ML112020459
| ML112020459 | |
| Person / Time | |
|---|---|
| Site: | Perry  |
| Issue date: | 07/26/2011 |
| From: | Michael Mahoney Plant Licensing Branch III |
| To: | Bezilla M FirstEnergy Nuclear Operating Co |
| mahoney, m NRR/DORL/LPLIII-2 415-3867 | |
| References | |
| TAC ME4373 | |
| Download: ML112020459 (8) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 26, 2011 Mr. Mark B. Bezilla Site Vice President FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant Mail Stop A-PY-A290 P.O. Box 97,10 Center Road Perry,OH 44081-0097 SUB..IECT: PERRY NUCLEAR POWER PLANT, UNIT NO.1 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE 10 CFR 50.55A REQUESTS IN SUPPORT OF THE THIRD 10-YEAR IN-SERVICE INSPECTION INTERVAL (TAC NOS. ME4373)
Dear Mr. Bezilla:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated January 24, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110320065), FirstEnergy Nuclear Operating Company (the licensee), submitted multiple Title 10 of the Code of Federal Regulations, Section 50.55a, requests in support of the third 10-year inservice inspection interval for NRC approval for the Perry Nuclear Power Plant, Unit No.1.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The speCific information requested is addressed in the enclosure to this letter. During a discussion with your staff on July 11, 2011, it was agreed that you would provide a response within 45 days from the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for NRC staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3867.
WtJ Michael Mahoney, oject Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-440
Enclosure:
Request for Additional Information cc w/encl: Distribution via Listserv
REQUEST FOR ADDITIONAL INFORMATION ON THE THIRD 10-YEAR INSERVICE INSPECTION INTERVAL PROPOSED ALTERNATIVES FIRSTENERGY NUCLEAR OPERATING COMPANY PERRY NUCLEAR POWER PLANT. UNIT NO.1 DOCKET NO.: 50-440 1.0 SCOPE By letter to the U.S. Nuclear Regulatory Commission (NRC) dated January 24, 2011, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110320065) FirstEnergy Nuclear Operating Company (FENOC, the licensee), submitted Proposed Alternatives IR-009, Revision 2, IR-013, Revision 2, IR-027, Revision 2, IR-043, Revision 2, IR-054, Revision 1, IR-056 , Revision 1, and PT-001, Revision 2, from the requirements of the American Society of Mechanical Engineers (ASME Code), Boiler and Pressure Vessel,Section XI, for Perry Nuclear Power Plant, Unit No.1 (PNPP). The proposed alternatives apply to the third 10-year inservice inspection (lSI) interval, in which the licensee adopted the 2001 Edition through the 2003 Addenda of ASME Code Section XI as the Code of Record.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a(a)(3),
states proposed alternatives to the requirements in 10 CFR 50.55a paragraphs (c) through (h) may be used when authorized by the Director of the Office of Nuclear Reactor Regulation. The applicant shall demonstrate that: (i) The proposed alternatives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Proposed Alternatives IR-009, IR-054 , IR-056 and PT-001, have been submitted under 10 CFR 50.55a(a)(3)(i), and IR-013, IR-027, and IR-043, have been submitted under 10 CFR 50.55a(a)(3)(ii).
The NRC staff with the assistance of Pacific Northwest National Laboratory (PNNL) has reviewed the information submitted by the licensee, and based on this review, determined the following information is required to complete the evaluation.
Enclosure
-2 2.0 REQUEST FOR ADDITIONAL INFORMATION 2.1 Proposed Alternative IR-013, Revision 2, ASME Code,Section XI, Examination Category C-G, Item CS.1 O. Pressure Retaining Welds in Pumps and Valves For ASME Code, Class 2, pump casing welds, the ASME Code requires 100 percent surface examination be performed on either the inner or outer surface of the weld on at least one pump from each group that has a similar design, size, function and service in a system. In accordance with 10 CFR 50.55a(a)(3)(i) or 10 CFR 50.55a(a)(3(ii), the licensee may propose an alternative provided that: (1) The licensee demonstrates that the proposed alternatives would provide an acceptable level of quality and safety; or (2) the licensee demonstrates that the ASME Code requirements are a hardship, or unusual difficulty, in complying with ASME Code examination requirements that are adequately described and the licensee can demonstrate that complying with the ASME Code requirements would result in no compensating increase in quality and safety.
The licensee has provided descriptions of the pumps, which are located in the concrete flooring, and would require disassembly for examination of the subject welds; however, the stated reason the pump would have to be disassembled is too general and insufficient to demonstrate it would provide a hardship or unusual difficulty.
Since the licensee requested the alternative under hardship they should discuss the specific causes of, or unusual difficulty, in support of an evaluation under 10 CFR 50.55a(a)(3)(ii). Examples of hardship or unusual difficulty include, but are not limited to: having to enter multiple technical specifications' limiting conditions for operations, as low as reasonably achievable (ALARA) concerns (including personnel exposure estimates), or creating significant hazards to plant personnel.
The second part of the licensee's proposal must show that, even if the pumps were to be disassembled and the subject welds were to be examined, these activities would not result in a compensating increase in the level of quality and safety. Discuss why the ASME Code-required examinations on the inside surface of the welds, requiring disassembly of the pumps, would not provide an increase in quality and safety, The licensee has provided a typical description and sketch of a pump casing. The deSCription implies that the floor is preventing examinations of all of the casing welds.
(a) Provide a sketch showing the restrictions. If the restrictions are associated with a pit, provide the approximate depth and annulus dimensions. Show the locations where the pump is in intimate contact with the supporting structure.
If the casing below the floor is completely encased in concrete, provide the approximate depth of the encased casing.
(b) Provide a discussion on the accessibility for remote visual examinations performed from the outside or inside casing surface (including welds below the floor).
-3 (c) Provide a percent estimate of a combined visual and surface examination of the pump casing.
(d) Identify the number of pumps in the group being represented by the residual heat removal pump A.
(e) Provide a discussion on the effects a crack pump casing weld would have on the functionality of the pump and the effects on the region outside the casing.
2.2 Proposed Alternative IR-027. Revision 2.ASME Code,Section XI. Examination Category D-A, Item D1.1 O. Welded Attachments for Vessels, Piping. Pumps, and Valves For welded attachments to ASME Code Class 3 pressure vessels, the ASME Code requires 100 percent visual vr-1, examination on one vessel from each group that has a similar design, function, and service in a system. In accordance with 10 CFR 50.55a(a)(3(ii), the licensee has proposed an alternative. When proposing an alternative under 10 CFR 50.55a(a)(3)(ii), the licensee must demonstrate that compliance with the specified requirements of the ASME Code would result in (1) hardship, or unusual difficulty, and (2) no compensating increase in the level of quality and safety would result. The licensee stated that the subject welds are covered in Pyrocrete, which is a hard and rigid material used for fire protection. In order to remove this material from the welds, cutting and chipping of the Pyrocrete would be required; however, this brief description of removal activities does not adequately demonstrate a specific basis for hardship or unusual difficulty. Provide a discussion on the specific causes of hardship, or unusual difficulty, in support of an evaluation under 10 CFR 50.55a(a)(3)(ii). Examples of hardship or unusual difficulty include, but are not limited to having to enter multiple technical specifications' limiting conditions for operations, ALARA concerns (including personnel exposure estimates), or creating significant hazards to plant personnel.
2.3 Proposed Alternative IR-043, Revision 2, ASME Code,Section XI, Examination Category B-M-1. Items B12.30 and B12.40, Pressure Retaining Welds in Valve Bodies When proposing an alternative under 10 CFR 50.55a(a)(3)(ii), the licensee must demonstrate that compliance with the specified requirements of the ASME Code would result in (1) hardship, or unusual difficulty, and (2) no compensating increase in the level of quality and safety. The licensee stated that no failures of the valve body welds have been experienced to date, no degradation mechanism has been identified for these welds, and further, the 2008 Edition of the ASME Code eliminated the surface and volumetric examinations for these welds. However, the licensee's statements do not satisfy the requirements for use of 10 CFR 50.55a(a)(3)(ii). Specifically, discuss why there is not a compensating increase in the level of quality and safety if the licensee were to inspect the subject valve body welds.
- 4 State the materials of construction for the gate valves listed in Table 2.3.1 below.
Table 2.3.1 - Examination Cateaorv 8-M-1 ASME Weld 10 Weld Type Code Item 812.30 1G33-F0101-SEAM Reactor Water Clean-Up 3" Gate Valve 812.40 1G33-F01 OO-SEAM Reactor Water Clean-Up 4" Gate Valve 812.40 1G33-F01 06-SEAM Reactor Water Clean-Up 4" Gate Valve 812.40 1G33-FOO01-SEAM Reactor Water Clean-Up 6" Gate Valve 812.40 1G33-FOO04-SEAM Reactor Water Clean-Up 6" Gate Valve 812.40 1E12-F0019-SEAM Residual Heat Removal, 6" Check Valve 812.40 1E51-F0013-SEAM Reactor Core isolation Cooling, 6" Gate Valve 812.40 1E51-F0063-SEAM Reactor Core isolation Cooling, 10" Gate Valve 812.40 1E51-FOO64-SEAM Reactor Core isolation Cooling, 10" Gate Valve 812.40 1E12-F0039A-SEAM Residual Heat Removal, 12" Gate Valve 812.40 1E12-F00398-SEAM Residual Heat Removal, 12" Gate Valve 812.40 1E 12-F0039C-SEAM Residual Heat Removal, 12" Gate Valve 812.40 1E 12-F0042A-SEAM Residual Heat Removal, 12" Gate Valve 812.40 1E 12-F00428-SEAM Residual Heat Removal, 12" Gate Valve 812.40 1E12-F0042C-SI;AM Residual Heat Removal, 12" Gate Valve 812.40 1E21-FOOO5-SEAM Low Pressure Core Spray, 12" Gate Valve 812.40 1E21-F0007-SEAM Low Pressure Core Spray, 12" Gate Valve 812.40 1E22-F0036-SEAM High Pressure Core Spray, 12" Gate Valve The latest Edition of 10 CFR 50.55a only lists the 2004 Edition of the ASME Code as the latest edition approved for use. State why it is appropriate to cite the unapproved 2008 Edition of ASME Code Section XI, or provide the date of a safety evaluation (SE) that allows the use of the 2008 Edition of ASME Code Section XI.
The hardship for examining these valves with surface or volumetric techniques is the amount of dose received from insulation removal and reinstallation, and the actual examination. Since the ASME Code only requires inspections of 8 of the 18 welds listed in Table 2.3.1, state if any of these valves are located in low dose areas and if there are other similar gate valves in these systems that can be examined that would not result in high personnel exposure as when examining the valves listed above.
-5 2.4 Proposed Alternative IR-054. Revision 1. ASME Code,Section XI, Examination Category B-D, Items B3.90 and B3.100. Full Penetration Welded Nozzles in Vessels The licensee proposed, in lieu of performing examinations on 100 percent of the reactor vessel nozzle-to-vessel welds and nozzle inside radius sections, to incorporate Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division 1", which requires a minimum of 25 percent of nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size. NRC Regulatory Guide 1.193, Revision 3, "ASME Code Cases Not Approved for Use," states that:
The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation dated December 19,2007 (ADAMS Accession No. ML073600374) regarding BWR Vessel and Internals Project {BWRVIP)-108: "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii, Electric Power Research Institute (EPRI) Technical Report 1003557, October 2002" (ADAMS Accession No. ML023330203) are met.
The evaluation demonstrating the applicability of the ASME Code Case shall be reviewed and approved by the NRC prior to the application of the Code Case.
The five criteria are related to the driving force of the probabilistic fracture mechanics (PFM) analyses for the recirculation inlet and outlet nozzles. It was stated in the December 19, 2007 SE that the nozzle material fracture toughness-related nil-ductility transition reference temperature (RTNDT}values used in the PFM analyses were based on data from the entire fleet of BWR reactor pressure vessels (RPVs). Therefore, the BWRVIP-108 report PFM analyses are bounding with respect to fracture resistance, and only the driving force of the underlying PFM analyses needs to be evaluated. It was also stated in the December 19, 2007 SE that, except for the RPV heat-up/cool-down rate, the plant-specific criteria are for the recirculation inlet and outlet nozzles only because the probabilities of failure for other nozzles are an order of magnitude lower.
FENOC provided their calculations and results, which meet the criteria set forth in Section 5 of the NRC SE mentioned above. However, there is a discrepancy in the specific values provided for the RPV inner radius {r} and wall thickness {t} at PNPP. The values provided in proposed alternative IR-054 for the RPV inner radius and wall thickness are 119 inches and 7.19 inches, respectively. The values provided in BWRVIP-108, Table 3-1 for the PNPP RPV inner radius and wall thickness are 120.2 inches (RPV inner diameter was provided in the table as 240.4-inches) and 6 inches, respectively. If the PNPP values stated in BWRVIP-108 are used, the licensee would not be in compliance with the SE for the recirculation outlet nozzle-to-vessel welds.
Verify and state the specific values for the RPV inner radius and wall-thickness provided in proposed alternative IR-054 and explain why there is an inconsistency in the values provided in the BWRVIP-108 report and those provided in your submittal dated January 24, 2011.
- 6 Additionally, verify and state the specific values for the nozzle inner and outer radius for the recirculation inlet and outlet nozzles provided in proposed alternative IR-054.
2.5 Proposed Alternative IR-056. Revision 1, ASME Code,Section XI. Examination Categorv B-N
- 1. Item B13.10, Interior of Reactor Vessel, and Examination Categorv B-N-2, Item B13.40, Welded Core Support Structures and Interior Attachments to Reactor Vessels The licensee is requested to withdraw any proposed alternative in IR-056 that applies the requirements of BWRVIP-183, Top guide Grid Beam inspection and Flaw Evaluation Guidelines. BWRVIP-183 is currently under review by the NRC staff, and it would not be appropriate to consider this portion of the alternative at this time.
July 26, 2011 Mr. Mark B. Bezilla Site Vice President FirstEnergy Nuclear Operating Company Perry Nuclear Power Plant Mail Stop A-PY -A290 P.O. Box 97,10 Center Road Perry,OH 44081-0097
SUBJECT:
PERRY NUCLEAR POWER PLANT, UNIT NO.1 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE 10 CFR 50.55A REQUESTS IN SUPPORT OF THE THIRD 10-YEAR IN-SERVICE INSPECTION INTERVAL (TAC NOS. ME4373)
Dear Mr. Bezilla:
By letter to the U.S. Nuclear Regulatory Commission (NRC) dated January 24, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110320065), FirstEnergy Nuclear Operating Company (the licensee), submitted multiple Title 10 of the Code of Federal Regulations, Section 50.55a, requests in support of the third 10-year inservice inspection interval for NRC approval for the Perry Nuclear Power Plant, Unit NO.1.
The NRC staff is reviewing your submittal and has determined that additional information is required to complete the review. The specific information requested is addressed in the enclosure to this letter. During a discussion with your staff on July 11, 2011, it was agreed that you would provide a response within 45 days from the date of this letter.
The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for NRC staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-3867.
Sincerely, IRA!
Michael Mahoney, Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-440
Enclosure:
Request for Additional Information cc w/encl: Distribution via Listserv DISTRIBUTION:
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