ML19322A370

Responds to NRC 790917 Ltr Re Four Items Requiring Resolution Prior to Plant Restart.Areas Include safety- Related Piping & Anchor Bolts on Base Plates of safety- Related Pipe Supports

ML19322A370
Person / Time
Site:	Surry
Issue date:	10/04/1979
From:	Spencer W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
References
817, NUDOCS 7910100220
Download: ML19322A370 (7)

Text

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s VIHOrNIA Er.ncraic ann PownN CObf PANY Ricnw oxo.Vs uorn:A 2020 October 4, 1979 Pir. Ja-es P. O'Reilly, Director Office of Inspection & Enforcement D Cy Serial No. 817 PSE&C/Ct1Rjr:mac: wang U.S. Nuclesr Regulatory Comission Wb Region II 101 Marietta Street, Suite ~i100 o }g gt Docket No. 50-280 bJ1J.!Saat Atisnta, Georgia 30303 -l.icense No. DPR-32

Dear Mr. O'Reilly:

RESPONSE TO I.E. LETTER DATED SEPTEM3ER 7, 1970 SURRY POWER STAT [dN UNIT 1

. ,e purpose of this letter is to provide responses-to your letter of September 7, 1079 where you identified four items that required resolution prior to Unit I being returned to service. Our responses are as follows:

1. . Seven piping runs were selected for analysis in response to item one II) of your letter of Septem5er 7, 1979. These piping runs meet the requirements of it s ene in the letter in that they are samples of safety-related oiping systems six inches in diameter or less. In addition, these particular piping runs have the following characteristics:

a. These piping runs are located inside the Reactor Containment Structure and are deamed inaccessible during plant operation.

h. This piping is required for Safe Shutdown as defined and identified by Vepco station personnel.

It is our understanding that the selection of these seven piping runs has been reviewed by NRC personnel during discussions with Vepco at the Surry Power Station on September 14,: 1979. These seven piping runs have been analyzed as five' sample problems: '

System ES No. Problem No.

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Safety Injection ' '

f1105'A?'

3000 1105 A3 Safety In'jection 1106 Al'

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. 3003 1106 A4 Resistance Temperature Detection 1103 AS 3001 Resistance Tempe*ature Detection 1103 A4

• 3002 Resistance Temperature Detection 1103 A5 3004 D

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, Mr.' James P. 0'Re111y, Director 2

~. It should be noted that an MKS is a present day isometric piping

. configuration drawing, whereas an MSX is the original piping configuration

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drawing.

The con'pletion of the analysis .of these seven piping runs coupled with the analysis; performed pertaining to the March 13, 1979 Show Cause Order will 4

complete the reanalysis of the Safe Shutdown piping runs 2 1/2 inches diameter and_ greater inside the Reactor Containment. For the purpose =of ,

seismic review,.-Safe, Shutdown requirerpents are. considered to be all.the

, Engineered Safeguards' as defined in the FSAR plus Auxiliary Feedwater and Residual Heat Removal wit at lesst;one heat exchanger.

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We are advised by the St6nd WWebster Enginee$1'ng Corp' oration that during the . design of >the Su.rry Power Station, the Boston Stone 8 Webster office >

prepared original MSK drawirigssto show support locationsf and functions for' 21/2 to 6 inch' pipe.. The MSK &awings were' based upon approved pipe drawings (FP drawings) with additional infohnition pertaining to support locations and loads. These MSK drawings were transmitted to the

construction site for use by the hanger designers', and to the fabricator of shoo supplied hangers, if, applicable, to develop construction drawings. Only.a limited number of these construction drawings have taen recovered to date.

The pipe stress problems for the present sample program tre based on. site

. generated ismittries of the oiping . systems (MXS @awings) as they presently exist in the plant .(as opposed to original MSK drawings). These MKS drawings show field verified pipe dimensions, locations and function of the hangers, and include _small diameter (21" vhes and less) branch piping to either the first anchor or second resu t. In addition, field as-built data is developed for all supports on 2 W to 6 inch piping.

The field MKS's are compared to presently avai:cble design documents in order to identify differences. Where significant differences are found, 3

the configuration defined on the field MKS's are reanalyzed using simplified methods (and in some cases NUPIPE computer methods) using approved criteria.

The analysis that is presently underway on the seven safe shutdown piping i . runs in the-sample is in accordance with the criteria and methods that have been reviewed with NRR during~their evaluation of.the March 13, 1979

'Show Cause reanal'l sis. These methods are consistent with the Surry FSAR c

and the recent " Safety Evaluation by the Office .of Nuclear Reactor

-Regulation",7and'have been accepted by the NRC. ,

Thefivesampleplroblemswerereviewed,as.outiined-inthefollowing.five

- steps: , ,

~1) :A comparison of.the'present day MKS's was made to.the'available;

' ~ design documents.(MSK's).. This comparison; indicated that the

following~. supports are not installed in accordance with the' ,

> presently available design documents:

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'3, Problem  ;

.Line Present Day.

Configuration N o .' -System . Designation .

3000 SI 3"-51-147-1503 Original AH moved around ' '

elbow and changed to strut.

3003 . SI -3"-51-145-1503 1 Original RH1 moved around >

elbow and. changed'to , .

strut.- .

,:(1). VC/L'C missing -

3001 RTD- '2"-RC-146-1502 .

.2"-RC-149-1502" '(1) LC" missing /(1) VC .

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.(?)VC/LC(m'issing

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-(1) VC/LCfadded 300? ,RTD 3"-RC-}16-150?1 -(1 , Anchor < missing

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VC/LC missing -

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"2"-RC1117-1502~ ' ((11)'VC niissing

'1) VC/LC missing

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3004 RTO ~'2"-RC-132-1502~ (2) LC missing 3"-RC-131-1502** (1) Anchor missing (1).VC/LC missing AC = ' Axial lCohstraint. .

VC='-Vertical'Constra' int

LC =' Lateral Constraint- "

'RH = Rod ilanger ,

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2) The, presently existing HKS's~for the five problems'were reviewed

• using simplified analysis methods'.- The results of this'. review '

showed'that additional supports were.. required las follows:- ,

Problem No. ,

' System' '

~ Required.- ,

3000 >

'Sk: 4 1 Anchoe, ~'lVC/Lb, ,

1. Guide-

' ?3003 SI 3 Guides;.;llLC, 1 Anchor: 1'

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- 23001, 3002,A- RT" , 1 LC, 1sVC, 1;VC/LC for ,

- 3004 ,

f. ~each problem .

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'3)' !Since;the;simplifieit method is; overly conservative, these five ,

- problems were then modeled and.run 'on the NUPIPE4 ciomputer _ ^ '

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. program using presentlyiexisting physical data'and approved: ..

O ~ cc ' ~ criter,ia .as previously described;" -

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LThe three RTD, bfpass lines are the most . difficult small pipe n systems in.the plant to layout and support becauseJof the.

thermal expansion of the pipe involved and because of the thermal motions imposed at tefeinations (and joints) of the:

pipe. In addition to controls on arrangements to meet stress.

the. NSSevendor,. technical requirements control relative ~

- criterian'thekbypas'sillegt, length'of 1ccat.ionion the loop' piping, and 3 absolute lengths of some segments of' the bypass legs. .

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The seven MKS?sidetail;579 linedfeet ofipipe of-which 273 feet >

are 2 1/2 to 6 inches and the 7emaining'305 lineal feat.are' ,

branchipiping4, inches.and smaller. The analysis using NUPIPE indicated that'all 273flineal 'fest of: piping L1/2 to 6 inches I w'a's Veithin allowable str'ess limits.' 2, l 4j Less than 3 feet (or less than one percent) of the branch pipe 2 .

i- inches and smaller exceeded the allowable deadweight stress -

limits when multiplied by a stress intensification factor. .> >

Multiplication of non-cyclic stressL(such as_ dead-weight Stress) by an; intensification factor is a conservatism not explicitly '

i required by the B31.1' code,1955. .Also, allowable stresses were computed using minimum material properties, whereas actualf ,

! properties shown by test results typically eiceed minimums such s

. that an allowable stress calculated from the real' material '

strength would be significantly higher. '

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None of the pipe > exceeded the' allowable thermal stress > and most ,

was determined to be less than half of allowable thermalt '-

! stress. The pipe was therefore well supported:to preclude ..

~'effect of therma,1 stress which is continuously;present in the  !

" RTD t ~ pass loops._ - .

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s Stress from the postulated earthquake condition exceeded

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allowable; in-only 3. feet of branch pipe 2 inches and sm' aller '

when calculated'with a low damping factor. These stres'ses were.

.such thatislightly higher damping values would re' ally act to? 7 '

reduce vibratory response. Using current ASME code criteria,the stresses would not exceed real~ ultimate strength,:so no' failures ','

. would occuriand public safety"would not be affected, although-

safety margins are r, educed. -

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The computers pipe stress analysis lof thel preserdly existing . ,- i

- configurations.using;NUPIPE indicated that. fewer,added supports:

i. / are required toimeet approved criteria than was indicated'by the t

conservative simplified methods. The 'added supports required by mL

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"the NUPIPE! analysis,Jusing approved Show Cause criteria,-are;asi GJ q ,

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- Additional Supports ,.

. Problem No. Svsten ' Required v' 3000 ' SI- Hone 3003 SI 1 suoport:to be removed

-3001 RTD 3 supports .

3002 RTD 3 supports 3004 RTD 3 supports.

4) The supports to be added will be based upon final approved

, NUPIPE stress runs in accordance with approved criteria.

5 )' ' Existing supports'en 2 1/2 to 6 inch piping are being reviewed and will be rean'alyzed using loads from the NUPIPE analysis, or in some cases to generic loads using approved criteria. The analysis using NUPIPE computer methods and approved Show Cause i

. criteria indicates th.at all 2 1/2 to.6 inch piping in the five-sample problems meet allowable stress limits. The branch piping 2 inch and smaller; which was necessarily incluaed in the computer model for.the larger piping, also met allowable stress limits with the exception'of. about one percentief. the small .

.' ' piping. We do 'not believe that safety is impaired and believe that the piping is fully capable of performing its function in.

the event of a design earthquake. However, any modifications ~

identified in the five sample. problems as a result of the use.of '

Show Cause criteria will be completed prior to startup.

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2. Anchor Bolt Test Program ST-75 was conducted to' inspect sanples- of 'all size anchor bolts'(Phillips Red Head Self-Drilling' Anchors) on baseplates- '

of. pipe supports for safety-related piping. . The objective of the prograi was to inspect approximately 200 bolts by measuring the anchor size, depth

' of set plug, shell projection, and thread engagement, and by . torquing the- "

bolts;to.a.value that would demonstrate a' factor safety of two. The , ,

. initial attempt was to obtain the sample by testing .one bolt per -!

accessible support inside.the containment, excluding those supports

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previously tested under the Anchor Bolt Sampling Program, ST-39, and  ;

excluding;those~ supports which had anchor bolt modifications made as a. ,

result _of the NRC Order to Show;Cause. Due to the: limited size of the , '

> sample obtained,J the criteria were expanded .to include.on,el bolt per accessible. baseplate inside the containment, allowing more than one. ,

'haseplate per support to be i_nspected and allowing inspection of different' anchor. bolts on those: supports inspected undar ST-39.JA sufficient number "of anchor boltsloutside.the 7containment was inc16dedito bring the totai

' sample size.to 'approximately 200 ~ anchor bolts. Of th'e final sample size. .

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of 209 b'olts, 35 were outside the containment. This methodology for: '

selecting bolts-was; discussed'with your inspector, Mr. R.(M. Compton, and it. was . agreed that the main objective.was not' toiselect bolts' previously'-

inspected nor any,new bolts.'iThe sampling program met this objective.

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6 1 3 a s The results;of'the Anchor Bolt Test Program show that 9S.2% {199 anchor

- holts).of the' sample . size of 209; anchor bolts were acceptable and have demonstrated a factor of safety ' of two? This sample size: represents 195 '

- 1baseplates on 164 supports. Ten anchor bolts were found not to be

- - Lacceptable' for the following reasons. Two anchor bolts did not'have sufficientithrea'd engagement, one anchor (bolt hadino thread engagement and one anchor bolt had only 1/8" of. thread engagement.' .Six anchor bolts had the bolt and'shell turning together. Five;of these were on one-b'aseplate r 4 end rotated 'when the back-off torque was ' applied in attempting. to remove .

the bolt and the: sixth anchor bolt rotated:during;appifcation of the proof

< load torque. Two ' anchor bolts were found on one. baseplate with the shells: .

rusted to the extent that a back-off torque off'250.ft.-lbs. was' recessary to remove the bolt and no measurements could be obtained for the shell or i depth of set plug. Results of this program will be maintained at Surry Power Station as Special Test-75 should you require any further l

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4 4 inspection. The detailsief the results1have,been= r,eytewed< by your Messrs.

E. J. Vallish and R. H. Compton; ' All' deficiencies? identified in this program will be corrected prior to re,tur,ni,ng the Unit (to servic'. e s .

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3. We are incorporating flexibility considerations..into'the-analysis of the concrete' anchors forlthe thirteen pipe supports with baseplates-in the five sample piping runs.) Current; support, loads are preliminary.since pipe stress analysis,is sincompletc.;at this time.l 4eyerthelessewe-are p ,

.oroceedingiith, flexibility review per the Stone & Webster! simplified 50-STEP procedure or more sophisticated computer analysis, and will use '

the final oipe stress loading to confirm the preliminary baseplate results. Flexibility analysis'is complete on five of these supports and, results are# acceptable using . preliminary loads. All reanalysis -is being.

F performed using field verified piping and support information. If

' hardware modifications are required as analysis continues, appropriate action will be taken in accordance with the plant Technical Specifications.

4 An additional number of supports'are beingfevaluated fordflexibility.

considerations as a result-of a previous; bolt sampling program.

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Approximately 94 supports are involved with this program with evaluations "

complete on 39.. A total of 35 supports have been found to1be acceptable _-

with flexibility criteria included ~as described above. .Any baseplates or, ,

anchor-bolts found to be overstressed dueitoLfinal (NUPIPE). loading ~n .

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• conditions with computer analy"s for the baseplates will .be subject to a appropriate. action in accordanes with the; plant Technical. Specifications. -

, 4. SpecialTest15: NhichcoveredtheAnchorBoltTest'Programincluded.a '

baseplate visual? inspection. The visual inspectio_n : involved a walkdown of 1,, all seismic Category;I piping systems inside the containment covered by:

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. the NRC Order, to Show Cause and I.E. Bulletin 79.-14 to look for any gross-

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N defects: such:asimissing or loose bolts and< nuts, and stewed bolts which, ,

would adversely; affect thesability of th'e sup' port to perform i_s.. intended ,

- ~ ;functioni A total- of.88 piping isometric 1 drawings 4MKS's) were walked' *

,dewn in the? inspection. Fifty-onel drawing's~'w ere for the;NRC Order to Show?

Cause. and 37: drawings were for the piping covered by:I.E. Bulletin 79-14. _

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. All noted discrepancies were tabulated and the line rewalked to verify.and; evaluate the discrepancies. These, items;which;were found to bc deficiencies incisde five U-bolts ihichJwerelloose. three baseplates which were not flush with the concrete surface, four baseplates with oversize-

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bolt holes which required piste

' Linstalledat.'an'angleand_ requip [Eashers,one.springhangerwhichwas' ed. realignment;, one bolt missing in.a

• U-bolt, one baseplate which required regrouting, ch6 skewed b31t, one. "

' anchor bolt missing,1 two rod hangers not connectedsto:the piping, .one ,

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'oneb'rokenanchoribolti,'a'ndSnewasherinstalled washer behind a which baseplate.' sas'Five need*14, lpipe suppbrtifiscrepancies were found on:the MK drawings prepared for.use in complying with I.E. Bulletin 70-14 These discrepancies include.one U-bolt and.one support which.were shown on the drawings but not installed, and two supports and one U-bolt which werej , , ,

installed but not shown'on the drawings.; These discrepancies will be

' reflected in the finalized MKS drawi,ngs which will:be used for analysist under I.E. Bulletin 79-14. ,

i A repair procedure was wr.itten as part of ST-75 and included repairs for those deficiencias noted in the visual . inspection as well .as those I

deficiencies which resulted from or were noted as a result of the Anchor' Bolt Test Program.. All of'these deficiencies wil.1 be corrected prior to i- start-up.

! The deficiencies found were small in numSer. Therefore, we believe the failure rate of the visual inspection ,to be acceptable. The results of the: visual inspection program have been reviewed by your Mr. E. J. Vallish.

, 'Approximately I?.5.MKS's make up the I.E. Bulletin 79-14 scope cf work.

Seven of these have been analyzed as the sample problems. We believe the remaining 118 MKS's can be analyzed in about five months and we will make j every'e.ffort to expedite this schedule. If during this review any j modifications are identified, they will be reported-to the station and l evaluated under tne requirements of the plant Technical Specifications.: ,

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< Very trulv yours, 4

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W. C. Spencer- .

. Vice President ~ Power Station

' Engineering and Construction Services ,

cc: Mr. Victor'Stell'o,. Director ~ .

y Office _oftInspection~and EnforcementL 1

' Mr. . Harold R. Dent' o n,t Director . .

Office of Nuclear" Reactor Regulation <

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