ML13175A100
| ML13175A100 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 08/05/2013 |
| From: | Veronica Rodriguez Plant Licensing Branch 1 |
| To: | Pacilio M Exelon Nuclear |
| Lamb J, NRR/DORL/LPLI-2, 415-3100 | |
| References | |
| TAC ME9491 | |
| Download: ML13175A100 (7) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 August 5, 2013 Mr. Michael J. Pacilio President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, I L 60555
SUBJECT:
OYSTER CREEK NUCLEAR GENERATING STATION - RELIEF FROM THE REQUIREMENTS OF THE ASME CODE, RELIEF REQUEST NO. 15R-02 (TAC NO. ME9491)
Dear Mr. Pacilio:
By letter dated August 28, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12243A287), Exelon Generation Company (Exelon, the licensee) submitted Relief Request (RR) 15R-02 for the U.S. Nuclear Regulatory Commission's (NRC's) approval. The licensee proposed an alternative to certain requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI.
RR 15R-02 is requested for the fifth 1 O-year inservice inspection (lSI) interval at Oyster Creek Nuclear Generating Station (Oyster Creek), which commenced on January 15, 2013. The 2007 Edition through the 2008 Addenda of the ASME Code,Section XI, is the current Code of record at Oyster Creek. Specifically, pursuant to Title 10 of the Code Federal Regulations (10 CFR) paragraph 50.55a(a)(3)(ii), the licensee requested to use the proposed alternative on the basis that compliance with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
The NRC staff has reviewed the subject request and has concluded, as set forth in the enclosed safety evaluation, that the proposed alternative described in RR 15R-02 provides an acceptable level of quality and safety. Accordingly, the staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR SO.SSa(a)(3)(ii), and is in compliance with the ASME Code's requirements.
Therefore, the NRC staff authorizes the alternative described in RR 15R-02 for the fifth lSI interval at Oyster Creek, which began on January 15, 2013, and ends on January 14, 2023. All other ASME Code reqUirements for which relief was not specifically requested and approved in the subject request remain applicable, including a third party review by the Authorized Nuclear Inservice Inspector.
M. Pacilio
- 2 If you have any questions regarding this matter, please contact the Senior Project Manager, John G. Lamb at (301) 415-3100 or bye-mail at John.Lamb@nrc.gov.
Sincerely, Veronica Rodriguez, A g Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219
Enclosure:
Safety Evaluation cc w/enclosure: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION REQUEST FOR RELIEF 15R-02 FOR THE FIFTH 10-YEAR INSERVICE INSPECTION INTERVAL OYSTER CREEK NUCLEAR GENERATING STATION EXELON NUCLEAR DOCKET NO. 50-219
1.0 INTRODUCTION
By letter dated August 28, 2012, (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML12243A287), Exelon Generation Company (Exelon, the licensee) requested approval of Request for Relief (RR) No. 15R-02 from certain requirements of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code), under the provisions of Title 10 of the Code of Federal Regulations (10 CFR), paragraph 50.55a(a)(3}(ii}, for the fifth 10-year Inservice Inspection (lSI) Program for Oyster Creek Nuclear Generating Station (OCNGS).
2.0 REGULATORY EVALUATION
lSI of the ASME Code Class 1, 2, and 3 components is to be performed in accordance with Section XI of the ASME Code, and applicable addenda, as required by paragraph 50.55a(g} of 10 CFR, except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(g}(6}(i). The regulations in 10 CFR 50.55a(a)(3}, state, in part, that alternatives to the requirements of paragraph (g) may be used, when authorized by the U.S. Nuclear Regulatory Commission (NRC), if the licensee demonstrates that: (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.
The licensee has proposed alternatives from ASME Code requirements pursuant to 10 CFR 50.55a(a)(3}(ii}. The ASME Code of record for the OCNGS fifth 10-year lSI interval program is the 2007 Edition through the 2008 Addenda of Section XI of the ASME Code. The fifth 10-year lSI interval began on January 15, 2013, and is projected to end on January 14, 2023.
Enclosure
- 2
3.0 TECHNICAL EVALUATION
RR 15R-02 3.1 ASME Code Component(s) Affected (as stated by the licensee)
Code Class:
1
Reference:
Table IWB-2500-1 IWB-5200
[ASME Code] Examination Category:B-P Item Number:
B15.20 Descri ption:
Pressure Testing the RPV [Reactor Pressure Vessel] Head Flange Seal Leak Detection Line Component Number:
[ASME Code] Class 1 RPV Head Flange Seal Leak Detection Line Drawing Number:
BR 2002 Sh.1 3.2 ASME Code Requirements (as stated by the licensee)
[ASME Code,Section XI,] Table IWB-2500-1, Examination Category B-P, Item Number B15.20, requires certain [ASME Code] Class 1 pressure retaining components be subject to a system leakage test with a VT[visual testing]-2 visual examination in accordance with [ASME Code,Section XI,] Paragraph IWB-5220.
This pressure test is to be conducted once per interval. The pressure retaining boundary for the test conducted at or near the end of each inspection interval shall be extended to all [ASME Code] Class 1 pressure retaining components per [ASME Code,Section XI,] Paragraph IWB-5222(b).
3.3 Licensee's Basis for Relief Request (as stated by the licensee)
Exelon Generation Company, LLC (Exelon) is requesting a proposed alternative in accordance with 10 CFR 50.55a(a)(3)(ii) on the basis that compliance with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality.
The reactor vessel head flange leak detection line is separated from the reactor pressure boundary by one passive membrane, a Silver-plated O-ring located on the vessel flange.
A second O-ring is located on the opposite side of the tap in the vessel flange. This line is required during plant operation and will indicate failure of the inner flange seal O-ring.
The configuration of this line precludes manual testing while the vessel head is removed. The configuration of the vessel tap, combined with the small size of the tap and the high test pressure requirement prevents the tap from being temporarily plugged. Also, when the vessel head is installed, an adequate pressure test cannot
- 3 be performed due to the fact that the inner a-ring is designed to withstand pressure in one direction only. Due to the groove that the a-ring sits in and the pin/wire clip assembly (See Figure 15R-02.1 1), pressurization in the opposite direction into the recessed cavity and retainer clips would likely damage the a-ring and thus result in further damage to the a-ring.
Pressure testing of this line during the [ASME Code,] Class 1 System Leakage Test is precluded because the line will only be pressurized in the event of a failure of the inner O-ring. Purposely failing the inner O-ring to perform the [ASME] Code required test would require purchasing a new set of O-rings, additional time and radiation exposure to de-tension the reactor vessel head, install the new O-rings, and then reset and re-tension the reactor vessel head. This is considered to impose a hardship and burden on the [OCNGS.]
Based on the above, the [OCNGS] requests relief from the ASME [Code,] Section XI requirements for system leakage testing of the reactor vessel head flange seal leak detection line.
3.4 Licensee's Proposed Alternative Examination (as stated by the licensee)
A VT-2 visual examination on the [ASME Code,] Class 1 portion of the reactor pressure vessel flange leak detection line will be performed during each refueling outage when the RPV head is off and the head cavity is flooded above the vessel flange. The static head developed with the leak detection line filled with water will allow for the detection of any gross indications in the line. This examination will be performed each refueling outage as per the frequency specified by [ASME Code,Section XI,] Table IWB-2500-1.
3.5
NRC Staff Evaluation
The ASME Code requires that all ASME Code, Class 1 components within the reactor coolant system boundary undergo a system leakage test at or near the end of each inspection interval.
The licensee has proposed to perform a VT-2 visual examination on the ASME Code, Class1 portion of the RPV flange leak detection line each refueling outage when the RPV head is off and the head cavity is flooded above the vessel flange. The static head developed with the leak detection line filled with water will allow for the detection of any gross indications in the line. This examination will be performed each refueling outage, as per the frequency specified by ASME Code,Section XI, Table IWB-2500-1.
The reactor vessel head flange leak detection line is separated from the reactor pressure boundary by one passive membrane, a Silver-plated a-ring located on the vessel flange. The design configuration of this line prevents manual testing while the vessel head is removed. The configuration of the vessel tap, combined with the small size of the tap and the high test pressure requirement prevents the tap from being temporarily plugged. In addition, when the vessel head is installed, the ASME Code-required pressure test cannot be performed due to the design of the inner O-ring to withstand the nominal operating pressure at 100 percent rated 1 Figure 15R-02.1 is not included in this safety evaluation. See the licensee's submittal dated August 28,2012, for Figure 15R-02.
-4 reactor power in one direction only. Due to the groove that the O-ring sits in and the pin/wire clip assembly, pressurization in the opposite direction into the recessed cavity and retainer clips would likely damage the O-ring and more than likely result in damage to the subject O-ring.
To perform the system leakage test in accordance with the ASME Code requirements, the RPV Head Flange Seal Leak Detection Piping would have to be redesigned, fabricated, and installed.
This would impose a burden without a compensating increase in the quality or safety on the licensee. The piping in question was examined at the beginning and end of each refueling outage during the fourth 1 O-year lSI interval with no structural abnormalities or deficiencies being noted. Based on the visual VT-2 results during these unpressurized examinations, it is reasonable to conclude that, if significant service-induced degradation had occurred in the piping or within the flange seal leak-off piping and inner O-ring seal, evidence of it would have been detected by the examinations performed. The NRC staff determined that the licensee's VT-2 visual examinations performed each outage on the unpressurized subject piping as part of the ASME Code, Class 1 leakage test provides reasonable assurance of leak tightness of the subject components.
3.6 Duration of Relief RR 15R-02 is requested for the fifth 10-year lSI interval for OCNGS. The fifth 1 a-year lSI interval began on January 15, 2013.
4.0 CONCLUSION
As set forth above, the NRC staff has determined that authorizing alternatives pursuant to 10 CFR 50.55a(a)(3)(ii) for 15R-02 is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest given due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility. Accordingly, the staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3)(ii). Therefore, the NRC staff authorizes the alternatives to the specified ASME Code,Section XI, contained in RR No. 15R-02 for the duration of the OCNGS fifth 1 a-year lSI interval.
All other ASME Code,Section XI requirements for which relief was not specifically requested and approved in the subject RR remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principle Contributor: T. McLellan Date: August 5, 2013
M. Pacilio
- 2 If you have any questions regarding this matter, please contact the Senior Project Manager, John G. Lamb at (301) 415-3100 or bye-mail at John.Lamb@nrc.gov.
Sincerely, Ira!
Veronica Rodriguez, Acting Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219
Enclosure:
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