ML20195C216
| ML20195C216 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 11/09/1998 |
| From: | Thadani M NRC (Affiliation Not Assigned) |
| To: | Geoffrey Edwards PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR TAC-M69469, TAC-M69470, NUDOCS 9811170049 | |
| Download: ML20195C216 (6) | |
Text
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November 9, 1998 Mr. Garrett D. Edwards Director-Licensing, MC 62A-1 l
PECO Energy Company l
Nuclear Group Headquarters Correspondence Control Desk P.O. Box No.195 l
Wayne, PA 19087-0195 l.
SUBJECT:
SECOND REQUEST FOR ADDITIONAL INFORMATION (RAI) FOR PEACH BOTTOM ATOMIC POWER STATION (PBAPS), UNITS 2 AND 3, UNRESOLVED l
SAFETY ISSUE (USI) A-48 (TAC NOS. M69469 AND M69470) l
Dear Mr. Edwards:
By letter dated September 3,1997, PECO Energy Company (the licensee) responded to the l
Nuclear Regulatory Commission (NRC) staff's letter dated June 5,1997, requesting additional J
l Information needed to complete the review of the licensee's resolution of the USl A-46 program at PBAPS, Units 2 and 3. The NRC staff has reviewed the licensee's response and finds that the questions have not been fully resolved and mors information is needed in order to complete the 1
review of the USl A-46 program implementation. A second RAI is enclosed.
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We request that the licensee provide its response to the enclosed RAI within 60 days from the date of receipt of this letter (which was mutally acceptable) to support a timely completion of the NRC staff's action on the above open item. This RAI was discussed with your staff on October 14,1998.
Sincerely, origial signed by:
Mohan C. Thadani, Senior Project Manager Project Directorate 1-2 Division of Reactor Projects -l/ll Office of Nuclear Reactor Regulation l
Docket Nos. 50-277 and 50-278 l
Enclosure:
Second RAI j
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November 9,1998 Mr. Garrett D. Edwards Director-Licensing, MC 62A-1 PECO Energy Company Nuclear Group Headquarters Correspondence Control Desk P.O. Box No.195 Wayne, PA 19087-0195
SUBJECT:
SECOND REQUEST FOR ADDITIONAL INFORMATION (RAI) FOR PEACH BOTTOM ATOMIC POWER STATION (PBAPS), UNITS 2 AND 3, UNRESOLVED SAFETY ISSUE (USI) A 46 (TAC NOS. M69469 AND M69470)
Dear Mr. Edwards:
By letter dated September 3,1997, PECO Energy Company (the licensee) responded to the Nuclear Regulatory Commission (NRC) staff's letter dated June 5,1997, requesting additional information needed to complete the review of the licensee's resolution of the USl A-46 program at PBAPS, Units 2 and 3. The NRC staff has reviewed the licensee's response and finds that the questions have not been fully resolved and more information is needed in order to complete the review of the USl A-46 program implementation. A second RAIis enclosed.
We request that the licensee provide its response to the enclosed RAI within 60 days from the date of receipt of this letter (which was mutally acceptable) to support a timely completion of the NRC staff's action on the above open item. This RAI was discussed with your staff on October 14,1998.
Sincerely, Mohan C. Thadani, Senior Project Manager Project Directorate I-2 Division of Reactor Projects - t/11 Office of Nuclear Reactor Regulation Docket Nos. 50 277 and 50-278
Enclosure:
Second RAI
-cc w/ encl: See next page j
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.4 Mr. Garrett D. Edwards Peach Bottom Atomic Power Station, PECO Energy Company Units 2 and 3 4
cc:
J. W. Durham, Sr., Esquire Chief-Division of Nuclear Safety 1
Sr. V.P. & General Counsel PA Dept. of PECO Energy Company Environmental Resources l..
2301 Market Street,826-1 P.O. Box 8469 Philadelphia, PA 19101 Harrisburg, PA 17105-8469 PECO Energy Company ATTN: Mr. J. Doering, Vice President Board of Supervisors Peach Bottom Atomic Power Station Peach Bottom Township 1848 Lay Road R. D. #1 Delta, PA 17314 Delta, PA 17314 i
PECO Energy Company Public Service Commission of Maryland ATTN: Regulatory Engineer, A4-5S Engineering Division Peach Bottom Atomic Power Station Chief Engineer 1848 Lay Road 6 St. Paul Centre i
Delta, PA 17314 Baltimore, MD 21202-6f.06
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Resident inspector Mr. Richard McLean U.S. Nuclear Regulatory Commission Power Plant and Environmental Peach Bottom Atomic Power Station Review Division P.O. Box 399 Department of Natural Resources Delta, PA 17314 B-3, Tawes State Office Building Annapolis, MD 21401 Regional Administrator, Region i U.S. Nuclear Regulatory Commission Dr. Judith Johnsrud 475 Allendale Road National Energy Committee King of Prussia, PA 19406 Sierra Club 433 Orlando Avenue Mr. Roland Fletcher State College, PA 16803 Department of Environment 201 West Preston Street Manager-Financial Control & Co-Ownt-Baltimore, MD 21201 Affairs Public Service Electric and Gas A. F. Kirby, lll Company Extemal Operations - Nuclear P.O. Box 236 Delmarva Power & Light Company Hancocks Bridge, NJ 08038-0236 P.O. Box 231 Wilmington, DE 19899 Manager-Peach Bottom Licensing PECO Energy Company PECO Energy Company Nuclear Group Headquarters
- Plant Manager Correspondence Control Desk Peach Bottom Atomic Power Station P.O. Box No.195 1848 Lay Road Wayne, PA 19087-0195 j.
Delta, PA 17314
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SECOND REQUEST FOR ADDITIONAL INFORMATION PEACH BOTTOM ATOMIC POWER STATION. UNITS 2 AND 3 UNRESOLVED SAFETY ISSUE (USB A-46
- 1. In regard to the in-structure response spectra (IRS) that was used to resolve the relay outliers for host cabinets located in the radwaste/ turbine building at the 150-foot floor elevation, Reference 1 indicated the use of clipping of spectra in accordance with Appendix Q of EPRI NP-6041. As stated in the staff's previous request for additional information (RAl),
dated June 5,1997, the methodologies described in EPRI NP-6041 Report have not received approval from the staff for the analysis of safety-related systems and components, including the resolution of USl A-46 mechanical, electrical, and structural component outliers. The clipping of spectra leads to a reduction of the amplified spectrum at the elevation of interest and is, therefore, not acceptable to the staff. You are requested to reevaluate the relay outlier resolutions that have involved the use of this methodology.
2.
In regard to the adequacy of seismic demand determination, it appears that there are several buildings (emergency cooling tower, diesel generator building and circulating water pumping structure) where for elevations less than 40 feet above the ground level, the corresponding IRS appears to be higher than 1.5 times the GlP-2 Bounding Spectra, and, hence, is higher than 1.5 times the ground response spectrum (since, as indicated in Reference 1, the ground response spectrum is completely enveloped by the Bounding Spectrum). You stated in Reference 1 that Method A of GIP-2 was used to verify the seismic adequacy of the equipment in these buildings. We note, however, that if the IRS for elevations less than about 40 feet above ground level is greater than 1.5 times the ground response spectrum then the use of Method A being not consistent with page 4-16 of GIP-2, under " Adequacy and Limitations," which states, "The restrictions and limitations on use of the around response spectrum for comparison to the Bounding Spectrum and the GERS [ generic equipment ruggedness spectra] are based on the conditior, that the amplification factor between the free-field response spectrum and the in-structure response spectrum will not be more than 1.5,... " You are requested to provide a technicaljustification for the use of Method A for the above buildings where the IRS at elevations less than about 40 feet above grade is greater than 1.5 times the ground response spectrum.
- 3. For cases where the IRS exceeds 1.5 times the ground response spectrum, provide a list of equipment, or an appropriate number of equipment items, in terms of percentage, whose seismic verifications were based on Method A of GIP-2.
- 4. On page ia of Calc. No. PS-0912 (Reference 1), you stated that the specific work was not performed as nuclear safety related, but was performed, reviewed, and approved in accordance with the preparer's (Duke Engineering and Services) Quality Assurance Program.
Discuss the calculational method or assumption and demonstrate that it meets the GIP-2 criteria for cable trays.
- 5. In reference to Sample 1 (page 2 of Attachment A of Reference 1), the sketch indicated that the supports have only two bolts, one at the top and one at the bottom. Discuss the lack of l
moment resistance capacity in the direction of the cables, and for this support, indicate why it is adequate with no such moment resistance capacity.
ENCLOSURE l
l
The sketch also indicated that the upper horizontal strut members and the lower strut members are connected by rods (there is.no indication in the sketch about what this line represents). Discuss how the lateral load is resisted. Discuss whether the rod hanger type indicated in the sketch is within the scope of the database referenced in GlP-2.
Also in the above Sample 1, the tensile and shear capacities are provided only for anchor bolts. Discuss the anchorage capacity of the concrete and the amount of a safety margin against spillage (shear) at the interface between the anchor bolts and the concrete. Explain whether or not the capacrty of the bolts is limited by the bolt itself or the capacity of concrete.
- 6. In reference to Sample 2 (pages 3 & 4 of Attachment A), it appears that there are two distinct cable supports. However, the calculation does not appear to address the support on page 4, which is a ceiling hung support. Provide an explanation for this.
- 7. In reference to Sample 3 (pages 17 and 18 of Cal. PS-0912 and the corresponding sketch on page 5 of Attachment A), there appears to be an inconsistency in the welded connection detail provided. On pages 17 and 18 of Cal. PS-0912, it is indicated that the strut is welded into the flange of the structural steel beam, whereas on page 5 of Attachment A, the strut is indicated as welded to the web. You are requested to provide a clarification as to which one is correct and how it was actually analyzed. Explain also how you concluded that the welded connection is acceptable.
- 8. On the cover sheets of calculations 0912 and 0939, you checked the box which corresponds to "non-safety-related" as opposed to " safety-related," implying that the affected cable tray supports are not safety-related. Confirm that supports with a non-safety-related classification meet the GlP-2 criteria.
- 9. Discuss how spatial interaction between the cable trays and supports and other objects were evaluated. In particular, there are many cable trays supported by suspended rod hangers whose moment resisting capacity may be insignificant for lateral seismic motion (free to swing). Discuss how the potential for interaction was addressed for such cases. Provide a drawing for the longest rod hanger or other cable tray supports together with the location of objects that may interfere with the cable tray supports. Demonstrate that there are sufficient spaces for the cable supports to swing freely without impacting with adjacent equipment or structures such as walls or columns.
- 10. Provide an updated status and estimated schedules for the resolution of all the identified outliers.
3-REFERENCES
- 1. Letter to NRC Document Control Desk, from G. A. Hunger, Jr., PECO Energy Company,
" Peach Bottom Atomic Power Station. Units 2 and 3, Request for Additional Information Regarding Generic Letter 87-02 on the Resolution of Unresolved Safety issue A-46," dated September 3,1997.