ML20116P172

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Forwards Responses to Questions Documented in NRC Relating to Util 920716 Submittal Re Seismic Qualification Issues,Per 920806 & 11 Telcons.Technical Evaluation Repts Encl
ML20116P172
Person / Time
Site: Brunswick  
Issue date: 11/17/1992
From: Mccarthy D
CAROLINA POWER & LIGHT CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20115A385 List:
References
NLS-92-266, TAC-M83211, TAC-M83212, NUDOCS 9211240379
Download: ML20116P172 (18)
v • d • e


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CP&L Carohna Power & t.lght Company 1

i NOV 171992 i,

SERIAL: NLS 92 266 4.

United States Nucisar Regulatory Commission ATTENTION: Document Control Desk Washington, DC 20555 4

3 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS 1 AND 2 I

DOCKET NOS. 50 325 & 50-324/ LICENSE NOS DPR 71 & DPR 02 FOLLOW UP OUESTIONS ON SEISMIC OUAllFICATION ISSUES (TAC NOS. M83211 AND M83212) he u pose of this letter is to provide written responses to Nuclear Regulatory Commission (NRC)

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staff questions related to seismic qualification issues at the Brunswick Steam Electric Plant, Unit 4

Nos.1 and 2. These questions, which were documented in an NRC letter dated August 25,1992, relate to Carolina Power & Light Company's (CP&L) July 10,1992 submittal. The questions were discussed with CP&L during telephone conferences on August 6,1992 and August 11,1992. The i

Company's responses to the NRC staff questions are enclosed.

Please refer any questions regarding this submittal to Mr. D. B. Waters at (919) 546 2710.

i Yours very truly, O. C. McCanhy

- Manager l

Nuclear Licensing Section i

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Enclosures cc:.

Mr. S. D. Ebneter Mr. R. H. Lo Mr. R. L. Prevatte 200029

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ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 j

NRC DOCKET NOS. 50 325 & 50 324 OPERATING LICENSE NOS. DPR 71 & DPR 62 FOLLOW-UP OUESTIONS ON SEISMIC OUALIFICATION ISSUES (TAC NOS M83211 AND M83212) of the NRC's August 25,1992 letter provides NRC staff comments on the operability criteria of piping. The following staff questions / comments relate to Study report M-020, which is i

referenced in the response to NRC Ouestion 1.F in CP&L's July 16,1992 submittal.

i NRC OUESTION 1:

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l The SRSS combination method for 2D responses is unacceptable to the staff. Provide clarification of the 1,38 factor used and why it was later found not to be required.

CP&L RESPONSE:

The Company's position on 2D SRSS analysis was provided to the NRC by letter dated May 29, 1979 (copy provided as Attachment 1). This letter also discusses the use of the 1.38 factor.

During the subsequent reanalysis performed afte' 5985 by CP&L, the predominant analysis method was 3D SRSS using the Regulatory Guide 1.92 r..adel and co directional combination.

l NRC QUESTION 2:

Confirm eat the use of Code Case N 411 is in compliance with the requirements of Regulatory Guide 1.84, Revision 26.

CP&L RESPONSE:

a The Company requested the use of Code Case N-411 by letter to the NRC dated May 22,1985, (copy provided as Attachment 2). Subsequent approval was received from the NRC by letter dated August 28,1985 (cory provided as Attachment 3).

4 NRC OUESTION 3-.

If the table is for Class 1 piping, what are the criteria for Class 2/3 piping, i

CP&L RESPONSE:

The table is for Seismic Class I, which was the UE&C designation for structures, systems, and components requiring seismic design. Seismic Class 11is non seismic. The Class 2/3 piping classification is an ASME designation; equivalent piping at the Brunswick Plant was decigned as l

Seismic Clas 1.

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NRC QUESTION 4:

Provide more spr:ific discussion on the OBE/DBE conversion factor of 1.2 to 2.0.

CP&L RESPONSE:

The 1.2 conversion factor was developed for conversion of OBE loads and stresses. The development of this factor was presented to the NRC by letter dated May 29,1979, (copy provided as Attachment 1). As documented in an NRC staff meeting summary dated June 12, 1979 (copy provided as Attachment 4), the NRC indicated acceptance of CP&L's method during a phone call on June 1,1979. The design turnover program for piping criteria uses specific DBE response spectra curves or a conversion factor of 2.0.

NRC OUESTION 5:

Frequencv cutoff at 20 Hz with no " missing mass" considered is clearly an unacceptable analysis procedure, CP&L RESPONSE:

The frequency cutc,ff at 20 Hz with no " missing mass" was based on the licensing basis for BNP.

Most of the UE&C calculations have been reanalyzed by CP&L. These calculations included a

" missing mass" correction to 90 percent mass participation for long term calculations.

Additionally, as stated in CP&L's letter to the NRC dated May 22,1985 (copy provided as ), a value of 33 Hz in conjunction with Code Case N 411 damping valve curves will be used in lieu of 20 Hz. The 33 Hz frequency cutoff was included in Amendment 4 to the Brunswick Plant Updated Final Safety Analysis Report, page 3.9.2 6.

. of the NRC's August 25,1992 letter provides the following NRC staff questions concerning CP&L's July 16,1992 submittal.

NRC OUESTION Q:

In your response to NRC Ouestion 1.H, you stated that "All masonry walls that were considered seismic in the originalIE Bulletin 8011 walkdown with structural angle restraints attached with expansion anchors that are located outside (underline added for emphasis) of the diesel generator building have been reviewed. The results of these inspections are summarized in the response of question I.B.* However, your response to NRC Question 1.0 stated that you have insoected all the walls in the diesel generator building, the walls with IE Bulletin 80-11 modifications in both tne conc building and diesel generator buuding, and a group of seismic walls outside the diesel gene stor building (i.e., the control building and reactor building). The answer to question I.B is not clear as to whether al] the walls with structural angle restraints attached with expansion anchors, classified as Categvy I or seismic, in the plant have been field inspected, design reviewed, and physically modified, if needed, to design standard. Please clarify your answer.

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i CP&L RESPONSE:

All Seismic Cat 6 gory I walls have been inspected for structuralintegrity and design reviewed for potentialloads in addition to seismic. Repairs are being made as necessary.

NRC QUESTION 7:

i in your response to NRC qaestion I.B, you stated that 6 walls, which were restrained by anchor bolted angles by original design and construction outside the diesel generator building, were 100 percent inspected and in a few cases anchors were discovered to be 5/8 inch sleeve anchors in lieu of 3/4 inch anchors, but all met IE Bulletin 8011. Since IE Bulletin 8011 does not address or provide any requirements with respect to anchors, we are confused by your response of meeting IE Bulletin 8011 as stated above. Expand or clarify your response.

CP&L RESPONSE:

Industry resolution for IE Bulletin 8011 contains criteria for acceptance of masonry walls. The Company's response actually refers to the resolution documentation for IE Bulletin 8011 which I

was provided by NRC letter dated January 30,1985,

  • Masonry Wall Design, IE Bulletin 8011' (copy provided as Attachment 5). This NRC letter issued the staff's Safety Evaluation and Technical Evaluation Report (TER). The statement referring to the six (6) walls is meant to indicate that the condition o' the walls is acceptable under the referenced TER, although they were not constructed per the original design.

i NRC O_UESTION 8:

In your response to NRC question I.E, you stated that those anchors with frozen studs were load tested. Explain how the load test was performed and the level of load which was used, such as allowable design loads.

l CP&L RESPONSE:

A manually operated calibrated torque wrench was used. Sufficient space was verified to exist between the plate and the anchor shell or shims were added during the test. The torque values specified in the procedure were greater than two times the manufacturers design load.

NRC QUESTION 9:

In your response to NRC question ll B, you stated that "An overall review of the IE Bulletin 80-11 program is underway for the Brunswick plant. The review will address existing masonry wall functions, including missile barrier, tornado barrier, ventilation barrier, or other functions for which it is not analyzed.' We are confused by the words 'for which it is not analyzed." Expand or clarify your response.

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i CP&L RESPONSE:

The Company's review for masonry walls under IE Bulletin 8011 included walls with functional requirements as indicated in item 1 of IE Bulletin 801' All missile barrier walls were reviewed under IE Bulletin 8011 for the appropriate missile losi -

The Brunswick Plarit has no exterior masonry walls serving as tornado barriers; therefore, tornado loading was not previously considered to control the design. A more detailed building analysis for tornado pressurization indicates that instantaneous pressures on internal walls in the diesel building control the cesign rather than seismic considerations controlling the design in some cases. The Company's functional review also identified a new sefety related function not considered in IE Bulletin 8011. The new safety related function is a series of walls in the diesel building that serves as a barrier between the supply and exhaust air for the diesel bays. These walls are being seismically restrained commensurate with their safety related function.

NRC QUESTION 10:

On page 4 of the enclosure with file No. B0913510c and Serial No. BSEP/821616, you stated that " Tests for proper installation were performed on 59 percent of the anchor bolts.............. All of the anchors with unremovable bolts or studs were successfully tested for pro Oad, however, demonstrating the load capability of the anchors." What criterion was used for ' proper installation?" Does 59 percent of the anchor bolts means 59 percent of the anchor bolts in the whole plant? What criterion was used for "preload?" Expand or clarify the words "however, demonstrating the load capability of the anchors."

CP&L RESPONSE:

This NRC questions contains a number of parts, each of which are addressed below:

Proper Installation

" Proper installation" was based on the following two tests:

1.

Thread engagement between the bolt and the sleeve. The test acceptance criteria is shown in the enclosed Table * (copy provided as Attachment 6) from the test procedure.

2.

Embedment depth, which for a self drilling anchor, was a function of plug depth (see the sketch provided as Attachment 7). The acceptance criteria is shown in the enclosed Table 3 (copy provided as Attachment 6) from the test procedure.

"59 Percent" "59 percent" means 59 percent of the total number of available anchor bolts in this phase of IE Bulletin 79 02 testing for the Brunswick Plant, Unit 2.

The following historical information may aid in putting this statement in perspective:

The Brunswick Plant IE Bulletin 79 02 testing effort was completed somewhat independently for the two units due to different outage schedules. Also, some piping.

systems or portions of systems were missed by our architect engineer when the scope of El 4

l IE Bulletint 79 02,07, and 14 was originally determined. Our July 26,1982 letter (copy provided as Attachment 8) was reporting completion of the anchor bolt testing for the Unit 2 portion of this additional scope. Hence,59 percent means 59 percent of the Unit 2 4

anchor bolts on the missed systems or portions of systems.

j Sut,sequently, in April 1992, when the masonry wall anchor bolt issue arose, CP&L i

management directed the IE Bulletin 79 02 effort be re examined to ensure the problems l

with diesel generator building walls did not extend to pipe support anchor bolts. The July 26,1982 letter needed clarification due to the number of " frozen

  • studs and nuts reported. Therefore, the /spril 1992 CP&L audit team retrieved and reviewed the individual anchor test data sheets to determine the reason for the frozen studs and nuts. The results of this audit have been discussed with the NRC in our letters dated At,ril 15,1992 (Serial No. NLS 92118); Mey 29,1992 (Serial No. NLS 92148) i i

it &ppears the reason for the relatively large number of froren studs and nuts found during l

this phase of the originalIE Bulletin 79 02 inspections is due to the fact that this is when the corrosion prone areas were covered. Hence, corrosion, not fraudulent installation, accounted for the frozen studs anu nute.

Deslan Load Test The "preload" or design load test involved the application, by hydraulic Jack or nut torque, of a pull out load equal to or greater than the allowable design load of the anchor. The acceptance criteria was that the anchor " pull out" during the loading be no more than 1/16 of an inch.

Load Caoability

" Demonstrating the load capability of the anchors" means the actual pull testing of the anchors, as described in the " Design Load Test" discussion abcve. This test demonstrates the anchor's capability to carry its design load.

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i ATTACHMENT 1 i

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CP&L Letter Dated May 29,1979 1

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ATTACHMENT 2 CP&L Let:er Dated May 22,1985 4

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Carolina Power & Light Company File NG-3514(B)

May 29, 1979 SERIAL:

GD-79-1401 Office of Nuclear Reactor Reguistion ATTENIION: Mr. I. A. Ippolitu, Chief Operating Reactors Branch No. 3 United States Nuclear Regulatory Commission Washington, D. C.

20555 BRUNSW'J CK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 DOCKET NOS.

50-325 and 50-324 LICENSE NOS. DPR-71 AND DPR-62 SEISMIC ANALYSIS OP SAFETY-RELATED PIPING Dear Mr. Ippolito At our meeting on May 21, 1979, Carolina Power and Light Company co:nmitted to provide the NRC Staff additional information concerning our response to IE Bulletin 79-07 on seismic pipe stress analysis.

On May 23 and 24, 1979, the Staf f identified to us, by telephone, and to a representative of United Engineers and Constructors, our architect engineer for the Brunswick Steam Electric Plant, several additional items that should be addressed in.our re sponse.

The remainder of this letter and attachments respond to those r eq ue s t s.

1 The analysis of the loads for the pipe supports for the first ten (10) lines reanalyzed for pipe stresses has shown that there were ten cases where the load exceeded allowable.

Table 1-1 summarizes the data on the 98 pipe supports on these ten (10) lines.

Table 1-2 presents the details of the ten (10) supports that were overstressed.

l While evaluating these ten pipe supports, it was determined that

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the supports had been underdesigned initially.

In no case was the overstressed condition a result of the new load from the i

seismic reanalysis.

As shown on Table 1-2, the new load actually decreased in five cases, increased less than 2.5% in l

f our cases, and increased 19% in only one case (which was l

already over capacity by 14.5%).

These ten supports were analyzed to determine if their structural integrity would be l

maintained under the identified loads.

Four of these supports were found to maintain stresses less than yield and thus would maintain stru.:tural integrity.

When it was determined that structural integrity would be l

compromised for the six supports under the calculated loads, Carolina Power & Light Company decided to shut down both units and make necessary modifications to these supports to reduce stresses to less than allowable.

These modifications have been initiated and th: new capacity is shown on Table 1-2.

.4 411 Fevetteoe St'est. # 0 Bor 1551. Aate gr N C. 27602 bf8-SY )

FAle NG-3514(B) SERLAL:

CD-79-1401 During this evaluation, it was noted that the overloaded pipe supports f ailed in two ways either concrete anchors or in torsion.

An investigation was begun to look at all pipe supports on safety related systems to determine if similar overloaded conditions may exist under the original load.

The resu1*s of this investigation will be available on June 1, and all necessary modifications vill be made prior to returning the units to operation.

2.

On iby 24, 1979, the Staf f informed us by telephone that the seismic stress analysis should be based on absolu;e sum if a two-dimensional seismic analysis was used, and that the square root of the sum of the squares (SRSS) was acceptable if a three-dimensional seirmic analysis was made.

The Staff further stated that a stress from a two-dimensional analysis calculated using SRSS and multiplied by a factor of 1.38 would be acceptable.

At the time that BSEP was licensed, two-dimensional SRSS seismic analysis was acceptable criteria, and it is n st apparent to us that the back-fit of a two-dimensional absolute sum seismic analysis has undergone the necessary requiremonet of 10CFR50.109.

Although CP&L does not accept the Staff's position, we have prepared a revision to Table 2 of our letter of May 21 demonstrating the effect of multiplying the two-dimensic 11 analysis results by the 1.38 f actor.

We have also taken credit for conservatism that exists in the relationship between the OBE and the DBE.

The results of this exercise show that only one line of the first thirty-nine e

reanalyzed lines exceeds total allowable stress by 22.

For this line, the total stress is still less than 0.9S.

y For the unreanalyzed lines shown in Attachment 3 of our May 21 letter (GD-79-1342), we have used the 1.38 f actor to establish criteria for priority of lines to be reanalyzed.

We do not plan to base our conclusions of acceptability on the use of the 1.38 factor, since it is not the appropriate criteria for BSEP.

In determining the criteria for priority of reanclysis of the remaining linds, SRSS stresses were estimated on the basis of a f actor of 1.5 increase, and this resultant was then multiplied by 1.38.

Credit for the conservatism of the OBE/DBE relationship was taken into account prior to applying the 1.5 increase.

When this was applied to the 411 lines that have not been reanalyzed, 39 of the 411 exceeded allowable stress, and are tabulated in Attachment 2 to this letter.

Our reanalysis priorities have been changed to include these 39 lines in those to be reanalyzed the week of May 28, and the results of this reanalysis should be available on June 1,1979.

We still j

anticipate completing the total reanalysis in accordance with I

our previously stated completion date of July 21, 1979.

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Mr. T. A. Ippolito May 29, 1979 i

3.

As a result of an I & E inspection at the Brunswick Steam

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Electric Plant to verify that the as-built dimensions were the i

same as the es-designed (as-analyzed) system, four deviations Were noted.

These are discussed in Attachment 3.

1 As stated in the meeting on May 21, 1979, and confirmed in our j

letter of May 22, 1979, Carolina Power & Light Company will i

verify as-built dimensions for all safety related systems at i

BSEP.

This verification is currently in progress for those lines outside containment.

The lines inside containment will be verified at the next scheduled outage.

Due to the time constraints on reanalysis, the reanalysis is being conducted l

concurrently with the as-built verification.

If any discrepancies are identified between the as-built /as-analyzed configurations, an evaluation by a stress analyst will be made to determine if the line should be reenalyzed. This evaluation will be based on evaluating the magnitude of the computed stresses for the area in question, and the impact (increase or decrease) on the stresses expected for such deviation.

If it is determined that the line needs to be reanalyzed to determine the new stress level, we will promptly reanalyze the line.

4.

During our recent meetings, the relationship of 7E Bulletins 79-02 and 79-07 has been discussed.

Some of the pipe supports analyzed in the first ten lines are anchored using concrete expansion aachors discussed in Bulletin 79-02.

In the 79-07 support reanalysis, these base plates were and will continue to be analyzed using IE Bulletin 79-02 as a guide.

The capacity established for the concrete anchors is 20% of the matufacturer's rated capacity.

Using this criteria, two supports on the first ten lines had to be redesigned and now have sufficient capacity.

As stated in item 1 above, the i

remaining supports using concrete expansion anchors are being investigated-to determaae their adequacy and will be reported on June 1.

A final report on all of our analyses and testing related to the concrete expansion anchors and IE Bulletin 79-02 will be submitted in compliance with the bulletin schedule.

5.

The Staf f requested information e n the location of the postulated pipe rupture for a LOCA relative to the point of highest stress.

The BSEP piping design did not use the mechanistic approach of locating the pipe break at the point of highest stress.

The postulated break for doubled-ended guillotine or longitudinal split was analyzed for the pipe break to occur at any point on the pipe, inside or outside containment.

6.

We have been informed that during a meeting between NRC, another licensee and United-Engineers & Construttore (UE&C),

some questions were raised by the NRC staff about the subject m

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!!r. T. A. Ippolito May 29, 1979 of valve operability.

In the event the staff may have any questions concerning this topic as it may apply to BSEP, we will be prepared to address this issue.

7.

Carolina Power & Light Company's criteria for determining if an overstressed condition is reportable is set forth belows a.

Lines Yet Tn Be Reanalyzed The stress using new seismic data and revised analytical criteria are estimated for the lines that are yet to be reanalyzed.

As stated previously, those with high estimated stresses are being analyzed first in the reanalysis program.

We will not use estimated stress as a basis for determining overstressed conditions which are r epo rtable.

b.

Reanalyzed Lines Those lines which have been reanalyzed and which show an apparent overstress condition will be evaluated in detail to determine if it is a reportablu item.

First, the known conservatisms will be removed f rom the analysis.

The line will be analyzed to determine if the stress at any single modal point exceeds FSAR criteria of 0.9Sy or 1.8 S e h

whichever is the higher.

If the pipe remains overstressed, this will then be considered a reporthble item and the NRC will be informed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

c.

Reanalyzed Pipe Supports When the reanalyzed pipe data is available, the pipe supports will be reanalyzed for the revised load.

If the load exceeds the apparent support capacity, the specific support will be analyzed in detail to determine if the stated capacity is the actual capacity without exceeding 0.9 S.

If the load still exceeds the capacity, a y

determination is made if the support will maintain structural integrity even if the allowable is exceeded.

If structural integrity is maintained, this is not considered reportable.

If structural integrity is not maintained, the support is taken out of the computer piping configuration, and the line is reanalyzed.

The results of this reanalysis' are evaluated to determine if other supports and the pipe can take the additional load without exceeding their structural integrity.

If the system maintains integrity, the item is not reportable.

If the system does not maintain structural integrity, the item will be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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In summary, CP&L has evaluated the data from the lines reanalyzed to date, and the estimates for revised stresses for lines yet to be reanalyzed, and it is our conclusion that the continued operation of the Brunswick Steam Electric Plant, Units 1 L 2 is warranted without undue risk to the public health and safety, while the reanalyses of seismic design continues.

The problem associated with those supports that were found to be overstressed is a result of initial underdesign of those supports, and is not related to the use of algebraic, square root sum of the square, or absolute summation of seismic stresses.

The modifications cf those supports which were originally under-designed vill be completed in early June, and at that time, both units will be returned to power.

As stated in our letter of Kay 15,1979, and in item 7 of this letter, 24-hour reporting criteria have been established if any piping or supports are determined to be overstressed during the reanalyces.

If you have any questions concerning this information, please do not hesitate to contact our staff.

Yours very truly, k1g/

,. E. Utley Executive Vice President Power Supply DLB/sg bec:

Messrs.

D. L. Bensinger C. S. Bohanan D. B. Waters / File NG/3514(B)

J. M. Johnson W. B. Kincaid S. McManus A. C. Tollison, Jr.

C. W. Woods (LIS)

File BC/A-4 File B-X-0274 r

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00-79-1401 ATTACHMENT 1 PIPE SUPP' AT ANALYSJS, An evaluation was performed on the pipe supports of the first ten lines that were reanalyzed in the seismic pipe stress reanalysis program.

There are 98 pipe supports made up of snubbers, vendor catalog pipe supports, and fabricated supports.

The recalculated loads compared to the original load and support structural capacity are tabulated on Table 1-1.

As can be seen on Table 1-1, the load did not increase appreciably due to the seismic stress reanalyses and recalculation of loads.

The load decreased for 30% of the supports and increased less than 25% for 60% of the supports.

The load increased greater than 25%

for only nine supports, but the new loade were less than 75% of capacity for these supports.

However, ten supports were found where the load exceeded the applicable allowable.

Further investigation revealed that these ten supports were underdesigned initially.

For these ten supports, the new loads were less than the old loads in five cases, increased less than 2.5%

in four cases, and in only one case, the increase was 19%.

These ten supports were analyzed in detail to determine if they would maintain their structural integrity under :he specified loads even if they exceeded allowable.

This is summarized on Table 1-2.

In four cases, including the one where the new load was 19% higher than the old load, the supports retained their structural integrity.

Six supports would fail.

The six supports that would fail under the specified load (old or new) were redesigned to have their stresses less than allowable.

The new design loads for these pipe supports are shown on Table 1-2.

It has been concluded from the analysis of 98 pipe supports that the seismic stress reanalysis does not contribute to overloided pipe supports.

However, it lias been recognized that there is a potential for certain supports to be overloaded due to an error in the initial design.

These errors have been found to be with concrete expansion anchors and with torsion of the beam support. An investigation has begun to examine the pipe supports of the other safeth-related piping for similar problems. The results will be reported at a later date.

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TABLE 1-1 4

SUMMARY

OF PII'E SUPPORT IIIADS UPSET LOAD CAPACITY OLD 1 DAD NEW IDAD RATIO' RATIO LINE POINT LIMITING PART C

(OL)

(L)

L/0L L/C REMARKS 17 65S Strut.

13800 7037 7326 1.04

.53 Cat. P.S.

95S Cat. F.S.

15700 9129 9311 1,.01

.59 1105 Snubber 3920 1134 1430 1.26

.36 137S Cat. P.9 4960 4192 4286 1.02

.86 172S S.S. Sup't.

1836 606 595 0.98

.32 1

1 1755 Cat. P.S.

11630 6329 6527 1.03

.56 220S X Cat. P.S. Etrut 3920 968 1176 1.21

.30 Y Cat. Sup't.

6230 5544 5588 1.007

.90 Z Cat. P.S.

3920 1483 1634 1.10

.42 2555 Cat. P.S.

8000 4088 4160 1.01

.52 402S Snubber 3920 411 576 1.40

.15 24 136 XZ Snubber 3920 1339 2690 2.00

.68 Y Snubber 3920 1352 1400 1.03

.36 154 I Snubber 3920

+1576

+1243 0.09

.36 Z Snubber 3920

+1576

+1243 0.09

.36 ISB 61S Z Snubber 3920 956 1078#

1.12

.27 61S Y Snubber 3920 1381 1912 1.38

.49 105S Snubber 3920 1544 2133 1.07

.54 18S Wald 75300 11850 11466 0.96

.75 60S Snubber 22177 11837 11578 0.97

.52 71S Snubber 13666 7993 8392 1.04

.61 107S Snubber 13800 3187 3196 1.002

. 23 73S Snubber 37600 25612 23214 0.90

.62 110S X Snubber 29090 5239 4700 0.89

.16 110S 2 Snubber 20736 14316 13468 0.94

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TABLE 1-1 (Cont'd)

UPSET IDAD-CAPACITY OLD IDAD NEW IDAD RATIO RATIO LINE POItTr LIMITING PART C

(OL)

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237 420 conc. Anchors 678 43t9 4014 0.91 (5.92) See Table 1-2 i

i 420 Y-RSSA-20 Strut 20000 11124 13311-0.11

.67 440 X-EX.W8x17 790.

2784 2490 0.89 (3.15)

(Torsion) 4 4

440 Z-Strut RSSA-10 10000 5641 5792 1.02

.58 4

466 Snubber 3920 1884 1945 1.03

.50 472 Wald-Post Ex.St1 5374 4001 4551

1. 13

.85 484 Snubber 13800 3546 3570 1.006

.26 503 Snubber 13800 3339 3539 1.05

.27 525 Weld-Post to Snub 6000

.4613 4884-1.05

.81 4

122 2092 Snubber 3920 2117

-2101 0.99

.54 q

i 2094 Snubber 23600 3818 3827 1.002.

.16 2220 Snubber 13800 9008 9048 1.004

.66 2143 Snubber 13800

~5674 5670 0.99

.41 t

2156 X -W6 x 15.5 660 2726 1775 0.65 (2.69)- See Table 1-2 (Torsion) 2156-Z-Snubber 23600 1521 1504 0.98

.06-2230 Snubber 13800 4804 4182 0.87

.30 2240 Snubber 13800 6179 3685 0.59

.27 2174 W6 X 15.5(Torsion) 660-6714 6874 1.02- (10.42) See Table 1-2 4

'2062 Snubber 13800 3044

-3046 1.00-

.22 t

i 121

-3084 Conc. Anchors 840-2970 2991 1.007 (3.56) See Table 1-2 3083 Conc. Anchors 12960 3866 3207 0.82

. 15 3067 Snubber 13800 6190 6244

- 1.008

.46 3066 Snubber 13800 8152 8010 0.98

.59-1 11500 10798.

9143 0.84

.79 l-

-305S Clamp

-(55) 4

-..-,+.-,n_-,

+an.,

-n+-e--.,m,

TABLE 1-1 (Cont'd)

UPSET IDAD CAPACITY OLD IDAD NEW IDAD

-RATIO RATIO LINE POINT LIMITING PART C

(OL)

(L)

L/0L L/C RDERKS

- 3048 Snubber 3920 2843 3214 1.13

.83 L/C =.95 for Emergency Conditic Snubber 13800 3302 3405 1.03

.24 3

l 3200 Snubber 3920 712 1026 1.44

.26 6

13 S Snubber 37600 31000 18000 0.58

.48 4

.13 3S Snubber 23600 14499' 13400 0.92

.57

'32/

Fab. Sup't 7460 3341 5463 1.63

.73

3 f

4 Snubber 13800 10189 9377 0.93

.68

+ JS Snubber 13800 4628 5752 1.24

.42 806S Saubber 13800 6099 6863 1.12

.50-103S Saubber 13800 5106 8402 1.64

.61 719S Struct. Supt.

8800 11005 8600

'O.78

.98 718S-Snubber 13800 7383 5753 0.77

.42 24S Snubber 13800 7651 6137-0.80

.44 l

7 25S Snubber 3920 1200 1249 1.04

.32 710S Snubber 3920 2114 2423 1.14

.62 i

724S Y Struct. Supt.

1000 3406 4112 1.20

.41 Z Snubber 3920 1601 1823 1.13

.47-l l

901S Snubbar 3920 2154 2566 1.19

.63 900S Saubber 3920 1818 2160 1.18

.55 722S-Struct., Supt.

12200 2826 3442 1.21

.28

~

108S Cat., Pipe-Supt.-

0 7144' 7693 1.07

.86 510 133 S truct. Steel 3563 1948 1961 1.006

.55 Supt.

125 Wald 3712 3361 3505 1.04

.94 13 2 Strue. Steel 1350 1546 1844 1.19 (1.36) See Table 1-2 Supt. Channel

-4 TABLE 1-1 (Cont'd)

U* SET IDAD CAPACITY OLD IDAD NEW IDAD RATIO RATIO j

LINE POINT LIMITING PART C

(OL)

(L)

L/0L L/C REMARKS 195 Strue. Steel 27600 7049 7126 1.01

.26 Supt.

148 Strue. Steel 6040 23 16 2276-0.98

.38 Supt.

120 Strue. Steel 9645 5510 5498 0.99

.57 Supt.

111 Bolts 1445 77 99 1.28

.02 i

137 Strue. Steel 3897 2265 2264 0.99

.59 Supt.

16 Bolts 5038 5601 5456 0.97 (1.08) See Table 102 230 Strue. Steel 3770 3415 3473 1.01

.93 I

Supt.

224 Strue. Steel 12992 11140 11311 1.01

.87 Supt.

i.

225 Strue.

4960 3207 3211 1.001

.67 Supt.

40 Strue. Steel 5250 5271 5255 0.99 1.0 l

Supt.

116 Strue. Steel 3920 522 524 1.003

. 13 Supt.

i 113 Strue. Steel 4960 3743 3744 1.00

.76 Supt.

e 30 Snubber 13000 8758 8881 1.01

.68 l

l 270 Struc. Steel 11800.

18800 18900 1.005 (1.6) 136 Strue. Steel 3170-6556 6570 1.002 (2.07)

135,

.(Pipe Saction) i-803S-l l

125

- 10725 Strue. Steel 4192 4122

'4093' O.99

.98 1065S Conc. Anchors 12960 6302 7882 1.25

.61 1051S Snubber 11500 3477 3476 0.99

.30 1120S Snubber 13800 1452

-1437 0.98

. 10 1140S -Clamp 11500 10085 10488 1.03

.91

i 5..

TABIE 1-1 (Cont'd)

UPSET If>AD i

CAPACITY OLD IDAD NEW IDAD RATIO FATIO LINE POINT LIMITIE PART C

(OL)

(L)

L/0L L/C RDRRKS 1150S X Beam (Torsion) 790 1470 1301 0.88 (1.65)

Z Snubber 3920 807 998 1.23

.25 1029S Snubber 13800 9111 9630 1.05

.70 i

I t

l e

I I

TADIE 1-2

SUMMARY

OF PIPESVIT(H L/C #1.0 t

OLD NEW CAPACITY CAPACITY TO RATIO IDAD IDAD RATIO REDESIGN LINE POItfr LIMITING PART (C)

YIELD (Cy)

(NL/C or Cy)

(OL)

(NL)

OL/NL IDAD

  • e,

237 420 Cone. An d ors 678

> 1. 0 4399 4014 1.09 Y 13430 X 4192

.440 8W x 17 I-Bm 790 21.0 2784 2490 1.11 6762 Torsion I??

2156 6W x 15.5 660 p1. 0 2726 1775 1.53 2769 s

I-Bm Torsion 2174 6W x 15.5 660 21.0 6714 6874 0.97 11342 I-Bm Torsion j

121 3084 Conc. Anchors 840 71.0 2970 2991.

0.99 4592

[

l i

510.

132 SS (Channel) 1350 2194

.84 1546 1844 0.83 16 Bolts 5038 8184

.67 5601 5486 1.02 i

270 Stru. Steel 11800 19656

.96' 18803 18900: 0.99 136, 135 Stru. Steel 3170 6510

~ 1. 0 6556 6570 0.99

  • w 803S Pipe Section i.

125 1150S I-Bm Torsion 790

>1.0 1470 1301 1.12 2243 i

  • Cy does not app}y
    • Not redesigned for short term fix
      • Redesign Imad consists of new calculated emergency load x 1.38 + transient 1

to be less than AISC allowables (0.67 Sy).

ATTACHMENT 2

--u SEISMIC PIPE STRESS ANALYSIS CRITERIA As stated previously, the criginal seismic analysis for pipe stress used algebraic summation within each mode. A reanalysis effort was undertaken for all safety-related lines using the UELC - ADIFIPE-2 Computer Code which employs the equare root - su=-of-the-squares (SRSS) load combination within each mode.

The results of the reanalyses, given to the NRC Staff in our re-sponses to IE Bulletin 79-07, in letters A ted April 24, May 15, and May 21, 1979, used the SRSS method.

On P 24, 1979, the NRC Staff notified CP&L that the use of SRSS with i hree-dimensional seismic analysis was acceptable, but for a two-d.ensional seismic analysis the absolute sum method should be employed within each mode. The anclysis f or Brunswick uses a two-dimensional seismic input approach.

At the time BSEP was licensed, the two-dimensional SRSS analysis was the acceptable criteria.

Therefore, the acceptability of stress levels should not be based on absolute sum.

However, to use the most conservative case for comparison purposes only, the stresses calculated using UE6C - ADLPIPE-2 were multiplied by 1.38 (a number acceptable to the NRC Staff) to obtain stresses for the Operating Basis Earthquake (OBE).

As discussed in Attachment 7 of our letter to the NRC GD-79-1342, dated May 21, 1979, the previous seismic analysis used a most conversativa approach of relating stresses for an OBE to that for a Design Basis Earthquake (DBE), known today as a Safe Shutdown Earthquake (SSE).

The stresses computed in the OBE were multiplied by 2 and used as the stresses for a DBE. As discussed on May 21, 1979 with the NRC Staff, our reevaluation of the OBE and DBE Amplified Response Spectra (ARS) l indicates that the relationship between the two ARS in the fraquency range that affects pipe stress is less than 1.2, and frequently less

,y than 1.0.

However, a value of 1.2 has been selected for use to convert-,

l OBE stresses to DBE stresses.

l For the thirty-nine lines already reanalyz ed, the conversative stresses for comparison purposes for the DBE and total are shown on y ble 2-1.

The DBE stresses in this trble are calculated as follcws: V DBE =

l CI~ OBE x 1.38 x 1.2, where CI~0BE is obtained using the UE&C - ADLPIPE-2.

j For the lines yet to be reanalyzed, the stresu for a DBE was estimated l

as explained in Attachment 7 to our May 21, 1979 letter using a factor of 1.5 to account for the highest expected increase in stress due to the reanalysis for SRSS (within each mode) in lieu of algebraic sum l

(within each mode) and which is based on the data from the reanalyzed lines.

For those lines identified on Attachment 3 to oue May 21, 1979 letter, the stress for a DBE were estimated as follows:

l

([ DBE (I OBE x 1.38 x 1.2 x 1.5

=

Est.

Orig, where CI 0BE was computed in the original analysis. Those lines whose Orig.

estimated stresses exceeded allowable are tabulated on Table 2-2.

i

2-EVALUATION As can be seen f rom' Table 2-1, one line (RHR-60, Residual Heat Removal) exceeds the allowable (1.8 S ) by 1.7 percent.

However, this stress-is h

less than the ' stress equal t5 0.9 Sy (32,400).

The BSEP FSAR allows the use of 0.9 Sy or 1.8 S, which.,ver is greater, as the allowable -

stressduringemergencycoNdition(DBE).

Therefore, the stresses are acceptable for all lines reanalyzed.

Table 2-2 shows that 39 of 411 lines yet to be reanalyzed exceed allow-able (1.8 S ).

These stress values are not necessarily based on coinci-h dent point maximums, but rather the summation of maximum stresses for each individual loading.

It should be restated that these stresses are estimated and that they were derived using a e,onservative factor-of 1.5 to cover the maximum increase expected for the reanlaysis (old i

algebraic to new SRSS combination within each mode). - As discussed in Attachment 7 and shown on Attachment 8 of our May 21. 1979 letter, in over 58% of the lines already reanalyzed, the new seismic stress J

was.less than the original seismic stress.

-In over 87% of the cases,

]

the new stresses were less than 1.25 of the orig 1nal stresses.

As discussed previously in our letter, Carolina Power & Light Company-commits to placing these lines in the highest reanalysis priority category, regardless of the priority category previously established on a function and size basis.

~

It should also be pointed out that of the 39 lines estimated to be overstressed, 27 are 2" or less in diameter.

i i

W i

i-i L

l f

w-w v

i e

t 3+*-%

v y--

79 r--

q--w-'*-Ty-g y-g 9vgg 9-e "r**

g-'

dFNq y-*wg 6-O'EPT*

w 9-*HWp

{74t

ATIACHMENT 2 PIPE STRESS REEVAIDATION it SUPHARY

^

q EMERGENCY CONDITION (PSI) a TOTAL

l -

LINE SIZE ORIGINAL ORIGINAL TOTAL SE1SMIC TOTAL' SEISMIC STRESS SYSTEM NAME ISO NO.

(NPS)

TOTAL SEISMIC 5/21/79 5/21/79 5/25/79 5/25/79 ALIIMABM ALIIM&BIE Main Steam MS-ISB 24 10724 3942 10640 3858 9976 3194 27000 37

!; S:fsty/ Relief SRVL-121 10, 6 23012 12280 21910 11180 19987 9257 27000 74 1,giv, Safety / Relief

'SRVL-122 10, 6 19685 15800 24439 13352 22143 11056 27000 82 Valys Snfaty/ Relief SRVL-237 10, 6 20432 12004 24588 16160 21809 13381 27000 81 Valva Sefsty/ Relief SRVL-125 10, 6 24270 13347 24316 20270 20830 16784 27000 77 Valn Feedwiter W-16 18, 12 18007 12420 20028 13296 17741 11009 27000 66

. Residual Heat RHR-6 20 19406 13582 12644 6820 11471 5647 27000 42 Removwl Core Spray CS-24 10 16952 10076 14366 7490 13078 6202 27000' 48 High Press Cool HPCIS-17 14 12200 6446 12502 6748 11341 5587 27000 42-Intet High Press Cool HPCIS-510 14, 12, 10 12004 7994 12092 8082 10702 6692 27000 40 Inict High. Press Cool HPCIS-10 14, 12, 10

. Inict' 9733 3886 11152 5530 10201 4579 27000 38 Residua 1 Heat RHR-l' 24, 20 24094 18584 17972 14366 15501 11895 27000-57 Removal Rezidual Heat RHR-2 20, 16, 12 13309; 7654 11471 5948 10448 4925 27000 39

~

I Removal

'I

.Raidual Heat RHR-5 24 9848 3896 9514 2960=

9005 2451 27000 33 Remotml ii Re idual Heat RHR-25 4, 6 18558 12904 18530 12876 16315 10661 27000 60 Removal l

Nuc1cer Steam NSS-14 24, 10 14745 8446 16335 10036 14609 8310 27000 54 Supply S:fsty/ Relief.

SRVL-124 6, 10 25536 15928 25984 16376 23167 13559 27000 lj Volva J

'i v

ATTAClutEITT 2 (COfff'D)

DfERGENCY CONDITION (PSI) j TOTAL LINE SIZE ORICINAL ORIGINAL TOTAL SEISMIC ' TOTAL SEISMIC STRESS SYSTDI HAME ISO NO.

(NPS)

TOTAL SEISMIC 5/21/79 5/21/79 5/25/79 5/25/79 ALIJNABLE ALI&dABLE S f;ty/ Relief SRVL-126 6, 10 23361 18000 22197 17422 19200 14425 27000 71

Velv, Residual IIcat RHR-52 14, 12 23271 17936 19539 14204 17096 11761 27000 63 Remov'1 Re:cter Core RCIC-21 3

7603 3588 7601 3586 6982 2967 27000 26 I mlet. Cool Resid Heat Rem RHR-173-B 1

3808 2186 3838 2216 3457 1835 27000 13 Dr^in Line Residual Heat RHR-28 20, 16, 12 15298 8814 13626 7142 12398 5914 27000 46 Removn1 Nuclu r Steam NSS-15' 24 10974 4420 10458 3904 9787 3233 27000 L6 Syatem Nuclear Steam NSS-120 10, 6 19443 8902 17899 8298 16472 6871 27000 61 Syrtem (ISC)

High Press Cool HPCIS-4 3, 6, 10, 12 23609 20876 25481 22748 21568 18835 27000 80 Inlet Nucic.cr Steam NSS-123 6, 10 21027 16098 18577 16782 15691 13896 27000 58 Syrtem (15C)

Nuc1:ar Steam NSS-187 10, 6 21856 11337 23596 16424 20771 13599 27000 77 Syrtem (ISC)

Residual Ifeat RllR-42 12, 14 18116 12976 17480 12340 15358 10218 27000 57 Remov'1 Re;idu21 Ileat RHR-3 14, 16, 20, 25317 18328 23379 15590 20698 12909 27000 77 Removal 24 R.ecidual lleat RilR-13 4, 8, 14 12532 10018 12620 10106 10882 8368 27000 40 Remov91 Residual lleat RHR-59 4, 6, 10 12664 9970 13344 10650 11512 8818 27000 43 Remov11 Residual lleat IUIR-60 4, 6, 10 34618 33658 32971 32012 27465 26506 27000 102 netnov,1 Re idual liest RHR-168 1

23580 22658 26910 26198 22404 21692 27000 83 Removm1 Re idual lleat RHR-61 4, 6, 3/4 21117 17038 19393 15314 16759 12680 27000 62

' iRemov'1 4

f

L

, ATTACHMENT 2 (CONT'D) t EMEHGENCY CONDITION (PSI)

TOTAL LINE SIZE ORIGINAL ORIGINAL TOTAL SEISMIC TUTAL SEISMIC STRESS SYSTEM NAME ISO NO.

(NPS)

' TorAL SEISMIC 5/21/79 5/21/79 5/25/79 5/25/79 ALIDWABLE AllDEBIZ f

High Presaure HPCI-11 16, 14, 6 21386 19790 17214 16618 14528 12932 27000 541 Cooltat In-liunction

. Reactor Core Injunction RCIC-196 1, 3/4 22706 19504 22504 19302 19184 15982 27000 71 coolina Rasidual Heat RHR-41 3, 4 26802 20728 26824 20750 23255 17181 27000 86 Removal Re2 dual Heat RHR-199

4. 1, 14, 23410' 20242 23398 20230 la918 16750 27000 74 Rearvel 3/4 Rearter Core RCIC-194 2, Ik, 1 24519 24118 24559 24156 20404-20001 27000 76 !

l Iml. Coolina 4

)

f J

t

)

il

!j Meismic stresses shown.are obtained by multiplying the OBE Seismic Stresses by 2.

    • Total stress (5/25/79) are based on:

(DBE x 1.2) 1.38 + (Total Stress - Seismic) 2 5/21/79 5/21/79 4

i I

f 3

ATTACIDiEfff 2 - TABE 2-2 IDCATION EMERpENCY CONDITION STRESS (PSI)

ISO /

INS. OR TOTAL SEISMIC TOTAL PROB.

SHEET OITISIDE LINE STRESS (DBE)

TOTAL SEISMIC ALIDWABLE STRESS [

M1 SYSTEM NO.

Coffr.

RT7R 5/71 /79 5/71/79 5175/79 5/75/79 (1 R Sij ALIDWABLE 2

Primary Steam Condensate 128 In 2

24225 20022 29070 24867 27000 108 Drain Inside Dry Well (East) and (West) 32 liigh Pressure Coolant 152 Out 1%

23242 22760 28750 28268 27000 106 Inj. (Main Pump to I

Barometer Cond.)

i

'34 High Pressure Coolant 154 out 3/4 23191 20780 28220 25809 27000 105 Inj. (Misc. Vents &

Drains Booster Pump) l.!

35 liigh Pressure Coolant 155 Ditt 2

23245 20990 28325 26070 27000 105 Inj. ' (Turbine Exh.)

Out 2

23346 21800 28622 27076 27000 106 Out h

23762 23476 29443 29157 27000 109 38 High Pressure Coolant 158 Out 1

27538 25060 33602 31124 27000 124 T

Inj. (Misc Vei.t, Ter.t Out 3/4 24715 22760 30223 28268 27000 112 q

& Drains Lines)

Out 3/4 25923 22010 31249 27336 27000 116

,j

'i 46 Core Spray System (C.S.

39 Out 3

29154 26788 35636 33270 27000 13 2 Min. Flow By-Pass Pump 2A)

Out 3

25456 21860 30746 27150 27000 114

]

48 Core Spray System 105 Out 2

25235 22438

'50655 27858 27000 114 1

(RHR Conn. from C.S. Pump 2A)

Core Spray System 105 Out 2

25235 22438 30655 27858 27000 114 (RilR Conn. from C.S. Pump 2B) 54 Service Water Sa!t Water 82 Oat 20 29536 25810 35782 32056 27000 133 Supply to RHR, Service j

Water Pumps (South)

t a

i f

ATTACHMENT ] - TAB E 2-2 (Cont'd)

~

i

(

IDCATION EMERGENCY CONDITION STRESS (PSI)

ISO /

INS. OR TOTAL SEISMIC TOTAL l

PROB.

SHEET OUTSIDE LINE STRESS

'(DBE)

TOTAL SEISMIC ALIDWABLE STRESS /-

ND..

SYSTEM NO.

CONr.

SIZE 5/21/79 5/21/79 5/25/79 5/25/79 (1.8 s )

ALIEWABM h

f 55

' Service Water. System

-106 Out 4

26D9

.25100 322D 31174-27000 119-6"' Return Line from Pump 6

Room Cooler 2A:

56 LService Water System 107 Out 6

24965 17948 29308' 22291 27000 109 3

6" Supply Header (South).

Out 4

29907 28212 36734 35039 27000 136 57 Service Water Syatem

'108 Out' 2

24300 2D52 29467 26519 27000 109-1 6 Supply Header (North) f.

C 70

. Reactor Water Clean-up 22 In-6 24303 17D4 28449 21280 25940 110 i

R.W.C.U.-Pump Suction i

f 80 Cont. Atmospheric Control 211 Out 2

26825 24396 32729

'30300 27000 121 j

Valt e' By-Pass Piping '

t I

l 81 Cont. Atmospheric control '212 Out-4 23948 21750 29212 27014 27000 108 Vent Purge Line From Drywell

.y 83 containment Venting 230 out 25268 14896 28873 18501 27000 107 89 Instrument Air System 179 Out 2

24598 23348 30248 28998 27000 112 f

Supply Line (North) West

{

ng,n v

93 Instrument Air System

'184 In 2

30632 29382-37742 36492 27000 140 Supply Header (North) s

-95 Instrument Air System 189.:

In-1%

25719 24512 31651 30444 27000 117

[

Pipe to Accum L!

i 96)

Instrument Air System 190 Out 3/4 30507 29382 37617-36492 27000 139 i

Supply Lines'to Filters i

D-0005 and D-0006 1

t

.-... ~ _... - _ _.... _ _ -. _ _.

t ATTACHMENT 2 - TABIE 2-2 (Cont'd)

~

IDCATION EMERGENCY CONDITION STRESS (PSI)

ISO /

INS. OR TOTAL SEISMIC TOTAL STRESS l

ALIDWABIE{ -

.PSDB.

SHEET OtFISIDE LINE STRESS (DBE)

TOTAL SEISMIC ALIDWABLE i.

NO, SYSTEM NO.'

CONI.

SIZE 5/21/79 5/21/79 5/25/79 5/25/79 (1.8 S )

h j

.97-Instrument Air System 181 Out 2

28127 26922 34642 33437 27000 128 Outlet from RCVR at "18R"'

Supply West Col. "T" 98 Instrument Air Sys. Inner 192 In 2

28344 27094 34900 33650 27000 129 Air Supply Header Outer f y Header Air.Sv l

I 201 In

.3/4 25822 18666 30339 23183 25920 117

-i

'Recirc Pump 2B j

99 Instrument Air System Li 206 In 3/4 9.4990 22410 30413 27833 27900 109 1101-Instrument' Piping, Piping at Temp. Equalizing'D005a l

l '

'102 Instrument Piping 207 In 3/4 21655 19622 26403 24370 26028 101

?

Lines 2E21-701 & 702 1

108 Nitrogen & Off Gas Services 309 Out 3/4 27170 24018 32982 29830 27000-122 Bldg. Instr. Air Interrupt-able

'i f

]

110 RHR 545 Cut 4

23765 18128 28151 22514 27000 104

}

f

'113 RHR Drain to 154 548 4

23806 21690 2S055 26939 27000 108

.}'

p f

116 RHR tumps 1A & IB 605 2

24752 20878 29804 25930 27000' 110 117 Service Water Sys.

606 6

26321 23760 32071

.29510 27000 119 125

'Instrumect Air.

690 3/4 26655 25529 32833 31707 27000 121 t

S 709 8'

15579 11038 29288 13709 -

27000 108.

129 Cont. -Atmos. Control Sys. Sup. Lines 710 1

132 Service Water

-716 1%

2314s 18610

.27646 23113 27C00 102 0

4

)

I n

V i

l

)

J-

=

^ -'

- - :. n., -

2 w+.- a c,,;.

1-. c.. a ATIACHMENT 3 AS-BUILT DRAWINGS i

i As a result of the NRC-I&E valk-through of approximately 67 pipe supports on saf ety related lines, four discrepancies _were identified:

1.

Isometric 17 High Pressure Coolant Injection main pump discharge line above el. 18'-9" dat.s point 45 does not agree with piping drawing. Actual location of support is 9'-2" from valve F006 in lieu of 7.0' as shown on the analysis isometric.

1 d

Comnent: he analyzed location-has been reviewed by stress

{

analysist and confirmed that the actual placement of the support will have little or no effect on the results of aaalysis for the i

3 following reasons:-

l 1.

me total =mrimum stress of the line is less than 50% of l

the code allowable stress.

see attachment 2 4

2.

- The placement of the support within approximately two pipe diameters of its analyzed position on this 14" Sch.- 120 j

pipe vill not advarsely affect the analysis.

l 1

2, Isometric 20 Core Spray data point 101 is-located approrimately.

~~.

l'-3" closer.to valve F015A than shown on the isometric.

Comment: Review by stress analysist confirms that since the data i

point is a snubber placing it closer _ to_ the valve is better than t

the original placement.

In addition t.he new placement will have no' adverse effect on the stress analysis since the new placement

{

is in the same plane as analysea.

1 -

c,,.,l, S*"

~~'*K"I-~

, ' ?-

?'~"*"

-+=v

+

-r~~'---~%=

=h"*"h~'"'"*~*~'"*"~

" " * ' * * * * * ' ' * " ' " * * * ~ ~ * ' * "

p.

,a y,

v e.

. ~. -.. _ _ - ~. -

e 6.,.._~

g

- j

.,.y.. _ - _ -, _ ~. ->

w.~.:.

-~-a: c - :. r- ~ ~-

~ ~ -

~ --x

.~

c r_.n me:x

.c 3.

Isometric 12 Reactor Core Isolation Cooling ptunp suction lines i

data point 272 vertical snubber is located on the opposite side.

l.

~

of an elbow than is shown on the analysis isometric.

1 i

Cournent:

Review by stress analysist confim that placement han no effect on the stress analysis.

The analysis progran treats l

the elbow as a point in the model, therefore transfer frem one side of an elbow to the other has no affect on the analysis re-1 i

suits as long as the snubber acts in the required direction.

Field check of the installation has veriff sd that the snubber is acting in the correct (vertical) direction.

4.

Isometric 18 Core Spray Pump suction line 25, data point 236, 4

i is eleven inches closer to ptrop than shown on-the analysis isometric.

Consnent:

Review by ctress analysist confirms that the location l

of the eupport within.one pipe diameter will not adversely effect 4

i the stress analysis. In addition the support is a-sliding dead-weight support which has-no'effect for seismic support.

It should be noted that in the above caset it has been determined by a stress i-i analysist that there is no adverse impact on the pipe stresses.

HJwever...

e l

Carolina Power and Light has couanitted to perform-an _as-built, verification on all. lines incidded in the reanalysis - to increase the confidence that-the as analyzed condition is consistent with the as-built condition. 'Ihirty I

additional supports have been checked by field personnel and no additional =

k problems have.been found.

J

[.

---_.e.,.U.

,m,

.e-.

---w.-,,-an,e ge, e-,

,,,%n.,-w--,_v,c,,,

,#--%,,,,.q,4,y-

,.m e.p+,.m-,.#-..

i+,,,wgy,--g-,.+.,c e. e, w w e (

I e

t u

a ATTACHMENT 2 CP&L Letter Dated May 22,1985 1

Sc/A-f CD&L Carolina Power & Light Company SERIAL: NLS-85-106 MAY 2 2 boo Director of Nuclear Reactor Regulation Attention:

Mr. D. B. Vassallo, Chief Operating Reactors Branch No. 2 Division of Licensing United States Nuclear Regulatory Commission Washington, DC 20555 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 & 50-324/ LICENSE NOS. DPR-71 & DPR-62 PIPING STRES5 ANALYSES DAMPING VALUES

Dear Mr. Vassallo:

Pursuant to the Code of Federal Regulations, Title 10 Part 50.55a paragraph (a)(3),

Carolina Power & Light Company (CP&L) hereby requests approval to utilize the damping curve developed by the Pressure Vessel Research Council (PVRC)in ASME Code Case N-411. Damping values extracted from the curve will be incorporated into seismic analyses for Operating Basis Earthquake (OBE) and Safe Shutdown Earthquake (SSE) events. The new damping values could be used for current piping modifications and future piping stress analyses as an option to the original damping values presented in the Final Safety Analysis Report (FSAR). The PVRC damping values will be used only for seismic response spectra analyses. They will not be applicable to time-history analyses.

If the PVRC damping values are approved for use at Brunswick, the following upgrades will also be incorporated when applying the new values:

1.

A three-dimensional square root of the sum of the squares (SRSS) earthquake combination will be used in lieu of a two-dimensional SRSS combination.

2.

Regulatory Guide 1.92 modal combinations accounting for closely-spaced modes will be used in lieu of a straight SRSS o' cli modes.

3.

A rigid cutoff value of 33 Hz will be used in lieu of 20 Hz.

7 4.

If, as a result of using the damping value curve presented in ASME Code Case N-411, piping supports are moved, modified, or eliminated, the expected increased piping displacements due to greater piping flexibility will be checked to assure that they can be accommodated and that there will be no adverse interaction with adjacent structures, components, or equipment.

r The original FSAR criteria for piping analyses, including Regulatory Guide 1.61 values, and the proposed PVRC damping with the upgraded criteria presented above will be considered as valid options for pipe stress analyses and modification work at Brunswick.

When performing an analysis, the damping values taken from the curve presented in Code Case N-411 are only to be used with tne upgraded criteria, not with the original FSAR criteria. That is, no analysis will combine damping values and criteria which are not consistent.

411 FayeMevine stree'

  • P O Box 1551
  • Rafeigh, N C. 27602 ffe f;z w + W -4 7/

4 Mr. D. B. Vassallo Page 2 Marked up copies of the affected FSAR tables and text (to be incorporated into the next FSAR revision) as well as the PVRC figure to be incorporated are attached.

Carolina Power & Light Company has reviewed this request in.ccordance with 10CFR170.12, and a check for $150 in payment of the required fee is enclosed.

Should you have any questions regarding this request, please contact Mr. Sherwood R.

Zimmerman at (919) 836-6242.

Yours very trul,

lWf '[O A. B. Cutter - Vice Presi ent Nuclear Engineering & Licensing ABC/RWS/mi (1339RWS)

Attachment cc:

Mr. L. W. Garner (NRC-BNP)

Dr. J. Nelson Grace (NRC-Ril)

Mr. M. Grotenhuis (NRC)

(

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TABLE 3. 7.1-1 DAMPING FACTORS _

t FEM PERCENT OF CRITICAL DAMPING _

OBE DBE Reinf orced Concrete:

'(a)

Primary Containment Structure 4

7 (b) Reactor Building and other j

Class I Structures 4

7 j

Steel Structures and Assemblies:

(Reactor Building & other Class I structures)

(a) Bolted or Riveted 5

10 1

(b) Velded 2

5 Vital Piping 0.S*

2*

Equipment 1

2 Soil - Structure Interaction Damping 4

7 i

For final reconciliation of pipe stress analysis or piping system backfits, damping values as defined in ASME Code Case N-411' (Figure 3.7.1.5) may be' utilized for both OBE and DBE.

1

.I 3.7.1-5 1

w

3.9-2.1.3 Piping Seismic Analysis The piping systems were dynamically analyzed using the " lumped mass response spectrum method" of analysis.

For each of the piping systems, a mathematical model consisting of lumped masses at discretc joints ccnnected together by weightless elastic elements was constructed. Valves were also considered as lumped masses in the pipe, and valve operators eccentricity was considered (Reference 3.9.1-1).

Stiffness matrix and mass matrix were generated and natural periods of vibration and corresponding mode shapes were determined.

Input to the dynamic analyses were the applicable 0.5 percent damped acceleration response spectra.

Increased damping values may have been applied for final stress reconciliation or piping system backfits ia accordance with /.ShE Code Case N-411 (Figure 3.7.1.5).

If so, the following criteria were also used?

a)

A three-dimensional square root of the sum of the squares (SRSS) earthquake combination in lieu of a two-dimensional SRSS combination.

b)

Regulatory Guide 1.92 modal combinations accounting for closely-spaced modes in lieu of a straight SRSS of all modes.

c)

A rigid cut off value of 33 Hz in lieu of 20 Hz.

d)

A pipe displacement check performed if existing pipe supports were moved, modified, or eliminated.

The increased flexibility of the curved segments of the piping systems was considered. The results for' earthquakes acting in the X and Y (vertical) directions simultaneously and 2 and Y directions simultaneously were computed separately. The maximum responses of each mode were calculated and combined by the absolute sum.

The response thus obtained was combined with the results produced by other loading conditions to computer the resultant stresses.

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3.9.2-6 (1483RwS/ccc )

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TABLE 3.7.3-1 CRITICAL DAMPING FOR STRUCTURES, PIPING, AND EOUIPMENT Item PERCENT OF CRITICAL DAMPING

~~

OBE DBE 1.

Concrete Structures 4%

7%

2.

Piping I#a %

  • 2%
  • 3.

Equipment a.

Pumps b.

Motors 1

2%

c.

Switchgear d.

Exchangers e.

Tanks f.

Estteries and Racks g.

Cable ray Systems h.

Diesel canerator Units j

i. Othe rs l

4 Cranes 4%

7%

l l

For final reconciliation of pipe stress analysis or piping system backfits. damping values an defined in ASME Ccde Cace N-411 j

(Figure 3.7.1.5) may te utili::ed for both OBE and DBE.

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Frequency (Hz)

Figure 3.7.1.5 Damping Value for Seismic Analysis of Piping (Applicable to both OBE & SSE, Independent of Pipe Diameter) 4 4

ATTACHMENT 3 A

h NRC Letter Dated August 28,1985 t

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1 NUCLEAR REGULATORY COMMISSION 6

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August 28, 1985

/ll/ S tf.S~- 6 d 8 Docket Nos. 50-325/324 Mr. E. E. Utley Senior Executive Vice President Power Supply and Engineering & Construction Carolina Power & Light Company Post Office Box 1551 Raleigh, North Carolina 27602 i

j

Dear Mr,

Utley:

SUBJECT:

USE OF ASME CODE CASE N-411 Re:

Brunswick Steam Electric Plant, Units 1 and 2 By letter dated May 22, 1985, you requested approval to utili:e the damping curve developed by the Pressure Vessel Research Council in ASME Code Case N-411. Code Case N-411 damping values would be used as an options to the original damping values presented in the Final Safety Analysis Report (FSAR).

Your letter dated May 22, 1985 submitted the information and commitments. We have reviewed your request pursuant to 10 CFR Part 50.55a paragraph (a)(3) and find that, although Code Case N-411 has not been listed in Regulatory Guide 1.84 or 1.85, it is acceptable based on the information and commitments provided in your application.

Our related Safety Evaluation is e>1 closed.

Sincerely, g

pOfficeofNuclearReactorRegulation Harol en on, re tor

Enclosure:

As stated I

cc w/ enclosure:

See next page h.

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SAFETY EVALUATION tv THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO USE OF ASME CODE. CASE N-411 CAROLINA POWER & LIGHT COMPANY BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 4

4-DOCKET NOS. 50-325 AND 50-324 i

i

1.0 INTRODUCTION

ByletterdatedMay)22,1985,(CP&L, the licensee, pursuant to the Code of Fe the-Carolina Power & Light Company Part 50.55a paragraph (a)(3), requested approval to utilize the damping curve developed by the Pressure Vessel Research Council (PVRC) in ASME Code Case N-411.

It further stated that damping values extracted from the curve will be incorporated into seismic analyses for Operating Basis Earthquake (0BE) and Safe Shutdown Earthquake (SSE) events; that the new damping values could be used for current piping modifications and future piping stress analyses as an option to the original damping values presented in the Final Safety Analysis Report (FSAR); that the-PVRC damping values will be used only for seismic response spectra analyses; and that they will not be applicable to time-history analyses.

2.0 EVALUATION We have-completed our review of the CP&L request for authorization to use i

the damping values in ASME Code Case N-411 foriapplication'in the response.

spectrum seismic analysis for current modifications and future stress analyses of piping systems at the Brunswick Steam Electric Plant, Unit 1 l

and 2 as discussed in.the licensee's. letter NLS-85-106 dated May 22, 1985, The damping values in Code Case N-411 could be used as an option to the i

l original damping values prdsented in the FSAR.

^

1 I-Code Case N-411. " Alternate Damping Values'for Seismic Analysis of Piping Section III, Division:1, Class 1, 2 and 3 Construction" is a conditionally l

acceptable Code Case and is approved by the staff for specific plant i

applications pending revision-of Regulatory Guide 1.61.

Utilities wishing -

l to use this Code Case shall submit in'their request the following l

information or comitments:

(a) Comit to. use the case for piping systems analyzed by response spectrum methods and not those using time-history analysis methods.

(b) Indicate if the case is to be used for new-analyses or for reconciliation. work and for support optim17ation.

N y 9 ? C O 5 n A (-

P

. (c) Due to the increased flexibility of the system comit to check all system predicted maximum displacements for adequate clearance with adjacent structures, components and equipment, and that the mounted equipment, can withstand the increased motion.

(d) When the alternate damping criteria of this Code Case are used, they will be used in their entirety in a given analysis and shall not be a mixture of Regulatory Guide 1.61 criteria and the alternate criteria of this Code Case.

CP&L has complied with these comitments in the letter of May 22, 1985.

Therefore, since the commitments with respect to the Code Case N-411 are documented in the referenced letter, the staff finds the licensee's request to use Code Case N-411 acceptable for use at the Brunswick Steam Electric Plant, Units 1 and 2 in the response spectrum seismic analysis of piping systems.

In an attachment to the letter of May 22, 1985, the licensee has included marked up copies of the affected FSAR Tables 3.7.1-1 and 3.7.3-1, page 3.9.2-6 and figure 3.7.1.5 which will be included in the next FSAR revision.

The staff finds this material acceptable.

In addition, the licensee has stated that the following upgrades will also be incorporated when applying the damping values of Code Case N-411:-

(1) A three-dimensional square root of the sum of the squares (SRSS) earthquake combination will be used in lieu of a two-dimensional SRSS combination.

(2) Regulatory Guide 1.92 modal combinations accounting for closely-spaced modes will be used in lieu of a straight SRSS of all modes.

(3) A rigid cutoff value of 33 Hz will be used in lieu of 20 Hz.

The use of these upgrades for use at the Brunswick Plant, Units 1 and 2 are consistent with design methodology being accepted by the staff for plants currently undergoing licensing review and is thus acceptable.

3.0 CONCLUSION

Based on our review and the above discussion, we find that the CP&L request to use the iamping curve in ASME Code Case N-411 is approved as requested.-

Principal Contributor:

R. Kirkwood Dated: August 28, 1985 l

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i Al TACHMENT 4 e

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i NRL Meeting Summary Dated June 12,1979 4

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JUNE 121M Dockot Hos. 50-325 and 50-324 MEMORANDUM FOR:

Thomas A. Ippolito, Chief, ORD #3, 00R FROM:

Roby B. Bevan, Project Manager, ORB #3, D0R

SUBJECT:

MEETING

SUMMARY

A meeting was held with representatives of Carolina Power and Light Compar1y (CP&L) and their contractors in Bethesda, Maryland on May 30, 1979.

The purpose of the meeting was to provide the staff with the current status of the seismic pipe stress reanalysis being conducted for Brunswick Units 1 and 2 required for IE Bulletin 79-07, and to review the licensees program of identifying and modifying overloaded pipe supports and hangers found incident to the reanalysis.

A list of attendees at the meeting is enclosed. Also enclosed is a C?&L letter dated Itay 29. 1979, with attachments, addressing the items to be dis-cussed at the meeting. On May 25, 1979, CP&L inforned D0R staff by telephone that, in the course of their reanalysis for IE Bulletin 79-i:7 they had found six supports stressed beyond code allowabic.

On lenning of this situation, they shut down both " nits and infonned the 00R staff and 1&E of their action.

By telephone on May 28, 1979, CP&L requested this May 30, 1979 rr.eet ing.

At the meeting, CP&L explained that the six supports had not been adequately designed to account for the torsion stress on I-beams, a condition that might exist fc some oth er supports in the plant.

They had, therefore, initiated a program to inve;tigate all pipe supports on all safety related systems.

This program would be completed and all modifications made, before the two units would begin a return to power operation.

CPat. also described modifWtions in their reanalysis program, both for already-analyzed and for be-analyzed cases, to accanmodate the staff position regarding the BSEP Final Safety Analysis Report commitment to use absolute summation techniques with the two dimensional model.

The Staff agreed to review their proposal, and infonned them by phone on

~

June 1,1979 tnat their method is an acceptable one.

CP&L nas previously committed (letter dated May 22,1979) to verify as-built dirensions f or all safety related piping in and suppcrt systems.

They reported that this program of " walking the lines" is in pr ogress.

During the current shutdown, all lines inside containment will be verified.

The reanalysis of the lines is being done concurrent with the as-built verification.

CP&L described procedures for handling deviation of as-built fran as-analyzed.

h&%W2-/ o*

Thomas A. Ippolito JUNE 1: 079 l

h, j

In response to a previous staff inquiry regarding the location of the postulated LOCA pipe rupture relative to the highest stress point, CP&L informed the staff that the piping design did not locate the break at the highest stress point, but instead analyzed for a double ended or a longi-tudinal break to occur at any point on the line, both inside and outside cont ainment.

In response to a previous staff request, Cpal discussei their program of mtnagement controls and reporting criteria to assure appropriate licensee action when problems are identified in the continuing program of pipe and support reanalysis.

CPAL expresse1 their intention to return to operation in a few days.

They 5{

therefore regaested (and we agreed) to meet with us again on June 4,1979 to review their status.

j Specifically, they would at that time provide the
I current status of their pipe and supports analyses, identifying modifications yet to be completed (if any), and verifying the as-built condition of all lines procedures for handling deviations found in their ongoing as-built verification program.

Roby B.

evan, Project Manager Operating Reactors Branch #3 Division of Operating Reactors

Enclosures:

As stated e

en d

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4 ATTACHMEf4T 5 1

1 NRC Letter Dated January 30,1985 i

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RECEIVEC FEB 4 585 4,"*,<

Janua ry 30, 1985 j

ypt-b - O nnefat Nos. 50-3?;/324

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Mr. E. E. titlev Executive Vice Presidant j

Carolina Power 1 Licht Company l

Post Office Pox 1551 Daleich, North Carolina 77607 l

Daar Mr. litlev-1 i

silRJECT! WASONRY WAt.t. DES'GN, IE BtILt.ET1H 80-11 Re:

Brunswick Steam Electric Plant, Units 1 and 2 3

~!

On vav B, 1980 we issued IE Bulletin 80-11. We have reviewed your reson' set as listed on pace 18 of the Technical Evaluation Report (TEDi, n

l attached to the enclosed Safety Evaluation (SE).

In addition, we have rrviewed your response dated December 71, 1984 Based on our review, we i

find that the proposed modifications are in compliance with the staff acceptance criteria. We find the schedule for comoletion'of-the fix designs proposed in your letter to be acceptable. We are aware that tiie schedule

{

for the actual modification work will he worked out in your 5-year proprem i

for the Rrunswick plant, i

This concip h nur review of your response to IE Pulletin R0-11.

1 Sincerely, 1

ff*.&,l ' a Domenic B. Vassallo, Chief i

Operating Resctors Branch #2 Division of Licensing

Enclosures:

As stated l

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See next page s

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j SAFETY EVAltlATION RY THE OFFICE OF NtlCLEAR RFACT09 REGULATION i

PELATED TO NASONDY WALL DESIGN, IE PULLETIN RD-11 l.

CAROLINA PCWER A LIGHT COMDANY l

RRUNSWICK STFAM ELFCTPIC PLANT, UNfTS 1 AND 9 l

DOCKET NOS. 50-3?S/3?4 I

j introduction and Background i

i IE Bulletin R0-11 reenrdina Masonry Wall Desian was issued on May 8, 1980 3

The Carolina Power A'i.icht Company' (CPAL) responded with letters' dated July

7. November 5 and PS, and December 0, 1980.

In response to requests for '

additional information, dated Auoust 2, 1982 and February Pl. 1984, responses dated.1uly 29, 1983 and April P7, 1984 were tuhmitted. A final responsa dated December 21, 19P4 was received and reviewed after the l

Technical Evaluation Report (TER) Attachment I was completed and therefore 1

is not included in the reference list on page 18 of that report.

The findings repneted in this Sarety Evaluation (SE) are based on the attached TER, prepared by Franklin Research Center (FRC) as'a contractor to l

NRC, and the NRC review of the December 21, 1984 submittal. This TER contains the details of construction techniques used, technical information reviewed.

acceptance criteria, and technical findings with respect to masonry wall i

construction at the Brunswick units. The staff has reviewed this TER, l

concurs vith its technical findings and it is hereby incorporated into this i

SF.

It is noted that on pape 3 of the April 27, 1984 response (RAI Sbi that the NDC position paDer on tha energy balance technique is referenced.

In f

view o' the CPAL decisinn not to use this technique, this position paper was not needed.

1 The staff sumary and evaluation of the major technical conclusions are included in this SE.

i 4

l Evaluation and Conclusion i

There were 87 masonry walls identified at the Prunswick units. The licensee qualified sixty of-these safety-related walls by using the working stress approach which is in compliance-with the staff acceptance criteria. The licensee planned to modify ten walls, which were originally qualified by-nonlinear: +echniques, by providing steel pilasters, steel prating to restrain walls, and steel anales installed at their boundaries. These.

modifications would render these ten walls in compliance with the staff acceptance criteria.

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The December ?1.1984 submittal, received after the TEP was completed, indicated the results of the evaluation of the addi+ional 17 walls referred to in the TER. That submittal also indicated that a field inspectinn determined that two of the 87 walls were "Nonsafety Related* rather than

" Safety Related." One of the latter walls was listed amono the ten walls to 1

j be uporaded and one was among the last 17 to be evaluated. Five among the 17 walls were found to be within allowable linits and the remaining 11 will i

raouire additinnal reinforcament to bring them within allowable limits, j

That makee a total of 20 walls reouiring modification.

j in sunnary, there was a final total of 85 safety-related masonry walls identified at the Brunswick plant. Sixty-five were found to be within

{

ecceptable liniis and 70 will require nodification. The design fix for nine of the PO walls is completed, the desian fixes for the remaining 11 are.cheduled to he completed in July 1985. The schedule for the actual nodifications will be in the 5-year plan currently under review, Pated on the above findinos, the staff concludes that Items P(b) and 3 of

)( Pulletin 80-11 have been fully implenenten at the Brunswick units and that there is reasonable assurance that the. safety-related masonry walls at the Bruntwick units will withs+and the specified desion load conditions' i

without impairment of f a) wall integrity or (b) the performance of required safety functions.

Principal Contributor:

N. Chokshi and M. Grotenhuis I

i hated:

Jeouary 30, 1985 f

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