ML19343C407

Submits Summary & Schedule for Followup Response to Infractions of NRC Requirements,Per IE Bulletins 79-02,07 & 16.Generic Analysis Has Been Completed Showing That Typical Branch Connections Contain Acceptable Stresses

ML19343C407
Person / Time
Site:	Brunswick
Issue date:	02/25/1981
From:	Dietz C CAROLINA POWER & LIGHT CO.
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
BSEP-81-0440, IEB-79-02, IEB-79-07, IEB-79-14, IEB-79-2, NUDOCS 8103240076
Download: ML19343C407 (6)

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C.:rcuna Nwer a Li~ht C r.m eany Brunswick Steam Electric Plant P. O. Box 10429 Southport, NC 28461 February 25, 1981 s

FILE: B09-13514 ' '/ / .

SERIAL: BSEP/81-0440 g 47

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Mr. James P. O'Reilly, Director ..N U. S. Nuclear Regulatory Commission Region II, Suite 3100

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(8 101 Marietta Street, N.W.

Atlanta, GA 30303 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 & 2 LICENSE NOS. ' b and DP -

DOCKET NOS. 0-32jAND 0-324 FOLLOW-UP RESPONSE TO INF - IONS 0 ,r.C REQUIREMENTS

Dear Mr. O'Reilly:

In our letter (BSEP/80-2139) dated December 31, 1980, we committed to provide a comprehensive summary and schedule for remaining work by February 25, 1981.

This letter constitutes our report.

Our December 31 letter reported we had completed Items 1-6, 8 and 11 of our review outline (reference Attachment A). The results of Items 7, 9 and 10 are discussed below.

A comprehensive review and tabulation (Pipe Stress Analysis Summary Tables) of all safety-related piping was undertaken. This tabulation provides a line-by-line checklist of analyses performed and seismic commitments, as well as a status of the IE Bulletin 79-02, 79-07 and 79-14 efforts. These Pipe Stress Analysis Summary Tables have been completed by our A/E and we are presently final'. zing review of them. As expected from a re-examination of this magnitude, some inconsistencies / potential problem areas were revealed. The following paragraphs provide a short description of each category of incon-sistency/ potential problem along with its associated operational monsiderations. Our plans and schedules for resolving these items are also discussed. The anclysis portion of the below categories has been divided into a Phase I and Phase II program. In general, the Phase I effort will resolve items on a generic or typical basis while Phase II will resolve those items requiring a case by case analysis.

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Mr. Jamts P. O'Reilly 2 February 25, 1981 A. Lines Originally Seismically Analyzed, but not Included in IE Bulletin 79-07 Efforts This situation was originally mentioned in our November 28, 1980 letter to ycu under "Results to date" in item 1. At the completion of the seismic line review program a total of 38 isometrics (including the seven iden-tified in our November 28, 1980 letter) were identified as not having been reanalyzed per IE Bulletin 79-07. Of these 38, 21 (including the seven above) have been reanalyzed with results showing no short term action necessary. The remaining 17 isometrics are currently in reanalysis with no short term fixes anticipated in the judgement of the piping analysis group.

Completion of this reanalysis is part of Phase I. The walk-through and anchor bolt testing requirements of IE Bulletin 79-14 and 79-02 for these lines are scheduled for completion by the end of each unit's next refueling outage.

B. Vents, Drains, Instrucent Connections The effect of these small (1/2" to 1" piping) branch lines were not part of the original seismic analysis. A generic analysis has been completed showing typical branch connections contain acceptable stresses. This analysis also shows the branch connections do not affect the process piping for sizes four inch and larger. Additional analyses are in progress to determine the effect on computer analyzed seismic Category I lines and safety-related seismic Category I lines between 21" and less than 4" in diameter as part of our Phase I seismic piping effort. Based on these analyses to date and the engineering judgement of the analysis group, no short term fixes are anticipated on these lines.

C. Unanalyzed Loads Due to Valve Eccentricity Approximately 25 motor-operated valves were found not to have been analyzed for eccentric loadings. Most lines in this category are large enough to be essentially independent of eccentric valve loads. The smaller lines are supported close to the valves which will reduce any effect of eccentric loading. It is the judgement of our piping analysis group that additional loads as a result of valve eccentricity will not result in any short-term fix items. Confirmation of this conclusion is part of our Phase I work.

The Reactor Recirculation System is also to be reanalyzed by GE considering valve eccentricity and will be completed as a part of the Phase I program.

Also during your Mr. L. Modenos' inspection of January 22 and 23, 1980, at our A/E's office, it was noted that material properties, valve location and '

type were not verified. We agreed at that time to verify these items by QA documents. We have since identified the motor-operated valves (MOV) and, air-operated (A0V) valves and completed field verification of location, orientation and type. No discrepancies were found with our inspection. A further verification of proper valve installation by QA documentation review will be performed along with the pipe material properties QA documentation verification.

This QA documentation verification is scheduled to be completed as a part of the Phase I program. This schedule is acceptable based on the results of the field verification program.

Mr. Jam;s P. O'Railly 3 February 25, 1981

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D. Verification of Acceptable Containment Penetration Nozzle Loads Some (approximately 30 for Unit Nos. I and 2 combined) containment penetra-tions were identified which require design verification for new loads resulting from the IE Bulletin 79-07 reanalysis. The majority of the containment penetrations have been verified with no short-term action required, and in the judgement of our analysis group the few remaining will not require short-term action. This design verification is scheduled for completion as part of Phase I.

E. Vendor Supplied Piping and Vendor - A/E Interface Piping For some vendor supplied equipment and A/E designed interconnecting piping, the method of seismic analysis is not evident. Industry practice at the time this equipment was procured indicates a very low probability that vendor computer analysis would have been the basis for the seismic design. The items in this category include small diameter piping on such items as diesel generators, HPCI-RCIC turbines and standby instrument air compressor. Examination of the actual sizes and configuration of this piping revealed that the runs are either very short or supported well within the requirements of the ANSI B31.1 spacing tables. Review of safety-related lines 2i" and larger will be conducted as part of Phase II.

F. Small Nozzle Loads on Safety-Related Components Nozzle loads from computer analyzed small piping connections may not have been reconciled with the equipment design. These nozzles will require evaluation. In the interim, however, our piping support analysis group has evaluated this situation and anticipate no short-term fix items as a result of this effort. This small nozzle situation is very similar to the vents, drains, and instrument connection discussed in Category B above.

Completion of this item is scheduled with Phase II.

G. Seismic Requirement Inconsistencies There were several lines where seismic requirements were unclear and required additional review to reconcile inconsistencies or misinter-pretation of the FSAR, CP&L generated Q-list, and the original plant stress report. This situation was mentioned in our previous letter dated December 31, 1980. Of these questionable lines, or portions of lines, the ones that had no analysis performed were reviewed first and were discussed in one of the above categories or do not have seismic requirements. The line-by-line tabulation will be updated to reflect these resolutions. The remaining questionable lines will be reviewed and resolved in conjunction with our Phase I program. If during the resolution process of the above categories any short-term deficiencies are found, you will be notified and our

, Technical Specifications will be invoked.

We plan to complete Phase I by July 31, 1981 and Phase II by March 31, 1982.

This is acceptable based on the limited potential to effect the operability of system piping as discussed above. Due to the nature of the Phase II effort, its completion date may require review. If so, we' will inform you of the change and the basis for it.

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Mr. Jamas P. O'Reilly 4 February 25, 1981

.I In regard to our bulletin requirements and LER commitments for the CRD System supports, the following summary is provided:

1. All CRD supports inside the drywell were inspected, reanalyzed and modified on both units to meet "short-term fix" criteria, prior to startup from our last refueling outages.

2. As stated in our December 31, 1981 letter, we had completed reanalysis for all essential CRD supports and the results were under review. This review process has determined that 10 supports failed the short-term fix (STF) requirements, and they were reported by telecon to your Mr. Candle Julian on February 23, 1981, and is being reported as LER 2-81-25. You were also advised that a group of 60 supports were still under review since the original analyses performed on them included excessive conservatism. The 10 supports which failed the short-term fix criteria and the 60 supports still under review have been modified to meet short-term fix requirements.

Beside: the 70 supports outside the drywell discussed above, 358 supports outside the drywell were found to meet the short-term fix criteria.

3. All CRD system anchor bolts (both essential and nonessential) were tested and repaired prior to startup frem our last refueling outages. Flexure analysis on the CRD base plates is complete for all essential CRD baseplates. Completion of baseplate flexure analysis on CRD piping not essential to safe shutdown is scheduled for completion as a part of the Phase II program.

As stated in our December 31, 1980 letter, approximately 300 additional supports per unit were identified for anchor bolt testing (IEB-79-02). A further review of these supports found apprcximately 15% had been upgraded by installation of wedge type anchors and the drawings were in the process of being revised.

Another 5% were found to be in the A0G Building on piping not presently required for operation. A testing program was re-established and is still in progress on the accessible supports. Some of the inaccessible supports have been tested during a recent short maintenance outage.

Over 50% of the newly identified supports needing testing have been tested.

Anchor bolt testing of these supports have not resulted in inoperability of any safety-related piping due to their as-found condition. One line was declared inoperative during a plant shutdown when the support bolts were broken during testing.

We plan to complete the accessible portion of these identified supports as a part of the Phase I program and complete the inaccessibles during the next refueling outages. We will continue to test as many inaccessibles as practical during any unscheduled outages. The remaining supports were installed by the original constructor and have had a low failure rate. All anchor bolts installed by Reactor Controls, Inc. , which experienced a high failure rate, have been tested and repaired. This schedule for completion of testing is acceptable based on the results to date. Technical specifications will be invoked as necessary during the testing program.

In our December-31, 1981 letter referenced above, it was noted that a separate sign-off for support weld size verification was not included in our original as-built inspection program. Support welds were inspected originally as part of

Mr. Jam:s P. O'Railly 5 February 25, 1981 r?

the Plant Construction QA program. Based on this original inspection and the need to minimize personnel exposure, we plan to verify these support weld sizes by a sampling program. This program will involve reinspecting a random sample lot of support welds and will be designed to give a high confidence level in the results. If during this verification program, any short-term fix items are found, you will be notified and our Technical Specification LCO's will apply.

This program will be completed as a part of the Phase I program.

As is evident per the preceding, our concerted effort to assure compliance to the subject bulletins is continuing and should be completed as outlined herein.

Very truly yours, C. R. Dietz, General Manager Brunswick Steam Electric Plant SWP/mcg Attachment cc: Mr. R. A. Hartfield Mr. V. Stello, Jr.

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1. . Perform a page by page review of the FSAR, listing all references to Seismic Class I or any implication thereof.

2. Prepare a comprehensive Pipe Stress Analysis Summary table for each safety-related P&ID. Safety-related status is based on the following sources:

a. CP&L Q-List

b. BSEP FSAR

c. CP&L's response to IE Bulletin 70-07

d. UE&C's Final Unit 2 Stress Report The purpose of this table is to provide a line-by-line checklist of analyses. performed vs. seismic commitments and to provide a detailed status of the As-Built, Pipe Support, and IE Bulletin efforts.

3. ' Provide a yellow line mark-up of each safety-related P&ID to ensure that all lines are listed on the summary table.

4. Provide a line by line markup of each safety-related P&ID to identify applicable isometric drawing references and ensure that every line has been analyzed.

5. Review the status of all pipe supports on a line-by-line basis.

6. -Review the CP&L Q-List on a line-by-line basis.

7. Provide verification on a line-by-line basis that As-Built information was received.

8. ' Itemize FSAR commitments on a line-by-line basis.

9. Verify on a line-by-line basis that all nozzle loads, penetrations, valve eccentricities etc. have been considered.

.10. Verify if fire proofing (Pyrocrete) has been used which would significantly increase deadweight.

11. . Review the Unit 2 Final Stress Report for consistency with those lines analyzed for IE Bulletin 79-07.

Attachment A

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