ML18136A034

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Forwards Evaluations Requested 790905 of All Plants Previously Using Pipe Stress Analyses Based on Unacceptable Algebraic Summation Techniques.All Stresses within Allowable Limits
ML18136A034
Person / Time
Site: Nine Mile Point, Brunswick, Surry, Robinson, Cook, Zion  
Issue date: 09/18/1979
From: Schwencer A
Office of Nuclear Reactor Regulation
To: Whittine J
VIRGINIA, COMMONWEALTH OF
Shared Package
ML18136A035 List:
References
790918, NUDOCS 7910100044
Download: ML18136A034 (23)
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SEPTEMBER 1 8 1979 Di stri butj_on

("'boc_Ke~~~~~:)ttorney, OELD

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ACRS (16)

Docket Nos. 50-280 V. Stello and 50-_281 Mr. James Whittine State Corporat'ion Commission Commonwealth of Virginia P.O. Box 1197 Richmond, Virginia 23209

Dear Mr. Whittine:

B. K. Grimes D. Eisenhut R. Vollmer T. J. Carter W. Russell A. Schwencer C. Parrish D. Neighbors At the meeting with us on September 5, 1979, to di_scuss *the Surry Power Station~ you.requested that we provide you with exampJes of licensees*

  • justifications for continued plant operations even though it was dis-.

covered that the plant design had been based on pipe stress analyses,

which. had used unacceptable algebraic summation techniques.

lrJe have evaluated _all plants which used _algeb_ra"ic summation and we have enclosed some sampl~s of our evaluations.

Enclosed are the NRC evaluations for the following plants.

H. B. Robinson, No. 2 Zion, Nos. l *and 2 Nine Mile Point D. C. Cook Brunswick May 22~ 1979 July 5) 1979 June 19, 1979 June 22. 1979 June 7, 1979 Some of these plants, such as Robinson9 may have been shutdown *at the time of* issuance of IE Bulletin 79-07.

You will note that in each of these**cases, the licensee was able, through reanalysis) to demonstrate that stresses are within allowsbl~ limits.

I trust these evahiations are responsive to your request.

~incerely, 1\\"\\ \\:';.

J ',,;,.':

  • J I,

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Enclosures:

Original Signed Bt A. ~chwencer, C~fef Operating Reactors Branch #1 Divisi6n of Operating Reactors (5)

MRC Safety Evaluations 1 8J,/' ~

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for 1 ants 1 i sted above

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Docket Nos. 50-280 and 50-231 Mr. James Whittine State Corporation Commission Commonwealth of Virginia P. 0. Box 1197 Richmond, Virginia 23209 Dear Mr. Whittine~

At the meeting with us on September 5, 1979, to discuss the Surry Power Station, you requested that we provide you with examples.of licensees*

justifications for continued plant operations even though it was dis-covered that the plant design had been based on pipe stress analyses

  • which had used unacceptable algebraic summation techniques. tJe*have evaluated all plants which used algebraic summation and we have enclosed some *samples of our evaluations.

Enclosed are tiie MRC evalua-tions for the following plants *

. /

H. B. Robinson, No. 2 Zion, Nos. land 2 Nine Mile Point D. c. Cook Brunswick May 22, 1979 July 5, 1979 June 19, -'1979.

June 22, 1979 June 7, 1979 Some of these plants, such as Robinson, may have been shutdown at the time of issuance of IE Bulletfn 79-07.

You will note that in each of

  • \\

these cases*, the licensee was able~ thf'ough reanalysis, to demonstrate that stresses are within allowable limits.

I trust these evaluations are responsive to your request.

Sincerely; Original Signed By; A" Schwencer ~ CJii ef Operating Reactors Branch #1 Division of Operating Reactors

Enclosures:

(5)

NRC Safety Evaluations for lants listed above.

OFFICE~

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  • e UNITED STATE;S.

NUCLEAR REGULATORY COMM*ISSION

_. _WASHINGTON, o. C. 20555

. September 18; 1979

  • Docket Nos. 50~280 and 50~281 _

Mr. James Whittine

  • State Corporation Commission Commonwealth of-Virginia P. 0. Box 1197 Richmond, Virginia 23209

Dear Mr. Whittine:

At the meeting with us on September 5, 1979, to discuss the Surry Power Station, you requested that we provide you with examples of licensees*

justifications for continued plant operations even though it was dis~

covered* that the plant design had been based on pipe stress analyses which had used unacceptable algebraic summation techniques."*

We have evaluated all plants which used algebraic summation and we have enclosed some samples of our evaluations.

Enclosed are the NRC evaluations for the following plants.

H. B. Robinson, No. 2 Zion; Nos. 1 and 2 Nine Mile Point D. C. Cook Brunswick May 22, l-979 July 5, 1979

  • June 19, 1979 June 22, 1979 June 7, 1979 Some of these plants, such as Robinson, may have been shutdown at the time of issuance of IE Bulletin 79-07.

You will note that in.each of these cases, the licensee was able, through reanalysis, to demonstrate that stresses are within allowable limits.

I trust these evaluations are responsive to your request.

Sincerely, 7 a ;Jj,1{tc/trv

(

A. Schwencer, Chief*

Operating Reactors Branch #1 Division of Operating Reactors

Enclosures:

(5)

NRC Safety Evaluations for plants listed above cc:

w/enclosures See next page_

I -

.*e*

Mr. *w. *.L~ *Proffitt**

Virginia Electric and Power Company.

cc~ Mr. Michael W. Miupin Hunton ano Williams Post Office Box 1535

  • Richmond, Virginia 23213 Swem Library College of William*and Mary Williamsburg, Virginia 23185 Donald J. Burke U. S. Nuclear Regulatory Commission Region II Offic~ of Inspection and Enforcement 101 Marietta Stre~t, Suite 3100 Atlanta, Georgia 30303. ~eptember 18, 1979
  • UNITED STATES
  • NUC.LEAR REGULATORY COMMiSSION

.. ~

MAY 2 2 1979 Docket No. 50-261 Mr. J. A. Jones Senior Vice President*

Carolina Power and Light Company 336 Fayetteville Street Ra.leigh, North Carolina 27602

Dear Mr. Jones:

This is in* response to your letters of April 27 and May 21, 1979, wherein you reported the results of a recent seismic reanalysis of the reactor coolant piping at H. B. Robinson Unit No. 2 using ari updated version of the WESTDYN code which uses an absolute sum technique.

On the basis that the reanalysis was conducted with both an approved*

method and acceptable results, we consider this matter resolved, and are closing out this action item. A copy of our safety evaluation addressing this matter is enclosed.

Enclosure:

Safety Evaluation cc: w/enclosure See next page Sincerely,

a. o,t,~

A.

chwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

).

e Mr. J. A. Jones Carolina Power and Light Company cc:

G. F. trowbridge~ Esquire Shaw, Pittman, Potts and Trowbridge

,aoo M Street, N~w.

20036

. Harts vi 11 e Memorial Library Home and Fifth Avenues

  • Hartsville, South Carolina 29550 John F. Wolf, Esquire, Chairman 3409 Shepherd Street Chevy Chase, Maryland *20015 Dr. A. Dixon Callihan Union Carbide Corporation P. O. Box Y Oak Ridge, Tennessee 37830 Dr. Richard F. Cole

.Atomic Safety and Licensing Board U. s.*Nuclear Regulatory Commission Washington, D. C.

20555 United States Nuclear Regulatory Commission Office of Inspection and Enforcement Region II 101 Marietta Street Suite 3100 Atlanta, Georgia 30303

_UNITED STATES.

NUCLEAR REGULATORY COMMISSIOl'J

SAFETY EVALUATION BY THE OFFICE OF NUCL~AR REACTOR REGULATION RELATED TO THE SEISMIC REANALYSIS OF REACTOR COOLANT PIPING CAROLINA POWER AND LIGHT COMPANY.

H.B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO~ 2 DOCKET NO. 50-261 In response to IE *su11etin 79-07, Carolina Power and Light Company identified that the reactor cool ant piping dynamic ana1ysi s performed by Westinghouse

.using the WE_STDYN code_ in 1970 was performed using algebraic summation for intramoda1 responses.

Westinghouse has reanalyzed the reactor coolant

. system piping using an updated ~ersion of WESTDYN which combines intramodal responses by the absolute sum method.

The results of this reanalysis show only minor changes in the pipe stresses, support loads and equipment nozzle loads.

All stresses are within the allowable limits specified in the H. B. Robinson Unit 2 Final Safety Analysis Report~

The updated WESTDYN code used in the reanalysis of the primary coolant piping is contained in WCAP 8252, Re.vision 1, May 1977, 11Documentation of Selected Westinghouse Structural Analysis Computer Codes 11 This code has beeri reviewed by the staff and is acceptable.

We find the reanalysis of the reactor coolant piping system has been performed with an acceptable computer code and the results of the reanalysis meet the applicable requirements specified in the app1icant 1 s safety analysis

' report.

Date:

MAY 2 2 1979

e*.. UNITE~*STA,.,.ES e : ~. l_;~~---~>(J.f?..,):)_'..

  • .NUCLEAR REGULATORY COMMISSION.

Docket Nos*. 50-295 and 50-304 Mr.* Cordell Reed Assistant Vice President Coo.11onwealth Edison Company Chicago, Illinois 60690

Dear Mr. Reed:

WASHINGTON, 0. C. 20555

  • Ju1y.5, 1979 This is in**response to your letters wherein yew reported the results of a recent s~ismic reanalysis of the reactor coolant pip1ng at Zion Station Units l and 2 using an updated version of the WESTDYN code which uses an absolute sum technique.

On the basis that the reanalysis was conducted with both an approved method and acceptable results, we consider this matter resolved, and are c1osi~g out this action item.

A copy of our safety evaluation addressing this matter is enclo~ed.

Enclosure:

Safety Evaluation cc: w/enclasure See next page Sincere1,¥,

/}

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0tt1..c~/t-~

~- Sc wencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

e

~:". *Cordell Reed

  • Cor.rnonwea 1th* Edi son Company cc:

Robert J. Vollen, Esquire 109 North Dearborn Street Chic~go, Illinois 60602 Dr. Cecil Lue-Hing Director of Research and Development Metropolitan Sanitary District of Greater Chicago 100 East Erie Street Chicago, Illinois 60611 Zion-8en'ton Public Library District 2600 Emmaus*Avenue Zion, Illinois 60099 Mr. Phillip P. Steptoe Isham, Lincoln and Beale Counselors at Law One First National Plaza 42nd Floor Chicago, Illinois 60603 Susa~ N. Sekuler, Esquire Assistant Attorney General Environ~ental Control Division 188 West Rtndolph Street, Suite 2315 Chicago,. Illinois 60601 July 5~ 1979

L__ ___ _

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.. UNf~EDSTAT~S *..*... *.* *.

. NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C, 20555 SAFETY EVALUATION BY THE ciFFICE OF NUCLtAR REACTOR REGULATION RELATED TO THE SEISMIC RtANALYSIS OF REACTOR COOLANT PIPING COMMONWEALTH EDISON COMPANY ZION STATION UNITS NOS. 1 AND 2 DOCKET NOS. ~0-295 AND 50-304 In response to* IE Bulletin 79-07, Commonwealth Edison Company identi-fied that the reactor coolant piping dynamic analysis performed by Westinghouse using the WESTDYN code in 1970 was perf6rmed using.

algebraic summation for intramodal responses.

Westinghouse has rean-alyzed the re~ctor coolant system pipirig using an updated version of WESTDYN which ~ombines intramodal responses by the absolute sum method.

The results of this reanalysis show only minor changes in the pipe stresses, support loads and equipment nozzle loads.

All stresses are within the allowable limits specified in the Zion Station Final Safety Analysis Report.

I The updated WESTDYN code used fo the reanalysis of the primary coolant piping is contained in \\.JCAP 8252, Revision l, May 1977, "Documentation of Selected Westinghouse Structural Analysis Computer Codes".

This code has been reviewed by. the staff and is acceptable.

We find the reanalysis of the reactor coolant piping system has been performed with an acceptable computer code and the results of the reanalysis me~t the applicable requirements specified in.the applicant 1s safety analysis report.

Date:

J~ly 5, 1979

UNITED STATES UJ *..

  • I tA...1.. '"--:...* :..-LL*.

NUCLEAR REGULATORY COMMISSION WASHINGTON, o. C. 20555

ket No. 50-220 Mr. Donald P. Oise Vice President - Engineering Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York. 13202

Dear Mr. Di se:

June l9, 1979 The Commission has evaluated your response to Inspection and Enforcement Bulletin 79-07 as it affects the Nine Mile Point Unit 1 Nuclear Power Plant.

Bulletin 79-07 requested the identification of system designs for which an incorrect piping design code was utilized.

By letters dated April 25, 1979 and June 12, 1979 you provided the documentation which identified the extent of the problem as well as the details of the reanalysis effort.

We have concluded that the requirements of Bulletin 79-07 have been adequately satisfied to. allow resumption of power operation. This is b~sed on the fact that the reanalysis was conducted using computer programs exploying accept-able seismic response combination techniques, and that the results of the rec;nalysis have been within the allowable limits.

Pl.ease note that a final

.finding of canpliance with the provisions of Bulletin 79-07 wil.l require canpletion of c001puter code verification.

By letter dated June 13, 1979 you canmi tted to providing the necessary i nfonnat ion for this effort *.

A copy of our evaluation addressing this matter is enclosed.*

Enclosure:

Evalua:ion cc n1/e;ic~ csLJ*re*:

1::e 1ex: ~cge Sincer_1:l ;,

I I

~--* 0:-/,;@

~o as A. Ip lite, Chief Op rating R actors Branch #3 Division o Operating Reactors

Mr. Donald P. Oise ct:

Eugene B. Thomas, Jr., Esquire LeBoeuf, Lar:ib, Lei by & MacRae 1757 N Street, N. W.

'r.'ashi ngton, D. C.* 20036 Anthony Z, Reisman Natural Resources Defense Council 917 15th Street, N. W.

Washington, D. C. -20005 Oswego County Dffice Building 46 E. Bridge Street Oswego, New York 13126 2 -

NINE MILE POINT UNIT 1 NUCLEAR POWER PLANT REVIEW OF PIPING REANALYSIS PER I&E BULLETIN 79-07 SAFETY EVALUATION REPORT INTRODUCTION By letters dated April 25, 1979 and June 12, 1979 the licensee (Niagara Mohawk Power Corporation) responded to Inspection and Enforcement (I&E)Bulletin 79-07 (dated April 14, 1979) for Nine Mile Point Unit 1 Nuclear Power Plant.

Seven systems located inside containment were identified as Gaving been designed using an incorrect pi p.i ng comp*uter code.

These systems hav':: been reanalyzed with a code which uses *a~ceptab le sei srni c response combi nat fon techniques and the new results showed that*piping str.esses remained within code allowable (ASA B31.l, 1955) and that piping supports, penetrations, nozzles and equipment loadings remained within original design conditions.

DISCUSS ION The affected systems identified by the licensee are:

A.

Reactor Recirculation B.

Shutdown.Cooling C.

Emergency Condenser Returns D.

Reactor Cleanup E.

Reactor Drain F.

Reactor Feedwater G*.

Control Rod.Drive NMPC has stated that the method of algebraic (considering signs) summation of the codirectional spatial* components was used for the original analysis of safety related piping systems as identified above at Nine Mile Point Unit 1 during 1972.

The computer code used was the 1972 Version of ADLPIPE Code.

The licensee has indicated that the following computer codes were used in the reanalysis of these systems.

ADLPIPE - Teledyne Engineering Services (TES)

TMRSAP - Teledyne Engineering Services (TES)

TES has stated that it used a version of ADLPIPE, dated 1973, thru the CDC Cybernet system.

~his code was used to generate responses to the individual seis~ic excitation c::,.7,;:,oner:ts.

The r.iodal response due to each excitation were ccxr.::ined (internally)

~" 0 c-:::,c-c-,..,,:,~ho."*.. h"'c:::e rec:::...,onses *.ier"' ""'ni:>'1 C"*""""~npr: ;,..,....,,..,...=1'" (-=nu-llv)

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=.~so by :he SRSS method.

Since the combinc:ion of :he res;::onses ':o the earth-

'..:cke CO,'":lponents is performed externally, i~S hc.s not ::;ee;. re*:;t..es-:e: to submit
~e For:ran listing of this version of ADL?I?E; they have, however, described
~e'.r ~e:hcd of co~binat~on and haye co~mi::ed :o so1ve a se: c~ ~~= generated
e-,c:-::-:crk ;:::i*ct::1e~s c.S ;:an. of :he cede or :echnic;ue *,er"f~c.:-c.~cn effort.

'r/e

~~1d :;:~s co~~~~~ent acceptable.

- 2 TES has al so stated that it used the code TMRSAP.

Thi.s is a pro;::,rietary version of the code SAP IV which uses absolute summation for the intra~odal ccr.:pcne~ts and SRSS for the intennodal ctlllponents.

No modifications were liiade to the**

  • . dynar.iic response portion of the code.

In addition, a set of f'iKC benchmark probleliis are presently being solved by:TES as part of the code verification effort.

NMPC has indicated that the company, in conjunction with an Ir.spection and Enforce-ment ins*pector, verified that the drawings accurately depicted the as-built condition of the plant.

The licensee has s*tated that with the exception of the cleanup sys~em discharge, all of the piping sys.terns identified above were reanal;'zed in April 19i9 using the updated ADLPIPE cc:xnputer code.

The results of these reanalyses show that

  • stresses of all piping remain within code allowable ranges and the strength of structural attachments is within the design conditions.

The.licensee has also stated that in late 1978 through early 1979, seismic analysis was perfonned for the re-route of the Cleanup System discharge to the.

Feedwater Syste~ by Teledyne Engineering Services using the TMKSAP Code.

The TMRSAP Code calculated the earthquake component effects simultaneously.

For each mode codi rect i onal components were.added absolutely after which mod a 1 values were cc:xnbined by the Square Root of the Sum of Squares method.

This code will be verified as part of the NRC Code verification effort.

The licensee has indicated that the relatively minor increased stress levels at certain points in the reanalyzed piping lines has no effect on their pipe break analysis.

The NMPC's response to *I&E Bulletin 79-02 states that the reanalysis has resulted in stress increases at ten restraints.

These restraints have been reanalyzed and found acceptable to support the increased loads.

Since the restraints where loads have increased are not supported by concrete anchor bol-::s, the reanalysis has no effect on the testing.and inspection perfonned under I&E Bulletin 79-02.

In response to I&E Bulletin 79-04 the licensee has indicated that no VELAN swing check valves are installed in Seismic Category I piping syster.is at Nine Mile Point Unit 1.

EV.!.LUATION The reanalysis method used was a lumped mass response spectre modal analysis.

The i.ic.~*ority of piping systems reanlayzed used Regulatory Guide 1.92 for the ccrnbin-

~:::,s of :-:rndal* responses and spatial cooiponents.

This dynas~c a::alysis procedure

acce;:::c.:1e.

The procedure used for the clean-up syster:i d-:schar-ge to the e~*:*..,a:er Syste:;1 is al so acceptable.

e

  • - 3* - '

For the seven systems of concern the licensee utilized the cc:tii;::,u:er codes as is, discussed above.

In addition to satisfying the code verification req?Jirer.ients, -the licensee has also agreed to provide the NRCtwo problems for confimatory analys*is Th'ese confi matory problems will be solved independently by consultants to the NRC at.Brookhaven National Laboratory.

The models submitted for these p1ping problems wil 1 be confirmed* by the licensee as corresponding to the "as-built 11 condition.

We find these commitments acceptable.

We find the licensee's responses concerning I&E Bulle:in i9-02 and 79-04 accept-able for the piping systems reanalyzed.

The reanalys1s has no e~fect on Nine Mile ~oint Unit l pipe break criteria for the piping syste~s inv61ved.

CONCLUSION Based on the discussion and evaluation presented above, we conclude that the requirements set forth in l&E Bulletin 79-07 are adequately satisfied to allow resumption of power operation provided that the infonna~ion necessary for code verification be provided within 30 days of unit startup.

1

    • ~

A

  • . UNITED ST ATES A

. NU~AR REGULATORY COMM1SSION1JIIIJ WASHINGTON, 0. C. 20555 June 22, 1979 Docket Nos. 50-315 and 50-316 Mr~ John E. Dolan, Vice Ptesident Indiana and Michigan Electric Company I nd:i ana and Michigan Power Company Post Office Box 18 Sowling Green Station New York, New York 10004

Dear Mr. Dolan:

This letter is submitted in response to your lette~s dated April 24, May 4, and June 20, 1979 which provided documentation en :he results of the seismic piping stress reanalysis for the D. C. Cook Power Station, Unit Nos. 1 and 2 in connection with IE Bulletin 79-07.

\\.ie have reviewed your reanalysis techniques and the results of the reanalysis and find* them acceptable.

In subsequent discussions,;,with Westinghouse on canputer code verification; Westinghouse has agreed to solve a set of

. benchmark problems using WESTDYN and to provide the NRC a problem for confir;:iatory analysis.

For ccxnputer code verificatior., we firyd this acceptab 1 e.

A copy of our evaluation addressing this matter is enclosed.

ncl osure:

c.fety ::valuation Report cc:

~:e~c1osure See..,ext page Sincerely,

(}

A. Schwencer, Chief Operating Reactors Branch fl Division of Cperating Reactors

L_

e 1*'.r-. John Del an

  • . * "d,* "n * -,,d M; ;- *
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.n. ec~nc ompany Indiana and :~ichigan Power Company.

cc:

Mr. Rcb.ert W. Jurgensen Chief Nutlear Engineer A~erican Electric Power Sefvice Cor~oration 2 8 roadway New York, New York 10004 Gerald Charnoff, Esquire Shaw, Pittman, Pot~s and Trowbridge 1800 M Street, N.W.

i~ashington~ D *. C.

20036 Citizens for a Better Environment 59 East Van Sutan Street Chicago, Illinois 60605

~aude Reston ?alenske Memorial Li brar-y 500 Market Street St. Joseph, Michigan. 49085 Mr. D. Sha11er, Plant Manager Donaid C. Cook Nuclear Plant

?

  • 0. Bex 458 Bridgman, Michigan 49106 Mr. R. Masse Donald C. Cook Nuclear Plant.

P. O. Box 458 Bridgman, Michigan 49106

- 2.:.

J~ne 2i, 19i9

. DONALD. COOK NUClE.l;R PLA~T. UN ITS l,e 2.

REVIEW OF PIPING REANALYSIS PER I&E ~ULLETIN 79-07 SAFETY EVALUATION REPORT

. ENGINEERING BRANCH DIVISION OF OPERATING REACTORS I~T~~DUCTION/SACKSR00ND

-In :heir April 24, 1979, response to I&E Bulletin 79-07,,!;Jiierican Electric Power Service Coiporation (AEP) identified 24 lines that had originally been analyzed using an earlier version of the WESTDYN computer progra~ that incorporated an algebraic sum of intramodal responses to seismic loadin;s. However, 19 out of these 24 1 i nes had _subsequently been reanalyzed using,.i:iso 1 ute surrrnat ion or

._SRSS method *
  • Sy 1etter dated May 4, 1979, the licensee (AE?) reported that a further check of their records showed that 23 of the 24 lines had previ.o:.:sly been reanalyzed.

In response to NRC questions discussed during telephone conversations, AE?

supplied.supplemental information on this subject in a letter dated June 20, 1979.

DISCUSSION The licensee identified the only line that had not been seisiilically reanalyzed was the 14 11 pressurizer suroe 1ine. *This line has the same~geometry on both units, but in opposite hand; th~refore,"'onlr one stress analysis was necessa.ry.

AE? has stated that a reanalysis of the surge line in the 11 2.s-built" condition has been completed*and the results show al1 piping stresses remain below their allowable values, as specified in the FSAR's.

Additionally, the pip~ supports

~Eet FSAR criteria and the nozzle loads have been found acceptable.

This piping run was reanalyzed by Westinghouse using their current.version of WESTDYN which combines the intramodal responses by absolute sumnation.

I&£Bulletin 79-02 was not addressed at this time.

The 11ce~see has stated that the reanalysis has no effect on pipe bteak criteria si~:e jreak lot~tions were net postulated based on stress levels.

~*,*.:. :_ _'.:.~: :*r~

~~ ~ea~!;!s1s ~ech~ique e~:lcye~ was a 1u~;e~ ~ass res;:~se s:ec~ra ~cdal

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~table since they of the pressLlrize;

  • .UNITED STATES NUCLEAR REGU~TORY COMMISSION WASHINGTON, 0. C. 20555 Docket Nos.*50~325 and 50-324 Mr. J. A. Jones Executive Vice President C*rolina Power & Light Company 336 Fayetteville Street Raleigh, North Carolina 27602

Dear Mr. Jones:

June 7, 1979 This letter is submitted in response to your letters dated April 24, 1979, May 15, May 21, May 22, May 29, 1979 and June 4, 1979 which provided docu-mentation of the results of seismic piping stress reanalysis for Brunswick Steam Electric Plant, Units 1 and 2 in connection with IE Bulletin 79-07.

Your submittals were also in response to the staff 1s reques:s for infonnation set forth in our meetings of May 16, May 21, May 30, 1.979 and June 4, 1979.

On the basis that the reanalysis was conducted using computer programs

  • employing acceptable seismic response combination tech~iques, and*accept-able results were obtained from the reanalyses, we conclude that the requirements set forth in IE Bulletin 79-07 will be adequately satisfied to allow resumption of power operation on completion of the modifications identified in our meeting of June 4, 1979.

However, as-built verification of all lines and supports and completion of code verifications will be required before your response to IE Bulletin 79-07 is con~idered finally complete.

A copy of our evaluation addressing this matter.is enclosed.

ncl osure:

E:valuation cc,.,;/enclosure:

See next pa;e Sincerely, t;;t

.,,// -~' -~'7.-t,J~ _,

. ~

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Thomas':..:II ppc 1 i to, Chief Operating Reactors Branch #3 Division of Opera:ing Reactors

e Carolina Power & L~ght Co~pany

  • cc:

Richard E. Jones, Esquire Carolina Power & Light Company 336 Fayetteville Street Raleigh, North Carolina 27602.

George F. Trowbridge, Esquire Shaw, Pittman, Potts & Trowbridge 1800 M Street, NW Washington, D. C.

20036 John J.. Burney~ Jr., Esquire Burney,** Burney~ Sperry & Barefoot 110 North*Fifth Avenue Wilmington, North Carolina 28401 Southport - Brunswick County Library 109 W. Moore Street Southport, North Carolina 28461

.e

e e

SRUNSWICK STEAM.ELECTRIC PLANT, UNITS 1 A~D 2 REVIEW OF PIPING REANALYSIS PER I&E BULLETIN 79-07 SAFETY EVALUATION REPORT INTRODUCTION/BACKGROUND

.In their April 24, 1979 response to I&E Bulletin 79-07 Carolina Power ana

~ight Company (C?&L) stated that the recirculation and main stea~ piping nad been analyzea by GE using a computer coae that corncinea directional seismic responses by algebraic summation.

All other safety re1atea piping was analyzed by UE&C using a computer code that combinec airectional seismic responses by algebraic summation.

CP&L has supplied su,:;plementai information on this_ subject at.*meetings with the NRC staff and in :etters aated May 15, 21, 22, 29, and June.4, 1979.

DISCUSSION CP&L has stated that a reanalysis of all affected piping in tne "as-cuilt 11 condition 'l'lill be completed with the results showing all piping stresses remaining below.their allowable values, as specified in tne i3S~P FSAR cy July 21, 1979.

Aaditionally, all loads on attached equipment (nozzle loads) wjll be acceptaole.

Upon completion of moaifications to certain pipe supports which were determined to oe originally underdesigned, all pipe supports attached to safety related piping or equipment in the plant will ce "operable 11 and within FSAR criteria. These modifications will be completea prior to return to power operation.

The recirculation and mainsteam lines were reanalyzed oy GE using PISYS.

The

. responses from two directions, the most disaavantageous comoination of one horizontal with.the vertical, were combined oy the aosolute sum and the results wer~ within FSAR allowable.

Refer to Evaluation Section for reanalyses done by ).JE&C.

The licensee 1s response to IE bulletin 79-04 states that no VELAN swing checK valves are in any of the affected piping.

Further, lE Bulletin 79-02 was addressed when.piping support modi fi cations were fauna to oe necessary.

The licensee has stated that the reanalysis has no effect on pipe oreak criteria since the postulated oreak was analyzed to occur a~ any point on the p-ipe, inside or outside containment.

EYALUATIOt~

he reanaiysis tecnn*,que em~loyea was a lumpea ;nass res;;c1se spec:ra moaai analysis.

This cyr.arnic analysis procecure is an acce;:,t.ac-*,e :-:-.e:::ca.

ihe

.:s.:::lute cc;;;.:>inat~on of responses in t*,.,,o airections is aiso a::ep:aole to

e e Once the support modifications are complete, the affected piping stresses, at7.ached equipment loads, and support designs will all° be in accoraance with FSAR criteria and acceptable to the staff,.

UE&C has reanalyzec category l (pressure oounaary) and category Z (other safety relatec) lines using Squ~re Root of Sum of Squares (SRSS) load comoination insteaa of a1georaic summation.

ihis analysis employee a conservative factor cf 2 to convert from OBE to DBE.

when the use of SRSS methodology for a lD analysis was questioned because it die not conform to FSAR commitment, the licensee appliea a conservative factor of 1.38 to the SRSS results to convert to absolute sums, which is an acceptable load combination methoa wi:~ a ZD analysis.

In such cases, credit for.conservatism in the OBE/DBE r~lationship was taken (a more realistic.*factor of 1.2 was *used instead of 2).

When this exercise was completea, one of the first 39 reanalyzed lines was found to exceed. total allowable stress by 2~, but was still less than 0.9 Sy as permitted by the

  • FsAR and was found acceptaole.

For the remaining 411 unreanalyzed lines, SRSS stresses were estimated from the algeoraic summation stresses by applying a factor of 1.5. The SRSS results were th~n converted to aosolute sums for use with tne 2D analysis by applying a factor of 1.38.

Credit was again taken for the conservatism in the OBE/DBE relationship.

When this exercise was completed, 39 aaditional lines were found suspect.

SRSS stresses were computed for these lines which eliminated the factor of 1.5.

However, several of these lin~s still 9ave stresses in excess of code allowable *. io resolve this proolem, the licensee recomputed the total stress using coincident point values instead of maximum values. ihis method is more realistic ana is acceptable to the staff. ihe new total stresses were all within code allowable.

1ne above procedure which the licensee took in completing the p1p1ng seismic stress reanalysis under IE Bulletin 79-07 is acceptaole to the staff.

We find the licensee 1s responses concerning I&E Bulletin 79-02 anu 79-04.

acceptable.

The reanalysis has no effect on BSEP pipe break criteria committed to in the FSAR.

The _staff still has some concern as to whether the reanalyses effort reflects the true as-built conditions in the plant.

However, CP&L has completed a walk-down of the piping and supports insiae the dry well to verify that the as-b~ilt condition has indeea Deen utilized. Aaditional1y, the licensee has committee to a field verification of all lines on both u~its DY June 15, i9i9.

  • ,.;e fine this commitment acceptable.

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~e 1ice~see 1as i1cicatea :na: the followin~ computer ;ro5rams ~ere usea in reana. ;_,. s ~ s : i :r,, s p I ant:

  • P!S'r'S - General Electric Company ACL?IPE United Engineers and Constructors GE has statec that the code PIS'r'S combines the responses cue to seismic
  • i:iuiti-axial excitation by absolute summation; the modal res;::onse.due to ea.en *excitation are combined by methods as specified in Regulatory Guide i.92.

A.Fortran listing of the dynamic response calculations section of PiS'r'S has :::een suomitted by GE ana these statements have :een certifiea and ccnfirmea~

GE is also presently solving a set of NRC genera.tea benchmark problems as part of the co~e verification effort.

UEC has i ndi ca ted that the code ADLP I P-1 I combines the,-esponses due to seismic excitation by.the SRSS method when used with tr,e response spectra technique.

ihis has Deen confirmed Dy examining the coae listing and Dy verifying the code Dy solving a set of benchmark problems.

In acdition to satisfying the code verification requirements, the licensee has also agreed to provide the NRC two problems for confirmatory analysis.

These conf i ri7ia 'tor.y prob 1 ems wi 11 De so 1 ved i ncepenaently Dy consul tan ts to the ~~RC at 5rookhaven National Laboratory.

The models suomitted for these piping problems wi11 oe confirmed by the licensee as corresponding to the "as-ouilt" condition.

We find these commitments acceptaole.

CONCLiJSiON Basea on the discussion and evaluation presented aoove, we concluae that the requirements set forth in I&E Bulletin 79-07 are aaequately sa:isfied to allow resumption of operation upon completion of the moaifications iaentifiea in our meeting of June 4, 1979.

i

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