensee's proposed alternative is authorized in accordance with 10 CFR 10 50.55a(a)(3)(ii) for the licensee's fourth 1 0-year in service inspection interval. All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the NRC staff remain applicable, including third-party review by the Authorized Nuclear lnservice Inspector.
Docket Nos. 50-321 and 50-366
Enclosure:
Safety Evaluation cc w/encls: Distribution via Listserv Sincerely,
}(-..
Robert P scarelli, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR ALTERNATIVE HNP-ISI-ALT-18, VERSION 1, REGARDING SYSTEM LEAKAGE TEST OF REACTOR PRESSURE VESSEL FLANGE LEAK-OFF PIPING SOUTHERN NUCLEAR OPERATING COMPANY, INC.
EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 DOCKET NOS. 50-321 AND 50-366
1.0 INTRODUCTION
By letter dated December 16, 2013 (Agencywide Documents Access and Management Systems (ADAMS) Accession No. ML13351A424), as supplemented by letter dated January 13,2014 (ADAMS Accession No. ML14013A194), Southern Nuclear Operating Company (the licensee) submitted for the U.S. Nuclear Regulatory Commission's (NRC) approval, a request for alternative (RFA) HNP-ISI-ALT-18. The licensee proposed an alternative to a certain requirement of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, for the Edwin I. Hatch Nuclear Plant (Hatch), Units 1 and 2. RFA HNP-ISI-AL T -18 relates to the inservice inspection (lSI) requirement of IWB-5220 for the system leakage test of the reactor pressure vessel (RPV) flange leak-off piping.
Specifically, pursuant to Title 10 of the Code of Federal Regulations {1 0 CFR) 50.55a(a)(3)(ii),
the licensee proposed an alternative to the system leakage test on the basis that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
2.0 REGULATORY EVALUATION
The 10 CFR 50.55a(g)(4) specifies that ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except the design and access provisions and the preservice examination requirements, set forth in the ASME Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 1 0-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code, incorporated by reference in 10 CFR 50.55a{b), 12 months prior to the start of the 120-month interval, subject to the conditions listed therein.
Enclosure The 10 CFR 50.55a(a)(3) states, in part, that alternatives to the requirements of paragraph (g) of 10 CFR 50.55a may be used, when authorized by the NRC, if the licensee demonstrates (i) the proposed alternatives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Based on the above, and subject to the following technical evaluation, the NRC staff finds that regulatory authority exists for the licensee to request and the NRC to authorize the alternative requested by the licensee.
3.0 TECHNICAL EVALUATION
3.1 The Licensee's Request for Alternative The component affected by this request is ASME Code Class 1, IWB-2500, Table IWB-2500-1, Examination Category B-P, Item No. B 15.1 0, leak-off piping of the RPV head flange seal leak detection and collection system. Below are the components for which the licensee proposed an alternative.
Hatch, Unit 1 The 1 inch piping is A-312, Grade TP 304 stainless steel, Schedule 80. The design pressure is 600 pounds per square inch gauge (psig) at 850 degrees Fahrenheit (°F). The 3/8 inch tubing is A-213, Grade TP 304 or TP 316.
Hatch, Unit 2 The 1 inch piping is SA-106, Grade B carbon steel, Schedule 160. The% inch piping is SA-106, Grade B, Schedule 160. The design pressure for the piping is 900 psig at 850°F. The 3/8 inch tubing is SA-213, Grade TP 304 or TP 316 with a design pressure of 600 psig at 850°F.
The Class 1 portion of piping originating from the reactor vessel flange is subject to a system leakage test. The Hatch, Unit 1, Class 1 boundary stops on the downstream side of 1 B21-F071 and 1 B21-F063 and includes pressure switch 1 B21-N002. The Hatch, Unit 2, Class 1 boundary does not contain any boundary valves, but includes pressure switch 1 B21-N002. The Hatch, Units 1 and 2, leak-off piping are contained completely within the drywell.
The code of record for the fourth 1 0-year lSI interval at Hatch, Units 1 and 2, is the 2001 Edition through 2003 Addenda of the ASME Code.
ASME Code,Section XI, IWB-2500, Table IWB-2500-1, Examination Category B-P, establishes requirements to conduct the system leakage test and the VT -2 visual examinations in accordance with IWB-5220 and IWA-5240, respectively, during each refueling outage. In accordance with IWB-5221 (a), the system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100 percent rated reactor power. In accordance with IWB-5222(a), the pressure retaining boundary during a system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup. The visual examination shall, however, extend to and include the second isolation (closed) valve at the boundary extremity.
The licensee proposed an alternative to IWB-5221 (a). The proposed alternative is to perform the VT -2 visual examinations of the accessible areas each refueling outage when the RPV flange leak-off piping is subjected to static pressure head of 9.7 psig for at least four hours during the reactor vessel cavity flood-up for refueling. The static pressure head of 9.7 psig is produced from the normal refueling water level of approximately 22 feet and 4 inches above the reactor vessel flange when the reactor cavity is flooded for refueling. For segments of the line that are inaccessible for direct VT-2 visual examination, the licensee will examine the surrounding areas below the line for evidence of leakage in accordance with IWA-5241(b). A small portion of the piping, specifically the connection between the piping and the vessel, is covered by insulation. The remaining portion of the piping is not insulated.
The licensee provided its basis for hardship or unusual difficulty caused by compliance with the system leakage testing requirement. The Hatch reactor vessel flange leak-off lines are used to detect leakage past the inner 0-ring seal. The RPV flanges are sealed with two concentric metal 0-rings designed to permit no detectable leakage through the inner or outer 0-ring seal at any operating condition. To detect a lack of seal integrity, a 1-inch vent tap is located between the two 0-rings. A monitor line is attached to the tap to provide an indication of leakage from the inner 0-ring seal. This monitor connects to a pressure switch that will alarm to the main control room (MCR) if pressure increases to 600 psig. The Hatch RPV configurations are such that the reactor vessel flange leak-off lines are not capable of being pressure tested at normal reactor operating pressure of 1060 pound per square inch absolute unless the inner 0-ring seal fails or is intentionally failed. Another option would be a design change to perform the ASME 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. In letter dated January 13, 2014, the licensee estimated the cumulative radiological exposure to be approximately 5 roentgen equivalent man per refueling outage for activities needed to perform the required leakage testing by failing the newly installed inner 0-ring. Additionally, development and implementation of a modification that would facilitate installation of a device, such as a threaded plug to isolate the piping, is not considered viable due to the potential introduction of foreign material into the vessel. Individuals performing these tasks would be subjected to considerable dose.
The licensee stated that there is no site-specific history of degradation associated with Hatch, Unit 1 or 2's vessel flange leak-off piping. In its letter dated January 13, 2014, the licensee confirmed that Hatch has not observed any degradation or corrosion of the RPV flange seal leak-off piping during inspections performed inside the drywell.
Duration of RFA HNP-ISI-ALT-18 is for the fourth 1 0-year lSI interval which commenced on January 1, 2006, and will end on December 31, 2015.
In its letter dated January 13, 2014, the licensee stated that testing of the RPV flange seal leak-off piping has not been performed for Hatch, Units 1 and 2, in a manner consistent with the requirements of the ASME Code,Section XI. Although the Hatch, Units 1 and 2, RPV flange seal leak-off piping are included in the scope of the VT-2 leak tests required by the ASME Code,Section XI, for Class 1 piping, the testing does not include pressurization of the seal leak-off piping necessary to allow identification of leakage from the RPV flange seal leak-off piping. The licensee stated that it became aware of its inadequacy of compliance as a result of the NRC enforcement action associated with failure of others to perform adequate VT -2 testing of the seal leak-off piping as required by the ASME Code,Section XI. The licensee stated that it incorporated plans to perform the ASME Code,Section XI, required system leakage test of seal leak-off piping during the upcoming Hatch refueling outage.
In its letter dated January 13, 2014, the licensee stated that pressurizing the RPV leak-off piping in the beginning of the refueling outage prior to removal of the RPV head is not possible due to design configuration. For pressurization, physical modification of the RPV flange leak-off piping would be required. Although the configuration of the two Hatch units is slightly different, both units would require modifications in order to allow pressurization of the seal leak-off piping prior to removal of the RPV head. Implementation of design modifications and performance of the system leakage test will result in significant radiological exposure. Although the time required for implementation of the design modification cannot be accurately estimated without first preparing the design, the dose fields in the Hatch, Units 1 and 2, drywells are estimated to be in the range of 30-100 milliroentgen equivalent man per hour (mrem/hr) and 14-80 mrem/hr, respectively, with the higher dose rates occurring in the vicinity of the reactor vessel flange.
In its letter dated January 13, 2014, the licensee stated that the Hatch, Unit 2, final safety analysis report, Section 7.6.7.2.3.5, states, "If both the inner and outer head seals fail, the leak is detected by an increase in drywell temperature and pressure." Additionally, leakage from both the inner and outer RPV flange seals will be detected by the reactor coolant system Leakage Detection Instrumentation required by Technical Specification Limiting Condition of Operation 3.4.5.a. Similarly, leakage from a 100 percent through wall leak in the RPV flange seal leak-off piping, concurrent with leakage of the RPV flange inner seal, would provide similar, if not identical, indications to the MCR (i.e., increased drywell temperature and pressure, concurrent with increased unidentified leakage to the dry floor sump).
3.2
NRC Staff Evaluation
The NRC staff has evaluated RFA HNP-ISI-ALT-18 pursuant to 10 CFR 50.55a(a)(3)(ii). The NRC staff focuses on whether compliance with the specified requirements of 10 CFR 50.55a(g),
or portions thereof, would result in hardship or unusual difficulty, and if there is a compensating increase in the level of quality and safety despite the hardship.
The NRC staff determined that it would result in hardship if the IWB-5221 (a) requirement for leak testing of the RPV head flange seal leak-off line piping is imposed upon the licensee. This hardship is due to the existing design and configurations of reactor vessel flange and flange leak-off line which makes system leakage testing of leak-off line piping at 100 percent rated reactor power either before or after the RPV head is removed unusually difficult. In order to test the subject piping according to the ASME Code after the vessel head is removed for refueling, the licensee would have to make design modifications to the existing RPV flange face. This could introduce foreign materials into the reactor, which would introduce a plant Foreign Material Exclusion program concern. Personnel carrying out these tasks would incur additional radiation dose, which would be an as-low-as-reasonably-achievable (ALARA) program concern.
In order to conduct the required system testing before the RPV head is removed for refueling, the licensee would have to install new test connections and valves to existing piping to be able to externally pressurize the leak-off piping. Personnel performing these tasks and conducting testing would be exposed to additional dose, which would be an ALARA program concern.
There also would exist significant occupational hazards from heat while the plant is at normal operating temperature during the beginning of the refueling outage. Similar modifications (i.e.,
installation of new connections and valves) to existing leak-off piping would be necessary if the licensee has to perform the ASME Code required leakage test after the new 0-rings and vessel head are installed following refueling. Personnel involved in these tasks would be exposed to additional dose, which would be an ALARA program concern. Furthermore, when the head is on and the subject pipe being externally pressurized for the purpose of conducting the ASME Code required leak testing of leak-off piping, the inner 0-ring seal may fail due to abnormal pressurization resulting in an unsuccessful test. Therefore, the NRC staff determines that the above items constitute a hardship.
The NRC staff finds the licensee's proposed system leakage test acceptable because the RPV leak-off piping will be subjected to the highest pressure that can be obtained without major design modifications to existing configurations of both the vessel flange face and the flange leak-off piping. Specifically, the leak-off line piping will be pressurized by the proposed static pressure head (9.7 psig) developed from the refueling water level above the reactor vessel flange during the vessel cavity flood-up, and will be examined by the IWA-5240 required VT-2 visual examinations after attaining the proposed test pressure. Any evidence of leakage, if it were originated from an existing flaw in the piping and its associated connections, would be detected by the VT-2.
Furthermore, the licensee stated that both Hatch units have no history of degradation in vessel flange leak-off piping. The NRC staff review of operating experience, including Hatch, has not identified any documented degradation due to stress corrosion cracking and fatigue in the vessel flange leak-off piping and associated welded connections.
The NRC staff determined that the licensee has sufficient leakage detection capabilities (i.e.,
detection of increase in drywell temperature and pressure as well as detection of an increase in unidentified leakage to the drywell floor sump) that provides warning to the control room operator in the 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. Even if, or in case, the proposed alternative was not effective in identifying a through-wall leak, if it were originated from an existing flaw in the subject piping, the plant existing leakage detection capability will be able to identify the leakage during normal operation and the licensee will take appropriate corrective actions.
Therefore, the NRC staff finds that the proposed system leakage testing using low test pressure, although it may not be as effective as using the normal operating pressure, is adequate to provide reasonable assurance of structural integrity and leak tightness of the RPV flange seal leak-off line piping.
4.0 CONCLUSION
As set forth above, the NRC staff determines that the proposed alternative provides reasonable assurance of structural integrity and leak tightness of the subject piping and that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii) and is in compliance with the requirements of the ASME Code,Section XI, for which relief was not requested. Therefore, the NRC staff authorizes the use of RFA HNP-ISI-AL T -18 at Hatch, Units 1 and 2, for the remainder of the fourth 1 0-year lSI interval which commenced on January 1, 2006, and will end on December 31, 2015. All other ASME Code,Section XI, requirements for which relief was not specifically requested and authorized herein by the staff remain applicable, including the third party review by the Authorized Nuclear In-service Inspector.
Principal Contributor: Ali Rezai Date of issuance: February 12, 2014
- memo transmitted SEdated 1/30/14 OFFICE N RR/LPL2-1 /PM NRRILPL2-1/LA NRRIDE/ESGB/BC NRR/LPL2-1/BC NAME RMartin SFigueroa GKulesa RPascarelli DATE 02/11/14 02/10/14 01/30/14
- 02/12/14