ML20235J396

Responds to NRC 870709 Ltr on Concerns Re safety-related Supports at Plant,Including Original Design,Derivation of Original Design Loads & Original Design Calculation for Main Steam Sys Designed for Plant

ML20235J396
Person / Time
Site:	Trojan File:Portland General Electric icon.png
Issue date:	07/10/1987
From:	Cockfield D PORTLAND GENERAL ELECTRIC CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
TAC-65726, NUDOCS 8707150530
Download: ML20235J396 (12)

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=_= Portland General ElectricCompany David W. Cockfield Vice President. Nuclear July 10, 1987

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Trojan Nucicar Plant Docket 50-344 License NPF-1 l

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

Dear Sir:

Support Design Verification Pursuant to the Nuclear Regulatory Commission (NRC) Ictter of July 9, 1987 and in accordance with Part 50.54(f) of Title 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 j NRC concerns also extend to other safety-related system supports designed j by Trojan's A-E. Portland Conoral Electric Company (PGE) 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 outago, a discrepancy was noted when main steam line support EDB-1-1-SS-81 (SS-81) was inspected. The discrepancy con-sisted of separation betwoon the baseplato and the baseplato grout, and betwoon the grout and the concrete wall in several locations. The dis-crepancy was ovaluated by the PCE Civil Engineering Branch of Nuclear Plant Engineering. During their ovaluation of the condition, design criteria for the support were reviewed and it was determined that the design of the support anchorage was inadequate for the specified dynamic load.

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 Attachment 1. A total of 479 supports have been reviewed and evaluated (the verification of 15 pipe anchors is still in progress). It was determined 10 supports, for which the A-E Civil Engineering Group was responsible, were inadequately designed for the originally speciflod bounding dynamic loads (loads due to turbino trip from 100 porcent 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 Engincoring 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 turbino trips from full power. k 8707150530 870710 k \

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Portland General ElectricCorpsy Document Control Desk July 10, 1987 Page 2 To ensure proper implementation of the verification action plan, a PGE

- management team was sont to the A-E's offices to perform a quality 1 assurance surveillance. The team was hesded by the Panager of the Nuclear -l Quality Assurance Department and included the Branch Manager of Mechanical  !

Engineering and the Branch Manager of Nuclear Regulation. The objectives of the surveillance were to verify the A-E had adequately defined the problem, had identified and was taking appropriate corrective action, and that the documentation for the corrective action program was in place and was being complied with. The team concluded these objectives were being met.

On June 22 and 23, PGE and the A-E met with three NRC reviewers at the A-E I offices. The NRC reviewers raised 10 concerns to which the A-E and PGE l have responded. These concerns and responses are provided as' Attachment 2.

I The investigation of this issue has been comprehensive end thorough. PGE l

has been carefully monitoring the A-E's progress and results and is con-vinced the support verification plan has been thorough and has been cor-rectly performed. The verification plan constitutes a short-term action to demonstrate the acceptability of supports to perform their safety-related function. In the long-term, PGE intends to verify designs for other. safety-related supports. The scope of work and schedule for completing-this review is under evaluation and will be determined by September 1, 1987. l The design verification for 15 pipe anchors in Containment is not expected i to be complete until July 14, 1987. Plant startup will not commence until- l these 15 verifications are completed and are determined to be satisfactory.  !

Upon completion of this verification action plan, PGE has reasonable f assurance that safety-related supports will be properly designed and the Trojan Nuclear Plant will be safe to return to operation.

Sincernly, f 4 Attachments c: Mr. John B. Martin .;

Regional Administrator, Region V l U.S. Nuclear Regulatory Commission Mr. R. C. Barr NRC Resider.t Inspector .

Trojan Nucioar Plant Mr. David Kish, Director 1

tate of Oregon j . k. . .artment of Energy l l

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v. C N Subscribed and sworn to before me this,10th day of July 1987.

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b f' Notary Public of Ore'gon '

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i Trojan Nucicar plant Document Control Desk July 10, 1987 j Docket 50-344 Attachment 1 l License NpF-1 page 1 of 3 l l

j SUpp0RT VERIFICATION ACTION PLAN i

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To investigate and resolve the design load deficiency for SS-81, the j Architect-Engineer (A-E) was contacted. The A-E could not determine if the j turbine trip dynamic load of 83.2 kips had been properly included in the l

l sup- port design. The A-E's investigation determined the design of this t support was the responsibility of their Civil Engineering Group, l t'

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The A-E, under direction from pGE, took action to identify the root cause l )

of the problem and to develop an action plan to determine the extent of thia j problem. The A-E determined the deficiency was isolated to their Civil '

Engineering Group and proposed an actinn plan to review all the supports 1

for which the Civil Engineering Group was responsible in addition to the Main Steam System safety-related supports. The review. included the u t

64 safety-related supports in the Main Steam System, 33 safety-related pipe I anchors, 265 safety-related pipe snubber anchorages and structural members l end five non-safety-related pipe support structures containing aptotal of l 41 individual supports. While the original design documentation (for these supporth was incomplete, the verification plan developed documen'tstion j demonstkwting the adequacy of the as-built configuration based on the support design criteria. <

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Since the deficiency for SS-81 was a failure to account for the dynamic l load in the support design, PGE expanded the scope of the program to include an analysis of all 76 sa gty-related supports.for other cystems j for which dynamic load analysesi had been performed. These supports were These additional supports were reviewed l l ,dysigned by a pCE subcontractor.

. itol confirm that the deficiency was limited to the Civil Engineering Group.

Thusf the plan encompassed a total of 479 supports. The results of the l

l verification plan to date are described below:

s

a. Main Steam System Supports. l Of the 64 safety-related main steam supports, 34 were designed entirely ,]

by a pGE subcontractor, and 30 were jointly designed by the XI subcontractor (hardware) andL 3 ha A-E (anchorage and structure) .

Fifty-fourofthesesupportqhereverifiedtobeacceptabicas-is'.

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' The other 10 supports were determined to be inadequately designed for.  !

the originally specified dydanic loads. The original loads were based l on a conservatively applied pressure ramp taken from a turbine genara-l l

' tor similar to Trojan's. This method of deriving the original dynamic I loads wds unnecessarily conservative, but was standard practice in the \ ,d

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early 1970s. During the verification effort, the A-E reanalyzed the s dynamic loads for the Main Steam System supports insids Containment l based on actual turbine trip data from Trojan startup testing, and the .

supports were reevaluated using the new dynamic loads. Three of the l

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Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987 Licanse NPF-1 Attachment 1 Page 2 of 3 l 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 j margin would have existed. In conclusion, the Main Steam System j supports meet the design bases for the system,

b. Civil-DesiRned Pipe Anchors. J The A-E evaluated the design of all safety-related civil-designed pipe anchors (33). These supports are on the Safety Injection, Chemical and 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 i verification of the design of the remaining 15 is still in progress.

c. CivC Designed Pipe Snubber Anchorages.

The A-E verified the design of cll safety-related civil-designed pipe I snubber anchorages (265).

These supports are on the chemical and

l Volume Control, Component Cooling Water, Pressurizer Relitf and Snfety, Pressurizer Spray, Reactor Coolant Instrumentation, Reactor Coolant Pump Seal Water, Reactor Coolant System Drain, Realdual Heat Removal, Safety Injection, and Steam Generator Blowdown and Ssmple 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 mois< tre separator reheater relief valve discharge lines, one for the main steam bypass line, and two for main steam stop valves. These five structures contain a total of 41 individual pipe supports and have been verified to be acceptable,

e. Supports With Dynamic Load Calculations (not des 1Rned by Civil.

Engineering Group).

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 ~,6 are safety-related. The designs for all 76 safety-related sup-ports 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 feodwater pump.

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Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987 License NPF-1 Attachment 2 page 1 of 7 NRC CONCERNS ON SUpp0RT DESIGN ISSUE

1. Description of Concern What is considered as "dcoign verification" by the Architect-Engineer (A-E) for 50 safety-related main steam supports designed by the subcontractor. ]

Licensee's Response The group of 50 safety-related main steam supports included designs solely by the subcontractor, combination designs by the A-E civil and the subcontractor, and designs by the A-E' civil only. The 50 safety-related main steam supports were originally designed based upon loads calculated by the A-E.

The A-E's current " design verification" effort (1987) of these supports consisted of a review and reconciliation. New dynamic loads had been calculated in June 1987 based upon actual turbine trip test data, more recent regulatory guidance, and improved modeling techniques. The new design loads for the supports were compared with the original design loads. If the new design load was less than or equal to original load, and if the structure appeared adequate for the new design load, the support was considered acceptable. If the new design load was larger than the original design load, a structural evaluation was performed by the A-E to qualify the support. This evaluation consisted of engineering judgement for small load increases or a detailed calculation for larger load increases. The evaluations addressed the following items:

a. Bolts,

b. Base plates,

c. Wolds,

d. Member sizos, and

e. Components. )

2. Description of Concern What is considered as " design verification" by the A-E for 14 safety-related main steam supports inside Containment designed by the A-E Civil Croup.

Trojan Nuclear plant Document Control Desk i Docket 50-344 July 10, 1987

.j License NpF-1 Attachment 2 l page 2 of 7

,icensee's L Response 1

In 1975, 14 safety-related main steam line supports inside_Contain-ment were added to the original design for the dynamic load result- 1 ing from a turbine trip. These 14 supports were added after the design of the other 50 main cteam line supports. These 14 supports were designed by the A-E Civil Group and were apparently based upon )

design calculations that had not been performed (see Concern No. 3). )

1 Design verification for these 14 safety-related main steam line  !

supports inside the Containment consisted of the following:

a. New design loads were calculated for the turbine-trip event (see response to Concerns No.1 and 6) . 1

b. New calculations were performed based upon the new design loads I to evaluate the structural adequacy of the supports. The  !

calculations considered the following items:

)

1 (1) Bolts, J l

(2) Base plates, (3) Welds, (4) Member sizes.

c. The support calculations will be documented, checked, and approved in accordance with the existing A-E procedures and the A-E Quality Assurance program.  !

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3. Description ef Concern j l

Why did the original calculations of supports designed by the A-E Civil Group contain only stiffness calculations but not stress l (strength) calculations?

Licensee's Response l The investigation,has not revealed why only stiffness calculations can be found for certain of the 14 main steam line supports designed f by the A-E Civil Group. Due to the time interval since the stiff-l ness calculations were performed, it is not expected it will ever be known why the structural calculations were not performed.

However, the A-E is performing a review of pipe-support anchorages and structural members to verify design adequacy and to assist in l'

1

1 1

-Trojan Nuclear plant Document Control Desk l Docket 50-344 July 10, 1987 )

License NPF-1 Attachment 2 )

page 3 of 7 )

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q root cause determination. The review is being conducted on the following designs l l

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

verification"), o

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

(" design verification").

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c. All of the civil-designed pipe snubber anchorage and structurel membero located inside Containment. l

d. A sample of the civil-designed pipe support structures in the i ma.4.n steam and 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 ., ember would have failed. For the main steam line supports this coaclusion is supported by satisfactory performance under actual load {

conditions.

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

Licensco'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 I (D loop). TF;. four structures are used to support the large hydraulic sr.ibi..rs for the main steam lines.

Initially, it was determined that the four main steam supports q required modifications to meet the original design loads. However, 4 upon review of the original design loeds, it was determined that they were inappropriate and overly conservative and new design loads were calculated (sco responses to Concerns 1, 2, and 6). The four main steam supports were not able to meet the newly calculated i design loads within the original criteria limits. Therefore, I modifications were made to restore the intended margins.

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

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Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987-d 1

Licence NpF-1 Attachment 2 page 4 of 7 I

June 1987. Although the seismic loads showed a small general l increase from a seismic reanalysis performed earlier in the year, I l

the largest changes (reductions) in support loads were still attributable to the dynamic. effects due to the turbine trip transient analysis, l

The acceptance criteria for load combinations and allowable stress limits in structural steel were not changed. Due to its availabil-ity and the fact that tha old drill holes could be used, a dif-ferent type of anchor bolt was selected for the support redesign.

ASTM A-193 grouted threaded rods were used in the support anchorage redesign in lieu of the grouted rock anchors that were originally q used. The acceptance criteria for the grouted threaded rods were j based on current capacity development guidance for this type of anchorage.

The following is a specific description of the support modifications:

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a. Supports SS-81, SS-88, and SS-92. l The anchor bolts were changed from grouted 1 inch diameter rock anchors to grouted 1-1/4 inch diameter ASTM A-193 threaded rods  ;

in order to rostore anchorage capacity margins. The 1 inch j thick steel base plate was replaced with a 1-3/4 inch thick base l plate to increase stiffness and capacity. The sinubber support l was changed from a gusset plate stiffened 12 UF 36 bracket assembly to a 10 x 10 x 1/2 inch thick structural tube braced frame to provide overall increased capacity and reduce localized l stresses. New grout was placed between the base plate and concrete wall,

b. Support SS-86.

A new frame made up of a 4 x 12 x 1/2 inch structural tube column and beam war. connected to the existing stiffened 12 WF 36 snubber support to redistribute a portion of the load from the

( existing support to the new frame, thereby substantially reducing the loads on the existing anchcr bolts. The new frame also significantly increased torsional resistance of the existing snubber support beam. The new column 3/4 it.ch thick steel base plate was anchored with 3/4 inch diameter ASTM A-193 grouted threaded rods. New base plate grout was placed where necessary.

5. Description of Concern

• Describe status and findings of the A-E verification review of 50 safety-related main steam supports designed by the subcontractor.

Document Control Desk .j Trojan Nuclear Plant July 10, 1987 i Docket 50-344 License NpF-1, Attachment 2 l

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page 5 of 7 Licensee's Response I The verification, of the 50 safety-related main steam line supports is complete; no discrepancies or findings were noted. All supports were found to be acceptable. (See response to Concern No. 1).

6. Description of Concern Mcw were the dynamic loads related to turbine trip defined in the original design of he main steam lines. What is the AE's.conclu-sion on verification of this load and its effects on support design.

Licensee's Response The original design turbine trip dynamic loads, for the design of l the main steam piping and the 14 supports added in 1975, resulted l from a force-time history analysis. This analysis, which was com-pleted in 1975, utilized the pressure transient data from a test cf a similar General Electric turbine at another nucinar power plant.

The analysis (original dynamic load) was validated by testing at Trojan and found to be conservative (eg, for main steam support SR-86, the original design load was 83.2 kips and the measured value was 22.4 kips). The differences between the original calculated design loads and the test data were attributed to the differences between the nuclear power plant from which the original test data was obtained (a Boiling Water Reactor) and Trojan (a pressurized Water Reactor). The test data were always less than the calculated loads and the amount of conservatism was never questioned.

Subsequently, it has been determined that this analysis was overly l

conservative. The main steam piping inside Containment was recently I reanalyzed to calculate more realistic design loads. The new cal-culation was based upon the data from the Trojan 100 percent power turbine trip tests, more recent modeling techniques, and regulatory guidelines. The predicted turbine trip dynamic loads on all main steam supports inside Containment were reduced by this reanalysis.

7. Description of concern The A-E analysis for support SS-38 used a nonlinear interaction formula for combining shear and tension. The manufacturer recommends a linear interaction formula. Why did the A-E use I rather than T + V s 1?

I T I +I V i1 T (VALL, TALL VALL GALL /

1 Trojan Nuclear Plant Document Control Desk-Docket 50-344 July 10, 1987 ,

License NPF-1 Attachment 2 l Page 6 of 7 .

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Licensee's Response ]

The nonlinear interaction formula is an appropriate representation of shear-tension effects as demonstrated by test data. This formula was utilized for the evaluation resulting from IE Bulletin 79-02, regarding Seismic' Category I pipe support base plates which use f expansion anchors. A report of this evaluation (using the formula '

in question on Page 4) was transmitted to the Nuclear Regulatory Commission via a letter from Mr. D. J. Brochl, PGE, to Mr. R. H. j Engelken, NRC dated November 21, 1979. The NRC closed-out PGE action on IE Bulletin 79-02 in Inspection Report 86-05, dated j April 14, 1986. .

8. Description of Concern i PGE or the A-E must determine what the allowable loads are for the Williams' groutable (Hollow-Core) rock bolts.

Licensee's Response j l

The criteria for the design of the grouted-in rock anchors are:

a. Calculate the maximum tension load to be resisted by the rock anchor. ]

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b. Determine the embedment required to develop the tension load {

determined in Item a. above, based on the ultimate pull-out ]

capacity of concrete calculated, based on Paragraphs 1504 and )

1707 of ACI Code 318-63. l l

c. Select not-cross sectional area of rock bolt by limiting the j stress on the net section of 0.9 fy. ]

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9. Description of Concern 1

Snubber angularity, due to tolerance in installation, should be i considered b the design calculations for supports.  !

Need to provide justification for not including.

Licensee's Response Y

The Trojan Nuclear Plant pipe support design criteria (DC-11760-003, Rev. 0) limits strut and snubber offsets to the lesser of 5' or 2 inches. The design practice of the industry during the era of Trojan Nuclear Plant design did not consider these effects. The 5*

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tolerance would result in a maximum potential out of plane loading I

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Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987 License RPF-1 Attachment 2 Page 7 of'7 of 8 percent of the design load, and generally has a negligible effect on the structure.

10. Description of Concern 1

Calculations for the 265 snubbers inside Containment did not always .l contain enough information for a reviewer to determine'what the engineer / checker was accepting by comparing to a given standard j{

(example, a structure was accepted without any calculations or ]

discussion when lengths were longer and the member's sizes were smr11er). l Licensee's Response Further documentation will be performed of the engineering judgement made during review of the 265 snubber supports inside Containment.

This documentation will be added to the'A-E's calculation files.

The following were considered in the review:

a. Supports which are similar to one of the generic supports but i differ in some way (ie, member sizes, anchorage, welds, loads, lengths) are evaluated in the following fashion: i (1) Differences between the subject support configuration / load and the similar generic support are stated.

(2) The reviewer determined the effects of the differences based on judgement or based on a calculation. The reviewer then documented his conclusion,

b. pipe supports that are not similar to the generic type supports are evaluated in the following manner:

(1) If the load is small and the support configuration is simple, the support is reviewed by judgement after a review of the ana,horage, base plate, and structural members. The conclusion is documented by a statement.

(2) If a load is substantia) or support configuration is complex, a calculation is performed on critical structural items.

TDW/kal 1426P.687

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