ML20235R880

Forwards Revised Pages to Util 870710 Response to NRC Re safety-related Piping Supports at Facility.Changes Do Not Affect Results or Conclusions of Verification Program in

ML20235R880
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/15/1987
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
TAC-65726, NUDOCS 8707210470
Download: ML20235R880 (8)

Text

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F- Em [ David W. Cockfield Vice President, Nuclear July 15, 1987 i

1 Trojan Nuclear Plant j Docket 50-344 License NPF-1

. U.S. Nuclear Regulatory Coramission l ATTN: Document Control Desk Washington DC 20555 i

Dear Sir:

Support Design Verification l We would like to clarify our July 10, 1987 response to your letter of July 9, '1987 concerning safety-related piping supports at the Trojan Nuclear Plant. In our j response we stated a total of 479 supports had been reviewed and verified. It has subsequently been determined that a total of 453 supports were reviewed and evalu- l ated because 26 supports were reviewed as part of two categories and were counted l I

twice. These 26 supports were reviewed as part of the safety-related pipe snubber and restraint anchorages and structural members category as well as the dfnamic loads category. Some of the snubber anchor &gcc were, in actuality, piping restraint anchorages and this has been clarified on the " Summary of Pipe Support Verification i Action Plan". Minor clarifications are also being made on Pagco 3 and 5 of Attach- l ment 2 to our July 10 letter. These changes do not affcet the results or conclu- l' sions of the verification program as described in our July 10, 1987 letter.

We have revised the first page, Attachment 1, and Pages 3 and 5 of Attachment 2 to our July 10 letter to reflect these changes and have included it as an attachment to this letter. Picase replace the appropriate pages of the July 10 letter with the attached pages.

Sincerely, ,

Attachment c: Mr. John B. Martin Regional Administrator, Region V  ;

U.S. Nucicar Regulatory Commission i

j Mr. R. C. Barr NRC Resident Inspector Trojan Nucicar Plant Mr. David Kish, Director State of Oregon f' Department of Energy I-Subscribed and sworn to before me this 15th day of July 1987.

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David W. Cockfield Vice President, Nuclear l'

July 10, 1987 1

Trojan Nuclear plant Docket 50-344 License NpF-1 j

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555 ,

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Dear Sir:

i support Desir,n Verification ]

l Pursuant to the Nucicar Regulatory Commission (NRC) letter of July 9, 1987 {

I and in accordance with Part 50.54(f) of Titic 10 of the Code of Federal Regulations, this letter is provided to address the NRC concerns regarding the safety-related supports at the Trojan Nuclear Plant. The NRC letter expressed concerns regarding the original design, the derivation of the original design loads, and original design calculations for Main Steam System supports designed by the Architect-Engineer (A-E) for Trojan. The NRC concerns also extend to other cafety-related system supports designed by Trojan's A-E. Portland Cencral Electric Company (PCE) had similar concerns and has directed a thorough review of this issue. A description of this issue, the scope of the review performed, and the results of the review are provided below.

During the 1987 refueling outage, a discrepancy was noted when main steam line support EBB-1-1-SS-81 (SS-81) was inspected. The discrepancy con- I sisted of separation between the baseplate and the baseplate grout, and l between the grout and the concrete wall in several locations. The dis- l crepancy was evalucted by the PCE Civil Engineering Branch of Nucicar Plant Engineering. During their evaluation of the condition, design critoria for j the support were reviewed and it was determined that the design of the i support anchorage was inadequate for the specified dynamic load. 1 This problem has been identified, investigated, and is being resolved. An action plan was developed by PGE and the Trojan A-E to verify the design ]

of affected supports. A description of this action plan is provided in i Attachment 1. A total of 453 supports have been reviewod and evaluated l l (the verification of 15 pipe anchors is still in progress). It was determined 10 supports, for which the A-E Civil Engineering Group was responsibic, were inadequately designed for the originally specified bounding dynamic loads (loads due to turbine trip from 100 percent power).

These supports are all on the Main Steam System and were all designed and installed late in construction in 1975 to account for the turbine trip dynamic loads. This design problem is limited to the A-E Civil Engineering Group and specifically to the Main Steam System designs performed in 1975.

Although these supports did not meet the original design criteria, they have functioned acceptably under numerous actual loading conditions from turbine trips from full power. ,

121 S W Salmon Street, Portana O'egon 97204

Trojan Nuclear plant Document Control Desk Docket 50-344 July.10, 1987 License NPF-1 Attachment 1 '!

page 1 of 4 l

SUPPORT VERIFICATION ACTION PLAN To investigate and resolve the design load deficiency for SS-81, the Architect-Engineer (A-E) was contacted. The A-E could not determine if the turbine trip dynamic load of 83.2 kips had been properly included in the support design. The A-E's investigation determined the design of this support was the responsibility of their Civil Engineering Group.

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The A-E, under direction from pGE, took action to identify the root cause of the problem and to develop an action plan to determine the extent of the problem. The A-E determined the deficiency was isolated to their Civil Engineering Group and proposed an action plan to review all the supports for which the Civil Engineering Group was responsible in addition to the Main Steam System safety-related supports. The review included the 64 safety-related supports in the Main Steam System, 33 safety-related pipe anchors, 265(1) safety-related pipe snubber and restraint anchorages and I structural members and five non-safety-related pipe support structures containing a total of 41 individual supports. While the original design documentation for these supports was incomplete, the verification plan developed documentation demonstrating the adequacy of the as-built configuration based on the support design criteria.

Since the deficiency for SS-81 was a failure to account for the dynamic load in the support design, pCE expanded the scope of the program to include an analysis of all 76(1) safety-related supports for other sys- I tems for which dynamic load analyses had boon performed. These supports were fully designed by a pGE subcontractor with the exception of 26 which were partially designed by the A-E Civil Engineering Group. These addi-tional supports were reviewed to confirm that the deficiency was limited to the Civil Engineering Group. Thus, the plan encompassed a total of 453 supports. The results of the verification plan to date are described I below:

a. Main Steam System Supports.

Of the 64 safety-related main steam supports, 34 were designed entirely by a pGE subcontractor, and 30 were jointly designed by the subcon- l tractor (hardware) and the A-E (anchorage and structure). Fifty-four 1 of these supports were verified to be acceptable as-is. '

i (1) 26 safety-related supports are included in both categories:

265 safety-related pipe snubber and restraint anchorages ,

and structural members.

76 safety-related supports for systems with dynamic load l

analyses.

Trojan Nuclear Plant Document Control Desk f Docket 50-344 July 10, 1987 License NPF-1 Attachment 1

' Page 2 of 4 The other 10 supports were determined to be inadequately designed for the originally specified dynamic loads. The original loads were based on a conservatively applied pressure ramp taken from a turbine genera-tor similar to Trojan's. This method of deriving the original dynamic loads was unnecessarily conservative, but was standard practice in the early 1970s. During the verification effort, the A-E reanalyzed the dynamic loads for the Main Steam System supports inside Containment based on actual turbine trip data from Trojan startup testing, and the supports were reevaluated using the new dynamic loads. Three of the 10 supports were verified to be acceptable without modification based on the revised dynamic loads. Seven of the supports were modified by PGE utilizing the revised dynamic loads in the design process. It is likely the original design of these supports could have performed satisfactorily for the revised loads but PGE felt insufficient design cargin would have existed. In conclusion, the Main Steam System supports meet the design bases for the system.

b. Civil-Designed Pipe Anchors.

The A-E evaluated the design of all safety-related civil-designed pipe  ;

anchors (33). These supports are on the Safety Injection, Chemical and l Volume Control, Reactor Coolant Loop Vent / Drains, Pressurizer Spray, Residual Heat Removal, and Reactor Coolant Pump Seal Water Systems. To date, 18 of these supports have been verified to be satisfactory. The verification of the design of the remaining 15 is still in progress.

c. Civil-Designed Pipe Snubber and Restraint Anchorages.

TheA-Everifiedthedesignofallsafety-relatedcivil-designedpipe snubber anchorages (265)( ). These supports are on the Chemical and  !

Volume Control, Component Cooling Water, Pressurizer Relief and Safety, Pressurizer Spray, Reactor Coolant Instrumentation, Reactor Coolant Pump Seal Water, Reactor Coolant System Drain, Residual Heat Removal, Safety Injection, and Steam Generator Blowdown and Sample Systems.

These 265 supports were verified to be satisfactorily designed.

d. Civil-Designed Pipe Support Structures.

The A-E verified the designs for five non-safety-related civil-designed pipe support structures. Two of these structures provide support for (1) 26 safety-related supports are included in both categories:

265 safety-related pipe snubber and restraint anchorages and structural members.

76 safety-ralated supports for systems with dynamic load analyses. I

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-- Trojan Nucicar Plant Document Control Desk Docket 50-344 July 10, 1987 License NPF-1 Attachment 1 Page 3 of 4 moisture separator reheater relief valvo discharge lines, one for the main steam bypass line, and two for main steam stop valves. These five structurcs contaia a total of Al individual pipe supports and have been verified to be acceptable.

e. Supports With Dynamic Load Calculatio,ns. l The A-E identified 229 supports on systems for which dynamic load cal-culations had been performed. Of these, 153 are not safety-related and 76(1) are safety-related. The designs for all 76(1) safety-related I supports were verified to be satisfactory. These supports are on the pressurizer relief valve discharge line and the main steam lines to the turbine-driven auxiliary feedwater pump.

(1) 26 safety-related supports are included in both categories:

265 safety-related pipe snubber and restraint anchorages and structural members.

76 safety-related supports for systems with dynamic load analyses.

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Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987 License NpF-1 Attachment 2 Page 3 of 7 root cause determination. The review is being conducted on the following designs:

a. All safety-related main steam line pipe supports (" design verification").

b. All of the civil-designed pipe anchors inside Containment

("derign verification").

c. All of the civil-designed pipe snubber anchorage and structural members located inside Containment.

d. All civil-designed pipe support structures in the main steam and l main steam relief systems in the Turbine Building.

e. A sample of all pipe support calculations with dynamic loadings.

Calculations in progress to date do not indicate that.any member  ;

would have failed. For the main steam line supports this ccnclusion is supported by satisfactory performance under actual load conditions.

4. Description of Concern Why do the four main steam supports need design modification? Any changes in loads or acceptance criteria? What are the modifications?

Licensee's Response The four main steam supports are EBB-1-1-SS-81 (B loop),

EBB-1-1-SS-86 (A loop), EBB-1-2-SS-88 (C loop), and EBB-1-2-SS-92 (D loop). These four etructures are used to support the large hydraulic snubbers for the main steam lines.

Initially, it was determined that the four main steam supports required modifications to meet the original design loads. However, upon review of the original design loads, it was determined that they were inappropriate and overly conservative and new design loads were calculated (see responses to Concerns 1, 2, and 6). The four main steam cupports were not able to meet the newly calculated design loads within the original criteria limits. Therefore, modifications were made to restore the intended margins.

The design loads were changed by recent calculations performed in 1987. The dynamic loads were reduced by a factor of approximately 3.1 for SS-81 and SS-86, and by a factor of approximately 1.6 for SS-88 and SS-92, as a result of the dynamic analysis performed in

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