IA-85-301, Forwards Draft Temporary Instruction Providing Guidance for Inspecting & Evaluating Licensee & CP Holder Responses to IE Bulletin 79-14, Seismic Analysis for As-Built.... Comments Requested by 790723.Supporting Documentation Encl
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4-g NUCLEAR REGULATORY COMMISSION R WASHINGTON, D. C. 20556 4o oooo / JUL 181979 MEMORANDUM FOR: B. H. Grier, Director, Region I J. P. O'Reilly, Director, Region II J. G. Keppler, Director Region III K. V. Seyfrit, Director Region IV R. H. Engelken, Director, Region V H. D. Thornburg, Director, Division of Reactor Construction Inspection, IE
, D. G. Eisenhut, Acting Director of Operating 9eactors, NRR D. B. Vassal.lo, Acting Director of Project Management, NRR FROM: Edward Jordan, Assistant Director for Technical P'rograms
SUBJECT:
TEMPORARY INSTRUCTION RE IE BULLETIN 79-14 The enclosed draft temporary instruction (TI) provides guidance for those persons inspecting and evaluating responses of licensees and construction permit holders to IE Bulletin 79-14, " Seismic Analyses for As-built Safety-related Piping Systems." Comments received by COB on July 23, 1979 will be considered for
' incorporation in the TI.
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efdsiard . Jordan, Assisjt t Director for echnical Programs
Enclosure:
As Stated cc: N. C. Moseley R. W. Woodruff H. Wong C. J. DeBevec A W. Russell, NRR j l V. Noonan, NRR L. Shao, NRR
Contact:
H. Wong, TP 492-8180 h Ik g 9 850703 HERRNAN85'301 PDR
- 7/13/79 TI 2515/
Issue Date: INSPECTION REQUIREMENTS FOR'IEB 79-14 I. OBJECTIVE The objective of this temporary instruction is to provide guidance for IE inspection and review of licensees' actions and written responses to IE Bulletin 79-14. Bulletin 79-14 requests that licensees assure that seismic analyses of safety-related piping systems accurately reflect the as-built configuration of the plant. II. BACKGROUND Recently, two issues were identified which are related 1TI the validity of seismic analyses. These are the analytical technique for combining 7 seismic loads and the validity of input information for seismic analyses. IE Bulletins 79-07 (combining seismic loads), 79-02 (as-built condition of pipe supports) and 79-04 (actual valve weights) address these issues. As a result of issuing IE Bulletin 79-07 and show cause orders to Came= M +4e licensees, the concern regarding the technique for combining seismic loads was essentially resolved. IE Bulletins 79-02 and 79-04, however, have led to discovery of some failures to conform to design documents which are outside the scope of these bulletins and could have an adverse effect on the validity of the seismic analyses. Based on this fact, IE and NRR concluded that it is necessary to request licensees to verify that other seismic analysis input information is correct by comparision of this input with the physical facility as constructed. IE Bulletin 79-14 was issued for this purpose. The bulletin requires that licensees establish an adhoc inspection program scheduled so that the required inspections are com-pleted within 120 days. Further, the bulletin requires that licensees resolve specific nonconformances by either making changes to the system such that it conforms to design or by correcting seismic analysis to demonstrate conformance of the as-built system to design criteria. It also requires that licensees take action to correct administrative problems which could allow this problem to recur. III. BULLETIN REQUIREMENTS To comply with the requests in IE Bulletin 79-14, it will be necessary for licensees to do the following:
- 1. Identify Inspection Elements The licensee must himself or through his contractors or consultants:
(a) identify the piping system parameters which were input into the seismic analyses, (b) identify specifically the design documents from which values of the parameters were obtained for the seismic analyses and (c) establish acceptance criteria which as-built values of these w -- . . . -- .
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s parameters must meet. System parameters which are important include but may not be limited to system geometry; locations and orientations of' anchor points and restraints; masses; locations of centers of gravity; sizes and cross sections of piping, supports and restraints; restraint clearances; and material properties. To competently comply with Item 1 in the bulletin the licensee must assure that the persons identifying these inspection elements are sufficiently conversant with the seismic analysis documents to identify all significant inputs and their sources. Inspection elements must be identified for those safety-related piping systems addressed in the bulletin. The licensee must then report to the regional office in accordance with Item 1 of the bulletin. w
- 2. Inspect Part of the Accessible Piping For each system selected by the li o nsee in accordance with Item 2 of the bulletin, the licensee is expected to verify by physical inspection, to the extent practicable, that the inspection elements meet th'e acceptance criteria. Where physical inspection can only be accomplished by removal of thermal insulation, the licensee is expected to do so. Where physical inspection is not practicable e.g. for valve weights and materials of construction, the licensee is expected to verify conformance by inspection of quality assurance records. If a nonconformance is found, the licensee is expected in accordance with Item 4 of the bulletin to perform an evaluation of the significance of the nonconformance as rapidly as possible to determine whether or not the operability of the system might be jeopardized during a design basis earthquake as defined in the regulations. This evaluation is expected to be in two phases, an initial engineering judgment (within 2 days) followed by an analytical engineering evaluation (within 30 days). Where either phase of evaluation shows that system operability is in jeopardy, the licensee is expected to meet the technical specification action state-ment and complete the inspections required by Items 2 and 3 of the bulletin as soon as possible. The licensee must report the results of these inspections in accordance with the requirements for content and schedule as given in Items 2 and 3 of the bulletin.
- 3. Inspect Remaining Piping The licensee is expected to inspect, as in Item 2 above, the remaining safety-related piping systems which were seismically analyzed and to report the results in accordance with the requirements for content and schedule as given in Item 3 of the bulletin.
4A. Evaluate Nonconformances With regard to Item 4A of the bulletin, the licensee is expected to include in the initial engineering judgment his justification for continued reactor operation. For the analytical engineering evaluation, the licensee is expected to perform the evaluation by using the same analytical technique used in the seismic analysis or by an alternate, less complex technique provided that the licensee can show that it is conservative.
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If either part of the evaluation shows that the system may not perform its intended functi'on during a design basis earthquake, the licensee must promptly comply with applicable action statements and reporting
. requirements in the technical specifications.
- 48. Submit Nonconformance Evaluations The licensee is expected to submit evaluations of all nonconformaces and, where the licensee concludes that the seismic analysis may not be conservative, submit schedules for reanalysis in accordance with i Item 4B of the bulletir or correct the nonconformances.
4C. Correct Nonconformances If the licensee elects to correct nonconformances, the licensee is expected to submit schedules and work descriptions in accordance with Item 4C of the bulletin.
- 4D. Improve Quality Assurance:
If nonconformances are identified, the licensee is expected to evaluate and improve quality assurance procedures to assure that future modifications are handled efficiently. In accordance with Item 40 of the bulletin, the licensee is expected to revise design documents and seismic analyses in a timely manner. IV. REQUIREMENTS FOR IE INSPECTION Evaluation of licensees' actions will consist of inspections and reviews of written responses in the field to assure that licensees responded to the bulletin in a timely and competent manner and reviews at headquarters to assure that licensees' actions are appropriate. For each site, the inspector will inspect the following: , 1. Development of Inspection Elements Review the organization and the qualifications of the persons who. developed the inspection elements and interview one of those persons. Determine that pertinent parameters and valves were identified by Item I in the bulletin. Inspect some seismic analysis source documents used for this purpose. Determine that acceptance criteria ! were developed in some rational way. Inspect some of the documentation of inspection elements and acceptance criteria which was prepared for use by personnel inspecting the piping systems for the licensee. Obtain assistance from the Vendor Inspection Branch if appropriate. 0
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- 2. Licensees' Inspection of Accessible Piping Observe in part the physical inspections of accessible piping systems performed by licensees in accordance with Item 2 of the bulletin.
Review licensees' reports to determine that they accurately reflect the work done. Independently inspect a segment of a piping system which the licensee has completed. For that segme.7t, inspect each inspection element. -- 3. Licensees' Inspection of Normally Inaccessible Piping In accordance with Item 3 of the bulletin, do the work described in Item 2 above. 4A. Nonconformance Evaluations Where nonconformances are identified, determine that evaluations are initiated as soon as is reasonably possible and are completed in accordance with Item 4A of the bulletin and Section III, Items 2 and 4A, above. Assure that action is taken promptly with regard to action statements in technical specifications. 4B. Subraittal of Nonconformance Evaluations Determine that licensees promptly submit all completed nonconformance evaluations to NRC per the distribution given in the bulletin. Also, determine that licensees promptly submit schedules as required by Item 4B in the bulletin where reanalysis is indicated by licensees. 4C. Correction of Nonconformances Where licensees elect to co rect significant nonconformances, determine that schedules and reports required in Item 4C of the bulletin are submitted promptly.
- 40. Imorovement of Quality Assurance For sites where nonconformances are identified, assure that necessary improvements to quality assurance procedures related to design changes due to modifications or maintenance are completed within 120 days of the date of the bulletin. Also assure that design documents and seismic analyses are revised in a timely manner as required by Item 4D of the bulletin and in accordance with Section III, Item 4D above.
V. Reporting Requirements The results of inspections required by Section IV above, shall be included in the usual inspection report. The regions shall transmit a copy of pertinent portions of inspection reports describing this effort to R. W. Woodruff, TP, ROI, IE and to Sac- 0... , , NRR. 5, 6. Hos ty,.4
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VI EVALUATION LICENSEE'S REP' ORTS i Reports submitted by licensees in accordance with the requirements of the bulletin will be evaluated at headquarters by a task group with the following membership: Q & La(CvW'.% NRR ( SV SWWY #* NRR ( & SjsTcwAC vew M h M RCI, IE (for construction only) C. J. DeBevec, TP, ROI, IE (for BWRs only) R. W. Woodruff, TP, ROI, IE (for PWRs only) RI (for RI reactors only) , RII (,for RII reactors only) C n., . . . . , w RIII # '. . 4..(for RIV reactors only)n RIV ! ~(for'RV' reactors only)
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This task group will prepare evaluations of the reports submitted for each " operating facility. Reports submitted by each holder of a construction permit will be evaluated, by the task group in.<.onjunction with the licensing review which leads to issuance of the operating license. VII. EXPIRATION This TI shall expire on January 2,1980. ,
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VIII. IE HEADQUARTERS CONTACTS H. J. Wong, R. W. Woodruff, E. L. Jordan (492-8180) IX. MMULE TRAINING SYSTEM INPUT (766 DATA) For module tracking system input, record the actual inspection effort against Module No. . Record task group review against. . t l l -
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Dear Mr. Fain:
j hp MdQ uu-This is in response to your letter,of November 16 in which g"g.s$igF* you raised questions about my having commented in a letter to my colleagues on calculations related to the seismic design of five nuclear power plants. We have checked with ~ the NRC to determine whether they continue to believe the calculations to have been-in error. The attached correspondence relates to our inquiry. The statement in the letter to my colleagues derived from . concern that a wrong calculational technique had been used and that a considerable time period had elapsed before the ' m' take was discovered. y,,, y 5;3&igtemenw_ MEW Msh%nsineww w - mne6irFCaet%s&E@ _
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C. .:.'.A.'".&, .*.". .*.e*." 3 Nove=ber 16, 1979 i Honorable Morris K. Udall, Chairman Co=mittee on Interior and Insular Affairs 1
' House of. Representatives Wz Sgton, DC 20512 .
Dea ' cirman: I have nad occasion recently to read your letter of November 9 to your colleagues on the subject of the proposed Markey Amendment to the bill authorizing funds for the Nuclear Regulatory Commission together with the detailed . statement enclosed with 'that_Jetter. On page 2 of your detailed statecent, in the four*h paragraph wherein you discuss "two precursor events" which raise questions in your mind about nuclear safety, the following sentence is included- 'O ? On March 19, 1979, Mr. William F. Allen, Jr., Chairman and Chief Exe.cutive Officer of Stone & Webster Engineering Corporation, testified before your Subcommittee on Energy and the Environment concerning this catter. Notwith-standing the characterization by the Nuclear Regulatory Commission of the seismic calculations in question as " erroneous," Mr. Allen pointed out to the Subcommittee that the#d esien was in fact adeouate*and that the systems involved Nould not lose their ability to function *as intended were a costu-lated sei wie evant to eccur. L ,.,__ _ _ -_
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__ I'n an October 25, 1979, letter to Senator Hart in response to questions
,W m propounded by the Senator, the Nuclear Regulatory Commission has again N" addressed this issue and I encluse for your information a copy of Question No. fa and the response theteto by the NRC. As you will note, the Co= mission g-has now concluded, based on a survey of other plants and the design proce-dures employed by other organizations, that the seismic calculations
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[ performed by Stone & Webster on the five plants in question were in accord with the state of the art of seismic analysis at the time. I hope that this information will be of assistance in your further consid-eracion of these issues. Very truly yours, 42'd' - - Gerald G. Fain Assistant Manager ' Washington Oparations Enclosure w w c.n-. v a k r.y %, fs4rL & (D k:y & <a ^ M%'*S M V " 'g Mt[e_tt . kh
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o ~ CUESTION 4: Please list each reactor which has been found since F. arch 13,
- including the five reactors which were the subject of the hearing, 'w have had an error in the seismic analyses of plant design. In your response, please include:
(a) whether the reactor was shutdown because of the error; (b) whether the shutdown was voluntary or by order; (c) the systems involved; . (d) whether the systems are safety related or non-safety ~ ,,
- related; and (e)' the resulting corrective measure if any. .
ANSWER: At the time of the original safety review of the plants in question, specific NRC (then AEC) guidance on acceptable methods for ccmbining seismic forces did not exist. *luclear industry practice to combine seismic forces for piping systems varied; some design organizations used algebraic sumation, others used square root sum _ofJbAst;ar.e@%adg'hassasad,.absol.u'4summationgethod . tic Ras ausaeveloped
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_ indusni}h g gg,.3 Myg,w w - =-- n => m A kke MQ m eGM.pnacceptable ~t' ay. 7n Dece ber 1974, when Regulatory Guide 1.92, " Combining Moda1' Responses and Spatial Ccmponents In Seismic Response Analysis", was issued providing specific, guidance on acceptable methods, the stafl did not review earlier plants to deter-mine if unacceptable methods had been, employed. Our efforts to reevaluate the seismic analyses and design of piping systems have - been directed at only safety related syst' ems since these are the systems which are of importance to assure the protection of the public health and safety. The
. list of the plants which have been found thus far to have used the algebraic sumation technique for the combination of codirectional responses to cultiple earthquake input components is contained in the accompanying table, including whether or not the reactor was shut down, whether the shut dcwn was voluntary or by order, a general description of the system involved, and any corrective measures. .
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- O SYSTEM CLASS Q-1 0-2 0-3 0 PIPE. RUNS ANALYZED 13 0F 20 21 OF 46 54 6F 54 -
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- MITHk.H CODE ALLOWABLE' VALUES 5 18 ' ll2
- STRESS GREATER THAN 3 3 12 ALLOWAdUE 0 70 0F 550 SUPPORTS EVALUATED 70% FOUND TO HAVE SOME LOADING COMPONENT GREATER THAN DESIGN SPECIFICATION -
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STRESS LEVELS IN PIPING, COMPONENTS AND SUPPORTS WILL SUBSTANTIALLY INCREASE UNKNOWN WHETilER DESIGN MARGINS WHICH. CURRENTLY EXIST WILL COMPENSATE ) FOR INCREASED STRESS LEVELS WHETHER EQUIPMENT WILL OPERATE UNDER. INCREASED LOADS i i , 1 4 l9 -
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ORDER 'T0 5110'l CAU5E l 0 NHY THE LICENSEE SHOULD NOT REANALYZE THE FACILITY , I . PIPING SYSTEMS FOR SEISMIC LOADS USING AN APPROPRIATE PIPING CODEJ, I j 0 WHY THE LICENSEE SHOULD NOT MAKE ANY NECESSARY I MODIFICATIONS FOLLOWING ANALYSESJ 0 NHY FACILITY OPERATION SHOULD NOT BE SUSPENDED PENDING SUCH REANALYSIS AND COMPLETION OF ANY RECUIRED MODIFICATIONS. O e 4 9 e O n! :! ,
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79 - 14 : AS BUILT I
.PR0bLEMS.
79-07 ALGEBRAIC MAINE YANKEE SUMMATION DBE ' 79-04 SYSTEMATIC VELAN VALVES EVALUATION PROGRAM _ - 79-02 UCS PETITION BASEPLATES FOR ANCHOR BOLTS
\ SEISMIC REANALYSIS
.~. s SEISMIC GAME PLAN OTHER ACTIVITIES TAP A-40
-CHANGING SEISMIC CRITERI A SSMRP
-COMPUTER PROGRAMS USED FOR ANALYSIS
. j l
m BAE PLATE RfXIBILITY #0 ANG0R BOLTS IEB79-02 NCH 8,1979) (JUNE 21,1979) REQUIREDACTION
' ACC0lNT FOR FIEXIBILITY IN ANCHOR BOLT LOADS VERIFY PROPER FACTOR OF SAFETY DESIGN REQUIREENTS FOR CYCLIC LOADS
. TESTING PROGRAM RESPONSESDUEJULY6,1979 PRELIMINARYRESULTS
' SOE PLANTS WITH EXTENSIVE INSTALLATI(N DEFECTS SOE PLANTS WITH FEW PRTLEMS
.} '
Scm GROSS INSTALLATION Pa LEMS (MISSING BOLTS, SLEEVES) SOME WITH MINOR INSTALLATICN PROBLEMS (MISSALIG4ENT, THREAD ENGAGEMNT) SO4E PLANTS HAVE REPORTED ENTIRE SUPPORT MISSING e
/
ELAN SWING CHECK VALE EIGHT IEB 79-04 (MARCH 30,1979) REQUIRED. ACTION IDENTIFYSYSEMSAFFECTED VERIFYCORRECTVALVEWEIGHTS RE-EVALUATEPIPINGSYSTEM
' IDENTIFY #1Y NECESSARY MODIFICATIONS -
IDENTIFY #1ALYTICAL TEONIQUE USED RESPWSESWEREDUEMAY1,1979 FINDINGS 48 0F 6 PLANTS USED VEL #1 VALVES
._ ,', . o PLANTS WHIG HAVE NOT YET RESPWDED ARE SHUTDOWN 5 PLNiTS REQUIRED iTDIFICATICH () HANGER OVERSTRESSED PLUS MISCELLANE0lJS HANGER ADJUSTENTS)
ADDITimAL VELAN VALVES
~. .. _ _ .
3 l ,N., STATUS OF REVIEW 0F 5 SHUTDOWN PLANTS SURRY 1 - PIPING ANALYSIS WILL BE COMPLETED 18 JULY 79 9 0F 49 PROBLEMS REQUIRE HARDWARE FIX EARLIEST START UP LATE AUGUST
- SURRY 2 -
SHUTDOWN FOR STEAM GENERATOR REPLACEMENT (EARLIEST START UP Nov. 79) EBASCO WILL DO SEISMIC REANALYSIS 3EAVER PROPOSAL FOR INTERIM OPERATION UNDER REVIEW VALLEY EARLIEST START UP 21 JULY 79 FEEDWATER LINE CRACKS
~
( .1 - BASE PLATE AND ANCHOR BOLT INSPECTIONS HARDWARE MODIFICATIONS DUE TO SEISMIC REANALYSIS FITZPATRICK - LICENSEE ESTIMATES THAT PROPOSAL FOR RESUMPTION OF OPERATION WILL BE SUBMITTED ABOUT 15 JULY 79 ALL PIPING ANALYSIS AND ALL SUPPORT ANALYSIS IN INACCESSABLE AREAS, INCLUDING ANY NECESSARY MODIFICATIONS, WILL BE COMPLETE PRIOR TO START UP MAINE YANKEE - APPROVAL FOR START UP ISSUED MAY 24, 1979 m 8 t g
= . . . ' . . . . . . . , , . . . > - . . . . . . , - .
ALEBRAIC SlMRTION i IEB B-07 (APRIL 14, D B) REQUIREDACTICN iDENTIFYAFFECTEDSYSTEMS REANALYZE USING APPROPRIATE CODE CCEE VERIFICATION RESPONSES NERE DUE APRIL 24, EB FINDINGS 25 OPERATING PLANTS REPORTED USE OF ALGEBRAIC SLM% TION CG)ES INVOLVED WERE , SHOCK 2 - STONE & %BSTER ADLPIPE ARTHUR D. LITTLE CO. ' WESTDYN WESTINGHOUSE
- DAPS GENERAL ELECTRIC PIPDYN 2 - FRMKLIN INSTITUTE NRR STAFF HAS COMPLETED TE REVIEW OF 16 PLANTS
-e
.I
.=*--=m..w ..-e..,,,u =. ,. *4w- = - . - = . m. +=, - e- e- e w
i OPEPATING REACTORS ESPONSE TO IEB 7M)7 IUIT CDE EMEKS BEAVERVAREY SHOCK 2 EXTENSIVE, ORDERS /D FITZPATRICK X MINE YANKEE SLRRY 1 SSRY 2 POINT BEACH 1 LIM ' COOUNG POINTBEACH2 \I y x brunswick 1 ADLPIPE&DPS EXTENSIVE x BRUNSWICK 2 y x INDIAN POINT 3 ADLPIPE&WESTDYN p SALEM 1 'PIPDYN EXTENSIVE x IroIANPorgr2 ADLPIPE 5 LINES COOPER SRVLINESONLY GItem 2 LINES x MILLSTONE 1 2 LINES , x MI asTa.E 2 6 LINES x NINE MILE POINT p LIMITED x COOK 1 WESTDYN 1LINE x COOK 2 WESTDYN P x ROBINSm 2 RCSONLY x TLRKEY POINT 3/4 x ZION 1 y ZION 2 Y Y X PILGRIM 1 DWS RCS & N IN STREAM ONLY X = PESOLVED .
.~ Pl. ANTS WITH SIGNIFICANT DIFFEENES IEMEN ORIGINAL IESIGN AND "AS BUILT" CmDITION OF PIPING SYSTEE Q DIFFERENCE REMARKS SURRY1
- MISL.0CATED SUPPORTS "AS BUILT" CCNDITICN CAUSED IN
' WRCNG SUPPORT TYPE MMORITY.0F. PIPE OVERSTRESS PRCB-
- DIFFERENT PIPE GECMETRY LEMS, NOT ALSEBRAIC SLM% TION BEAVERVALLEY NOT SPECIFICAU.Y IDENTIFIED "AS Bu1LT".CCNDITION RESULTED. IN e
LICENSEEREPORTED"ASBUILT 50TH PIPE #iD SUPPORT OVERSTRESS CCNDITIONS DIFFER-SIGNIFIC#4TLY FROM CRIGINAL DESIGN" FITZPATRICK
- DIFFERENCESSIMILARTOSURRY LICENSEE IS USING "AS BUILT" CCNFIGLRATICN FOR REANALYSIS
~
PILGRIM ' SNLEBER SIZING WRCNG PLANT SHUTDCH4 TO RESTORE.. SNLEBER PIPE ATTACH ENT WELDS ORIGINAL DESIGN CONDITICN SNLEBER SUPPORT STEEL ASSENLY, ,. BRlNSWICK U2 ' PIPE SUPPORTS LNDERSIZE BOTH LNITS SHUTDOW TO RESTOE ORIGINAL DESIG1 CmDITICN GINNA - PIPESUPPORTSNOTBUILTTO SUPPORTSWEREREPAIREDDURING ORIGINAL DESIGN REFLELING OUTAGE ST.LUCIE MISSINGSEISMICSUPPORTS INSTALLED / CORRECTED SUPPORTS e SUPPORTS ON MONG PIPING BEFORE START UP FROM REFUELING NINEMILEPOINT ' MISSINGSEISMICSUPPORTS INSTAU.ED SUPPORTS BEFORE START UP FRCN REFLELING INDIANPOINT3 MISLOCATEDSUPPORTS LICENSEE PERFORMING "AS BUILT"
' SUPPORT CCNSTRUCTION VER(FICA" ION TD BE CCNPLETED BY JULY ..
DAVIS-BESSE1
- MAIN STEAM LINE SUPPORTS @ PORTS WOULD BE OVERSTRESSED.
MISSING GUSSETS HEPAIRS WILL BE CCWLETED PRIOR TO START UP
@ e .. -
. - - - - , ,-,-s- - -
w- - - . - --
, , , , .y-, - - -
_ .. _ . . _ _ _ . . . _ . .. . _ .. . _ . 1 1 1 AS BUILT PROBLEMS IEB 79-14 (JULY 2, 1979) REQUIRED ACTION IDENTIFYINSPECTIONELEMENTSIN30 DAYS INSPECT 1/2 0F ACCESSABLE SYSTEMS IN 60 DAYS INSPECT ALL REMAINING SYSTEMS IN 120 DAYS
~
PHASED RESPONSES ARE REQUIRED . ACTION REQUIRED FOR NONCONFORMANCES IDENTIFIED EVALUATEEFFECTdNOPERABILITYANDCOMPLY WITH ACTION STATEMENTS OF IECHNICAL SPECIFICATIONS ANY NONCONFORMANCES IDENTIFIED WHICH RENDERS SYSTEM (IRAIN) INOPERABLE, EXPEDITE REMAINING INSPECTIONS SUBMIT EVALUATION OF NONCONFORMANCE UPON FSAR
~
CRITERIA 0R, CORRECT / REPAIR NONCONFORMANCES CORRECTIVE MEASURES TO UPDATE / REVISE DESIGN DOCUMENTS (DRAWINGSs ETC.)
- Seth W Stee p9. * > gem men o .-
MAIE YANKEE SEISMIC ESI64 PRESENT kSIGN HOUSNER SPEbTRA MCHORED AT 0.'Is IFLICENSEDTODAY LA Y 1.60 SPECTRA MCHORED BETWEEN OERAli SEISMIC CWSEPNATISM AT 1%INE YMKEE WAS REVIEWED PRIOR TO START UP ONGOING STAFF MTIW S 9 a
b $2 d t,dk A
-= u
{Q dwq.u i
- STATUS OF SEP SEISMIC REVIEWS ,[f' *y, ucn h IN#
- E TWO PLANTS NO SPECIFIC SEISMIC DESIGN BASI IWOPLANTSDESIGNEDTOUBC{FPMte&xdt .. ,
, gg
- OTHER PLANTS DESIGNED TO "G" VALUE SIMILAR TO CURRENT CRITERIA (EXCEPT WEST C0AST PLANT)
FOR OLDEST Six SEP PLANTS - LICENSEES INSTITUTING ANALYTICAL STUDIES -. NEWER FIVE SEP PLANTS BEING EVALUATED BY STAFF AND CONSULTANTS
- DRESDEN 2 EVALUATION NEARING COMPLETION - SEVERAL ISSUES REQUIR EITHERRETROFliTINGORDETAILEDSTUDIES LESSCNS LEARNED FRFI REIEd 0F SEP PIMS Sm.rMES GENERALLY ADEQUATE .TORISON MUST BE ADDRESSED PIPING GENERALLY ADEQUATE IF INELASTIC EFFECTS CONSIDERED SUPPORTS REQUIRE CASE BY CASE REVIEW
- FRAGILE EQUIPMENT h AT11ERIES, SMALL LINES WITH l'0VS) REQUIRE CASE BY CASE REVIEW OPERABILITY AND FtNCTICNABILITY OF EQ91 PENT REQUIRES MORE DETAILED EXAMINATICN LOWER QUALITY ASSURANCE THAN REQUIRED TODAY
- PROCUREENTPROBLEMS 6
~
m UCS ETITIm FOR SEISMIC EEVALUATIm l FOR ALL OPERATING RE-EVALUATE THE FOLLOWING: 1E 3i MAGNITUDE (INTENSITY)dFSSE hmERMINE FREEflELD GROLND MOTIm BETERMINE MOTIm 0F STRUCTURES DETERMINE MOTIM OF PLANT EQUIPENT : 1 COMPARE SEISMIC LOADS, LNDER APPROPRIATE LOAD C0f6INATIONS, l CN STRUCTURES, SYSTEMS AND COMPONENTS WITH All.DWABLE LDADINGS l l INSPECT 'EE PLANT TO DETERMINE METER "AS BUILT" PLANT l CONFORMS TO DESIGN SPECIFICATION l i
. . l l
\
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""~']_ _"*
# 4'""*R6*ep W e e w w.- e see - e . . _+_,
SEISMIC ISSUES EFFECTING OPERATING PLANTS
- 1. CRITERIA CHANGES THAT HAVE TAKEN PLACE DURING THE PAST 15 YEARS HAVE MADE SOME PLANTS MORE CONSERVATIVE THAN OTHERS:
SOIL-STRUCTURE INTERACTION R$SPONSESPECTRA DAMPING VALVES
~
- 2. COMPUTER CODES: ,
h.@ ^ M,ETHODS UTILIZED TO PERFORM CALCULATIONS 4
- 3. QUALITY ASSURANCE:
FACILITIES NOT CONSTRUCTED AS DESIGNED CALCULATIONS DO NOT REFLECT "AS BUILT" CONDITION EQUIPMENT NOT PROCURED AS SPECIFIED 6+ 4 _ _ _ _ _ _ - - - _ ^ ~ . _ _ _ _ _ . _ _ .
. PROPOSAL TO RESOLVE SEISMIC ISSUES
l. DETAILED STUDY OF CRITERIA USED FOR EACH OPERATING PLANT, BASED ON IHIS STUDY DETERMINE FRIORITY OF UPERATING PACILITIES WHICH SHOULD BE ANALYZED FIRST.
. SAMPLE EXAMPLES OF THE CRITERIA IHAT CAN BE USED TO ASSESS THE.
PLANTS ARE: SEISMIC INPUT RESPONSE SPECTRA i DAMPING VALUES , LOAD COMBINATIONS ETC..
- 2. UTILIZE FINDINGS FROM SYSTEMATIC EVALUATION PROGRAMS .
INCORPORATE SEISMIC ISSUES THAT ARE GENERIC TO.THE STUDIED . PLANTS STUDY MODIFICATIONS THAT SHOULD BE IMPOSTED ON OTHER j OPERATING FACILIITES USING FINALIZED SEP RESULTS TO DEVELOP LONG RANGE PROGRAMS
- 3. CONTINUE TO QUANTIFY SEISMIC CONSERVATISM: .
SSMRP TAKS ACTION PLAN (TAP) A-40 1 , l EFFECTIVE g LEVEL VS PEAK ACCELERATION OTHERS 1
- 4. DEVELOP CAPABILITIES IN VERIFICATION OF COMPUTER CODES IN THE MAJOR DESIGN OF STRUCTLRES, SYSTEMS, EQUIPMENT AND COMPONENTS BENCHMARK PROBLEMS TO TEST THE VALIDITY OF COMPUTATIONAL ii)
METHODS UTILIZED IN THE COMPUTER CODES. I i I g ease.m eegmene ven emome. ee . .. . ..- * . .- . . . ...+c... . .-** +. m._ , .-_-.--..-...._m...,,_,__m _~,,,.,.,_m;._,._,_,,. . . - - , - . _ , . - _ , , . . , - . . _ , _ _ . . .m ..,.m__---.. ,_m , . _ . ,.~ _ -
I
- 5. REVIEW THE IMPACT DUE TO'THE' DIFFERENCE BETWEEN "AS' BUILT"~AND
~
DESIGNED CONDITIONS AND OTHER QUALITY ASSURANCE PROBLEMS PROPER ACTIONS WILL BE TAKEN ACCORDINGLY 6 ~. USE FOREIGN DATA TO BETTER' ASSESS AND' W Y SEISMIC M 'b CONSERVATISM THE FINAL SEISMIC PROGRAM SHOULD BE FORMULATED BASED ON THE
. EXPERIENCES AND RESULTS GAINED FROM THE PRESENT SYSTEMATIC EVALUATION PROGRAM AND SEISMIC SAFETY MARGIN RESEARCH PROGRAM OO J
9 e n-e+.. . . . . . . , . . . , . . . _ , . . . . . . . . . . , . . . . ..
TASK ACTION PLAN A-40 THE OBJECTIVE OF THE TASK ACTION PLAN A-40 IS TO RESOLVE SOME OF THE SEISMIC ISSUES ON A SHORT TERM BASIS. PHASE I - RESPONSES OF STRUCTURES, SYSTEMS AND COMPONENTS ELST0-PLASTIC SEISMIC ANALYSIS SITE SPECIFIC SPECTRA NONLINEAR STRUCTURAL DYNAMIC ANALYSIS
~
. SOIL STRUCTURE INTERACTION PHASE II -SEISMIC INPUT DEFINITION STUDY OF EARTHQUAKE SOURCE MODELING ANALYSIS OF NEARFIELD GROUND MOTION a
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n-o OUTLINE I. Introduction and Overview A. In previous Commission briefing we have discussed several issues related to seismic capabilities of operating reactors.
- Systematic Evaluation Program briefing
- Unresolved Safety Issue briefing (A-40)
- Briefing related to Show Cause Orders of March 13, 1979 B. Our purpose today is to
- provide an update and overview of all related ongoing seismic review issues
- discuss lessons ~1 earned from our current seisnic efforts
- provide a conceptual background of future actions in the seismic review area - Our " Seismic Game Plan" (Slide 1) ,
II. Chronology of Seismic issues for operating reactors (Late 77) A. SEP Seismic Reviews - 11 older plants (seismic review is controlling task in completion of review) , (Feb. 78) B. A-40 Seismic Design Criteria - An Unresolved Safety Issue (March 0, 70) C. Ba:cplate and Anchor Bolt Bulletin (March 13,79) D. Show Cause Orders - 5 plants - Algebraic Summation (March 28,79) E. UCS Petition for Seismic Reanalysis of Operating Reactors (March 30,79) F. Velan Valve Weight Bulletin (April 14,79) G. Algebraic Summation Bulletin (May 3,79) H. Commission Briefing on Maine Yankee (Issue of adequacy of original SSE) (May30,79) I. Commission Briefing on Beaver Valley (advised that "As Built" is a problem) (Slide 2) A. Baseplate Flexibility IEB 79-02 (issuedMarch8,revisedJune21) B. Required Action
- Account for flexibility in anchor bolt loads
- Verify proper factor of safety
- Address how cyclic loads were accounted for in design (i.e.,preload)
- Testing requirements (Sample size, torque or pull test)
o , O n C. Responses are due July 6,1979 D. Preliminary Results
- Extensive Installation defects on some plants
- Some plants have few problems
- Some cases of missing bolts, sleeves
- Some lesser significant installation problems (misalignment, thread engagement, hole size)
- Some plants have' found entire seismic support missing (St. Lucie and Nine Mile Point)
(Slide 3) IV. A. Velan Swing Check Valve Weight - IEB 79-04 (issued March 30, 1979)
- 8. Required Action (3,4,6,8,10 inch valves only)
- Identify affected systems ,
- Verify correct valve weights
- Re-evaluate piping system
- Identify necessary modifications
- Identify analytical technique used C. Responses were due May 1,1979 D. Findings
- 48 of 65 plants used Velan Valves
- Remaining plants are shutdown and have not yet responded
- 5 plants required minor modifications (1 hanger overstressed plus miscellaneous hanger adjustments)
- Additional Velan Valves have problems (Visit to Velan by IE/NRR)
- Will include valve weight problem in "As built" bulletin which will be discussed later
. n
.ide 4) V. A. Algebraic Summation of Intra-Model Response in Computer Codes.
IEB 79-07 (issued April 14, 1979) B. Required Action
- Identify Affected Systems Reanalyze using appropriate code
- Code verification plus listings C. Responses were due April 24, 1979 D. Findings
- 25 plants affected
- Codes involved were Shock 2 - Stone and Webster
- Arthur D. Little Co.
ADL Pipe Westdyn - Westinghouse Daps - General Electric Pipdyn 2 - Franklin Institute
- Staff review of 16 plants I's complete (Brunswick 1/2 and IP 3 are partial and units are operating)
(511d25) E. Plants which used Algebraic Summation
- Varried in extend of use from 100% of safety systems to <
1-2 lines
- RCS main piping reviews have been completed (No modifications were required)
- Some plants voluntarily shutdown to repair "As Built" problems (i.e., Pilgrim and Brunswick 1/2) i i -
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r - . .. l l i 1 m holide6) VI. Maine Yankee Seismic Design
- Present = Housner anchored 0.lg
- Today = Reg. Guide 1.60 anchored 0.13 - 0.2g
- Overall seismic conservatism at Maine Yankee was reviewed prior to start up
- Ongoing staff actions (Review other Operating Reactors to determine if any others are as significant as Maine Yankee, etc)
"(Slide 7) VII. SEP Reviews and Lessons Learned A. Status
- 2 plants no seismic design
,- - 2 plants built to Uniform Building Code
- Other plants designe do a "g" value
- 6 plants (oldest) - licensees are conduting evaluations
- 5 plants (newer) - staff and consultints are evaluating
- Dresden 2 is nearing completion
'B. Lessons Learned
- Structures are generally adequate however effects of torsion must be addressed ,
- Piping is adequate if inelastic effects are considered
- Supports require a case by case review
- Fragile equipment requires a case by case review (Batteries, small lines with motor operated valves)
. - Operability and functionability of equipment requires more detailed examination
- Earlier QA not as rigerous as today (procurement problems i.e. purchase spec required 0.2g but no document action to verify)
talide 8) VIII. UCS Petition for Setsmic Reevaluation A. Re-evaluate all plants IAW today Standard Review Plan B. Inspect plant to verify "As Built" condition IX. Plants with Significant Differences between Original Design and "As Built" condition of pfptng systems A. Discuss Slide . B. Discuss Bulletin C. Inspection scope
~
- Safety related piping systems with diameters 2".
Seismic analysts input parameters as shown on design drawings and specifications for:
- Pipe run configurations
- Supports .
- Valves
- Pipe whip restraints D. Inspect hardware for conformance to design documents
- Half of accessible systems within 60 days
- Other half within 90 days
- All normally inaccessible systems at next scheduled cold shutdown E. Correct nonconformances
- Hardware changes
- Seismic reanalysis F. Satisfy the requirements et tech spec action statements re operability G. Provide assurance that future modifications will be reflected in design documents and seismic analyses.
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2, BME PLATE R.EXIBILIT( AND ANOR BOLTS IEB79-02 (MARCH 8,1979) 1 (JLNE 21,1979) REQLl! RED ACTIm
- ACCOLNT FOR FLEXIBILITY IN McHOR BOLT LDADS WRIFY PROPER FACTm 0F SAFE 1Y DESIGNREQUIREENTSFORCYCLICLOADS ,
TESTINGPROGRAM RESPWSES DLE JULY 6,1979 PRELIMINARY %SULTS SOE PLANTS WITH EXTENSIVE INSTALLATIm kFECTS SOE PLANTS W11H FEW PRmLEMS SGE GROSS INSTALLATIm PRmLEMS (MISSING BOLTS, SLEEVES)
' SOE WITH MINOR INSTALLATIm PROBLEMS (MISSALIGNENT, TifEAD ENGAGEENT)
SWE PLMTS HAVE REPORTED ENTIRE SUPPORT MISSING
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% ANALYZE USING APPROPRIATE C00E CG)E VERIFICATION RESPONSES WERE DUE APRIL 24,1979 FINDINGS 3 OPERATING PLANTS REPORTED USE OF ALGEBRAIC SinuTIm C@ES INVOLVED WERE SHOCK 2 - STWE & MiBSTER AX. PIPE - ARTHUR D. LITTLE CO.
WSTDYN - W STINGHOUSE DAPs - GENERAL El.ECTRIC PIPMN 2 - FRANKLIN INSTITUTE . NRR STAFF HAS COMPLETED TE REVIEW OF 16 PLANTS . s
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STATUS OF EVIEW 0F 5 SMIGN PLANTS Smay 1 - PIPING ANALYSIS RMODIFIQTICNS TO REDUG PIPE STESS WILL BE C04PETED Jz JULY /3 110F @ PRWLEMS REQUIRE HARDWARE FIX SLFPORT ANALYSIS WILL BE C0ffLETED IN AUGUST i
%SLfE OPERATICN IN SEPTEf6ER SURRY2 -
SHLEXH4 FOR STEAM GENERATOR REPLA NNT NO REANALYSIS RESULTS YET NO SCEDULE BEAVER VALLEY - PROPOSAL FOR INTERIM OPERATION LNER REVIEW MCDIFICATICNS IN PROGRESS TD BE COPPETED BY 3 REGNTLYJDENTIFIED ADDITIONAL PIPE STRESS PRGLEMS AND ABOUT MJ SUPPORTS TO BE ANALYZED FITZPATRICK - LICENSEE ESTIl%TES THAT PROPOSAL FOR REST #fTION OF OPERATICN WILL BE SLEMITTED AB0ur JULY 2,19B ALL PIPING ANALYSIS AND ALL SUPPORT ANALYSIS IN INACCESSABLE AREAS, INC13JDING ANY NEGSSARY MCDIFICATIONS,
. WILL BE C0ffLETE RRIOR TO START LP MAINE YANKEE - APPRWAL FOR START UP ISSUED MAY 214,195 9
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COOLING POINTBEACH2 \I y x BRUNSWICK 1 ADL PIPE & IAPS EXTENSIVE x BRUNSWICK 2 y x INDIAN P0!nr 3 ADLPIPE&WESTDYN p SALEM 1 PIPDYN EXTENSIVE x INDIAN POINr 2 ADLPIPE 5 LINES 000PER SRV LINES ONLY GINNA 2 LINES l I x MILLSTONE 1 2 LINES x MIEST%E 2 6 LINES x NINE MILE POINT p LIMITED WESTDYN 1 LINE x COOK 1 x COOK 2 WESTDYN P x ROBINSON 2 RCSONLY x TmKEY P0!gt 3/4 x ZIm 1 y Zion 2 N Y X PILGRIM 1 IMPS RCS & % IN STREAM ONLY x = RESOLVED
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MAIE Y#4(EE SEISMIC ESI@ PRESENTDESIGN HOUSNER SPECTRA M010 RED AT 0.1s. IF'LICENSEDTODAY
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OERALL SEISMIC CONSERVATISM AT MAINE YANKEE WAS REVIEWED PRIOR TO START UP
' 04GOINGSTAFFACTIONS e
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STATUS (F F SEISMIC EVIEWS TWO PLANTS NO SPECIFIC SEISMIC DESIGN BASIS TWO PUWTS DESIMED TO lEC OTHER (EXCEPT EST Pl>NTS COAST PLANTDESIGNED TO)"G" VALE SIMILAR TO CmREN
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FOR OLDEST SIX F PLANTS - LICENSEES INSTITUTING ANALYTICAL STUDIES M FIVE F PLANTS BEING EVAL!JATED BY STAFF AND CWSULTANTS DRESDEN 2 EVALUATIm NEARING CCNPLETION - SENERAL ISSLES REQUIRE EITER RETROFITTING OR DETAILED STUDIES 6 0 l
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(ISSuiS LEARG FM EIH OF SEP PLANTS STRUCERES CEERAU.Y ADEQUATE - TOISON MJST BE ADDRESSED
- PIPING GENERALLY ADEQUATE IF IVELASTIC EFFECTS CMSIDERED l
SUPPORTS REQUIRE CASE BY CASE REVIEW
- FRAGILE EQUIPENT-(BATTERIES, SMALL LINES WITH ES) REQUIRE CASE BY CASE REVIEW OPERABILITY AND ftNCTICHABILITY OF EGUIPMNT REQUIRES MORE DETAILID EXAMINATIm LOWER EJALITY ASSURANCE THAN REQUliED TODAY PROCUREFE R PRS LEMS
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UCS ETITIm FOR SEISMIC EEVALUATIm l l FOR ALL OPERATING REACTORS dLI M iY I h!NE MAGNIT10E (INTENSITY) OF SSE
' DEW.RMINE FREEf! ELD GROLND MGTIm
- DETERMINE MGTIm 0F STRUCTURES DETERMINE MOTICN OF PLANT EQUIPtENT -
- CGiPARE SEISMIC LOADS, LNDER APPROPRIATE LDAD COSINATIONS, m STRUCTURES, SYSTEMS AND CCNPONENTS WITH ALLDWABLE LDADINGS INSPECT TE PLANT TO DETERMINE METTER "AS BUILT" PLANT CONFORMS TO DESIGN SPECIFICATION
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SEISMIC ISSUES EFFECTING OPERATING PLANTS
- 1. CRITERIA CHANGES THAT HAVE TAKEN PLACE DURING THE PAST 15 YEARS HAVE MADE SOME PLANTS MORE CONSERVATIVE THAN OTHERS:
SOIL-STRUCTURE INTERACTION RESPONSi SPECTRA DAMPING VALVES
- 2. COMPUTER CODES:
METHODS UTILIZED TO PERFORM CALCULATIONS
- 3. QUALITY ASSURANCE:
FACILITIES NOT CONSTRUCTED AS DESIGNED CALCULATIONS DO NOT REFLECT "AS BUILT" CONDITION EQUIPMENT NOT PROCURED AS SPECIFIED
- - - - _. . - --= . . _ _ - . -- .
o . 4 i i PLANTS WITH SIslFICMT DIFEENES EMEN
) ORIGINAL ESIGN AND "AS BJILT" CmDITIm 0F PIPING SYSTDE
,M DIFFERENCE % MARKS SURRY1 'MISLDCATEDSUPPORTS "AS BJILT" CmDITim CAUSED IN '
- MtmG SUPPORT TYPE PWJORITY OF.PJPE OVERSTRESS PRCB-
- DIFFERENT PIPE GEOETRY LEMS, NOT AL1EBRAIC SLM% TION
! BEAVER VALLEY ' NOT SPECIFICALLY IENTIFIED "AS BUILT"jgNDITIm RESULTED IN
- LICENSEE REPORTED "AS BUILT BOTH PIPE MD SLPPORT OVERSTESS C mDITIONS DIFFER SIGNIFICANTLY
!2 FROM m!GINAL DESIGN" FIT 2 PATRICK ' DIFFERENCES SIMILAR TO SURRY LICENSEEISUSING"ASBUILT" j CmFIGLRATIm FOR REANALYSIS
~
- PILsRIM ' SNLEBER SIZING WRmG PLANT SHUMM TO RESTORE
- SNLEBER PIPE ATTAOMNT WELDS ORIGINAL ESIGN CONDITION
' SNlEBER SUPPORT STEEL ASSEPSLY BtLNSWICK V2 PIPE SUPPORTS LNDERSIZE Bom LNITS SHUMM TO RESTORE ORIGINAL E SIM Cm DITIm
- GImA 'PIPESUPPORTSNOTBUILTTO SUPPORTS E RE REPAIRED DURING ORIGINAL DESIGN REFELING OUTAGE i
ST.1.UCIE ' MISSING SEISMIC SUPPORTS INSTALLED / CORRECTED SUPPORTS i ' SUPPORTS m Mt0NG PIPING BEFORE START UP FROM REFLELING l N!Pc Mitt POINT 'MISSINGSEISMICSUPPORTS INSTALLEDSLPPORTSBEFORESTART UP FRm REFLELING , l INDIANPOINT3 MISLOCATEDSUPPORTS LICENSEEPERFORMING"ASBulLT"
' SUPPORT CCNSTRUCTIm R CA"Im TO BE COPFLETED l
DAVISUSSE1
- MAIN STEAM LINE SUPPORTS SJPPORTS WOULD BE OVERSTRESSED. l MISSING GUSSETS EPAIRS WILL BE COPFLETED PRIOR l
- TO START UP j i
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4 PROPOSAL TO RESOLVE SEISMIC ISSUES l
- 1. DETAILED STUDY OF CRITERIA USED FOR EACH OPERATING PLANT. !
BASED ON THIS STUDY DETERMINE PRIORITY OF OPERATING FACILITIES WHICH SHOULD BE ANALYZED FIRST. SAME EXAMPLES ARE: SEISMIC INPUT %.___ ,, RESPONSE SPECTRA DAMPING VALUES, LOAD COMBINATIONS ETC.
- 2. UTILIZE FINDINGS FROM SYSTEMATIC EVALUATION PROGRAMS INCORPORATE SEISMIC ISSUES THAT ARE GENERIC TO THE STUDIED PLANTS STUDY MODIFICATIONS THAT SHOULD BE IMPOSTED ON OTHER OPERATING FACILIITES USING FINALIZED SEP RESULTS TO DEVELOP LONG RANGE PROGRAMS
- 3. CONTINUE TO QUANTIFY SEISMIC CONSERVATISM:
SSMRP TAKS ACTION PLAN (TAP) A-40 EFFECTIVE G LEVEL VS PEAK ACCELERATION OTHERS
- 4. DEVELOP CAPABILITIES IN VERIFICATION OF COMPUTER CODES IN THE MAJOR DESIGN OF STRUCTURES, SYSTEMS, EQUIPMENT AND COMPONENTS BENCHMARK PROBLEMS TO TEST THE VALIDITY OF COMPUTATIONAL METHODS UTILIZED IN THE COMPUTER CODES.
~' - ~ ~ ^ -
. . .. _ _ _ . . . . . _ . _ . . 1 7' ....--~ .. . L.
j a 4 , s 5. REVIEW THE SEVERETY OF THE DIFFERENCE BETWEEN "AS BurLT" i AND DESIGNED CONDITIONS AND PROPER ACTIONS TAKEN Acc0RDINGLY . . 6. USE FOREIGN DATA TO BETTER ASSESS AND QUANTIFY SEISMIC CONSERVATISM
- THE FINAL SEISMIC PROGRAM SHOULD BE FORMULATED BASED ON THE EXPERIENCE AND RESULTS GAINED FROM THE PRESENT SYSTEMATIC EVALUATION PROGRAM AND SEISMIC SAFETY MARGIN RESEARCH PROGRAM.
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- tfa1/79 f seism 1C. ANALVKES ANb As.BulLT PIPNG AcTioW
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- 6- Seismie wa. lysis inpu+ pa.<a.m +,,, as sk a,,,, ,,.,
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- 1. Inspee+ ka.edwe fer cdarm.,ee +s der,gn docused, a.. Hs.l+ af accessiI.te syks wi%;n 40 cims, b . o A m.e he if wi+4 h 90 J% s.
e, A ll a w r a ll y ina.eeessille sy.sks a + nexf seduled eat l sk&down . i g, co rvae.f- noncador w eees. c., FIM w e c.ka.n3et. . 6, S e is ,te. rea.ne.ly is . S= Msfy +4e requir*~*nh df +nck spec. ack 4. sk+s-enn re. opera.hilih . 5 Peav ,'de a.ssv e nner %+ fufure mel%ca.%s will he ecFlee w i.m des ay n dae-+x a.J seis~ta. l w tyses.
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a _ . . : _. , = 2 , . _ ,t.;f_f. . __ . . 1... , . _ . . f TASK ACTION PLAN A-40 THE OBJECTIVES OF THE IASK ACTION PLAN A-40 IS TO RESOLVE SOME OF THE SEISMIC ISSUES ON A SHORT TERM BASIS. PHASE I - RESPONSES OF STRUCTURES, SYSTEMS AND COMPONENTS ELST0-PLASTIC SEISMIC ANALYSIS SITE SPECIFIC SPECTRA NONLINEAR STRUCTURAL DYNAMIC ANALYSIS SOIL STRUCTURE INTERACTION PHASE II -SEISMIC INPUT DEFINITION STUDY OF EARTHQUAKE SOURCE MODELING ANALYSIS OF NEARFIELD GROUND MOTION 9
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440a SEISMIC ISSUES EFFECTING OPERATING PLANTS
- 1. Criteria changes that have taken place during the past 15 years have made some plants more conservative than others:
Soil-Structure Interaction - Response Spectra i Damping Valves
- 2. ComputerCddes:
Logic of the algorithm utilized to perform calculations
- 3. Quality Assurance:
. Facilities not constructed as designed Calculations do no reflect "as built" condition 4
l l l l l
l l PROPOSAL TO RESOLVE SEISMIC ISSUE f
- 1. Continued operation of existing facilities based on seismic conservatism:
Each discipline, i.e. geosciences, structural, mechanical, and electrical impose their own conservatism. Change in spectrum shape accompanied by increase in damping Enveloping response spectra and time histories ' Elastic dynamic analysis Multi-directional earthquakes Peak widening of floor response spectra System Redundancy Other loads (wind and pressure) influence design Justification for continued operation can be strengthened by making a simplified risk assessment study as done in Diablo Canyon and recently proposed by H. Levin.
- 2. Initiate a task force to work on the seismic issues including disciplines:
Geosciences Structural Mechanical
~
Electrical 3 .. . _ . . . - . . . - .
2 s
- 3. Detail study of criteria used for each plant. Determine priority of operating facilities based on this study:
,- SeismicInputt.[ h . / SS( ,I4 20 42)
- , - Spectra
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26 f. f.r a reg SZ) f Damping Valdes ty Load combinations and factors >k SS 2' '
- 4. Utilize findings from SEP; M 6W .4P M -
Incorporate seismic issues that are generic to the studied plants , Study modifications that should be imposed on other operating facilities
/
/ Using finalized SEP results to develop long range programs
,(' 5. Continue to quantify seismic conservatism:
SSMRP k j d ;)t 1 Task Action Plan (TAP) A-40 Effective g level vs peak acceleration Others
- 6. Develop a h program to verify computer codes in the major design p 9areasofstructures, systems,equipmentandcomponents:
- .ft
_y . ~~ t Benchmark problems to test the validity of computational schemes y utilized in the computer code i ,Li 5 Benchmark problems to test the validity of the design l eYf[ . . , . . . , . . . _ . . . . . . . . , . . . . , _ . .
O 3 7. u) Use foreign data to better assess quantify seismic conservatism: h 4
,~ _
THE FINAL SEISMIC PROGRAM SHOULD BE FORMULATED BASED ON THE EXPERIENCE
<w' AND RESULTS GAINED FROM THE PRESENT SEP AND SSMRP. A #:M* ' '.
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NUCLEAR-0UAKE : ~- l .
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BY STAN;. BENJAMIN *
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4 Ss THE NUCLEAR REGULATORY C0!iHISSION'S STAFF RSSERTED THURSDAY. - STAFF NEMBERS TOLD - THE .. C0iiiiISSION THAT A BUI.LETIN HOULD 500N BE ' ' N dI-d U-- to 1.0 r.f L L .,.,,ic. P L n-,t i i ,e,, t i t t.t n x . .li n i.u.G d i v u t e dn-- VF He-li,g Reut
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nT v Ttti R*iiNING liHETHER PIPE DESIGNS ARE SAFE AND leiETHER THEY HERE ACTUALLY. BUILT AS.. DESIGNED. . D.E n,.... ....r iu unto exun uUn.,.E-Pn00F o des.I.,.1.i,4 nxt--. Cu,,1d IDt.o - n st D S e nt 3.u,ud- uc-t C a-.id tei. OF THE POSSIBILITY THAT INADEOUliTE CONSTPUCTION C0liLD LEAD TO THE . BREAKING OF THESE PIPESs WHICH ARE VITAL TO REACTOR SAFETY SYSTEMS DURING fin EARTHOUAKE. THE FAILURE OF THESE SAFETY SYSTEMS COULD CAUSE . c A NUCLEAR ACCIDENT AS SEVEREs OR UORSE, THAN THE ONE THREE MONT.HS AGO . 8: i h.xi e. e .u. i L E t ac la- r.i o is 1tnn o n n- xi3idatnus -- i-- rA.
. HILLInli RUSSELL, HEAD OF A COMMISSION. TASK FORCE ON SEISMIC .
DESIGN, PRESENTED A LISTING OF SIGNIFICANT DIFFERENCES BETHEEN ORIGINAL DESIGN AND AS-BUILT-CONDITION OF PIPING SYSTEMS IN THE 11 PLANTS. .
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z e ce d . y), E s M . "T~*s v. to - h 5'-/MG . M. aus LlI4l% w.Q f ' 3e4/ A. Show Cause Order of 3/13/79
- 1. Addressed Algebraic Summation of INTRA M0DAL Response -
Code involved SHOCK 2. Required licensee to show cause why:
- a. should not reanalyze
- b. should not modify as necessary
- c. shut down during reanalysis and modification MGeaWD
- 2. Technical Concern
__a. 3-D earthquake input results in 9 Force Vectors - I-D b >KaM N l 1.e., X earthquake causes X, Y, Z response, Y W w sa, sap S.D m '- wael4$ earthquake causes X, Y, Z response and Z earthquake ' A T AM. u _- _ causes X, Y, Z response,
- b. Problem - how to mathematically combine co-linear responses.
(SRSS, Absolute Sum, Mary 'iethc9, or Algebraic)
- c. Algebraic method is " unsatisfactory" because responses could cancelt resulting in the limit in no loading at all.
- 3. Safety Concern
- a. Systems were affected which could both cause an accident (LOCA) and cause the failure of mitigating systems (ECCS). -
- b. Stresses at Beaver Valley evaluated over weekend of 10-11 March were significantly (3-6x) over allowable.
- cud 3 1_
- 4. Licensee's Response to Orders
- a. Plants were shut down
- b. No hearings were requested
- c. Licensees proceeded with analysis and modifications.
LIDE 2 _ B. Status of Evaluation Today
- 1. Maine Yankee - reanalysis & modifications are complete.
Order was lifted 5/24/79.
- 2. Beaver Valley
- a. Piping reanalysis is complete - 3 piping systems require modification (2 shock suppressors, piping reinforcement at small branch line welds to larger
- pipe - river water). -
- b. Support reanalysis - 97 modifications - Proposing to make 15 and defer others for 6-7 weeks until refueling.
- c. Modifications for interim operation will be complete by July 6.
( Qu e b * ** '*" * .
- 3. Fitzpatrick
- a. Piping reanalysis will be complete by July 1.
- b. Support analysis in inaccessible areas and related modifications by July 1 also,
- c. Will request interim operat-ion during completion of support analysis in accessible areas.
's .
- 4. Surry 1
- a. 33 of 69 piping analysis complete - Mr^:b* ^, h 3*
t; . 160 of 187 support analysis complete - W -- ^b
- c. Only 1 support to ,mesm6p.- p 400 75 lA M 10 M A PvW hisc% 64,60W.
- d. Other minor modification not related to order are being accomplished.
LIDE 3 C. Status of Staff Review
- 1. Related Areas - M 3T8d b Mf& gMWMW
- a. Code verification - W M h M'*/ M
- b. Hand calculation methods " equivalent static"
- c. IEB 79-02 Status
- d. All safety systems & method of analysis - tm "M h"
- e. Other codes used (Shock 0/1)
- 2. Soil Structure Interaction at Surry & Beaver Valley
- a. New method proposed by licensee e.sts
- b. Comparison to R. G. 1.60;and R. G. 1.61
- c. Approved by staff on 5/25 (during 1.5 factor'and Beaver Valley 1.2 factor to account for variation of soil properties).
ggop Z BO 79 07 yas(emat. pg( o pas.j -
- c .. : ,
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- ORDER TO SHOW CAUSE e WHY THE LICENSEE SHOULD NOT REANALYZE THE FACILITY l
I PIPING SYSTEMS FOR SEISMIC LOADS USING AN APPROPRIATE I PIPING CODE; - I 5 E
- i.
- WHY THE LICENSEE SHOULD NOT MAKE ANY NECESSARY MODI-FICATIONS FOLLOWING REANALYSIS; i
E. l
';
- WHY FACILITY OPERATION SHOULD NOT BE SUSPENDED.PENDING
- SUCH REANALYSIS AND COMPLETION OF ANY REQUIRED MODI-r E FICATIONS E
E i.. t Ii
PIPING REANALYSIS STATUS R$ PORT"AS '0F 6/114/79 l MY BV F S-1 S-2 ; 19 86 96 69 67' PIPE ANALYSES To DO_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ' _ _ COMPLETED WITHIN ALLOWABLE 19 83 73 33 0 - COMPLETED ABOVE ALLOWABLE 0 3 0 0 0 (HARDWARE CHANGE REQUIRED) PIPING SUPPORTS TO EVALUATE 2 -732 875 887 808 COMPLETED WITHIN ORIGINAL 0 635 263 160 - 0 DESIGN ; COMPLETED ABOVE ORIGINAL 2 97 23 1 0 DESIGN (HARDWARE CHANGE - REQUIRED)
- ESTIMATED ANALYSIS COMPLETION DATE 5/2 7/1 7/1 6/30 (10/1) (9/15)
ESTIMATED. START.UP.DATE. .5/24 7/6 5CL 6/20 10/1*
._5STIMATED COMPLETION OF STEAM GENERATOR REPLACEMENT
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- FOR COMPUTER CODES USED; VERIFY WHERE ALG SUM USED PROVIDE CODE LISTING e
- FOR HAND CALC. METHODS DESCRIBE / JUSTIFY METHODS
~
- PROVIDE STATUS OF RESPONSE TO IEB 79-02 IDENTIFY ALL SAFETY SYSTEMS AND ANALYSIS METHODS
. FOR CODES USED FOR PREVIOUS EVAL. OR REANALYSIS i- PROVIDE INFORMATION ON VERIFICATION i
i t e
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OPERATING REACTORS ESPONSE TO IEB 79-07 LNIT CODE RB%RKS BEAVERVALLEY SHOCK 2 EXTENSIVE, ORDERS /D FITZPATRICK x MAINE YANKEE SLRRY l l SLRRY 2
~
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%d A. Background Mg April 26,1979 - Briefing on status of five shutdown plants b ~
F'.A. s - Gl d
.2 May 3,1979 - Briefing on Maine Yankee C 1m - o s>
May 24,1979 - Order issued approving restart of Maine Yankee
~ ' ga d a .m AI M M , B. Today's Briefing is on Beaver Valley h f[,
.ns Objective is to present current status of the review.
' N g , 7, 1. O'
- 2. During the April 26 briefing the staff advised that the licensee's g A Qt estimated completion date for analysis was May 6,1979. The licensee's analysis is not yet complete, gh a
- 3. 83 safety-related piping problemi to be analyzed:
W '
- 80 complete within allowable
- 3 require hardware.nodifications
. (2 snubbers + branch line reinforcement 9 welds)
- Plus 18 new problems which were reanalyzed for water hammer and 4
SSE but not OBE
- 1 of 18 complete for OBE - 0.K.
- 4. 741 safety-related pipe supports: ..
- 623 complete within allowable
- 15 require modification
- 103 yet to be comr.leted
- 5. Verification of con nuter codes (SHOCK 3 and NUPIPE-SW)
- same codes which were reviewed and approved on Maine Yankee
- code listings
- bench marking
- sample problem for independant analysis
+ m eese w endsg
~'
- 6. Approval of soil structure interaction methodology:
- accounts for energy lost in soil under site
- reduces stress on piping and supports
- method used did not consider variation of soil properties
- staff approved subject to using 1.2 " bump factor" to account for soil properties variation (approved by letter 5/25/79)
- Delay in staff approval had minimal impact because entlysis did y
not have to be redone. Only one system (river water) had pipe stress above allowable due to 1.2 factor C. Summary
- 1. Licensee's reanalysis is not yet complete.
- 2. Modifications required based upon reanalysis have not been implemented.
- 3. Staff review of SSI, codes, and hand calculation complete.
- 4. Staff review of reanalysis results, necessary modifications are in progress.
4 D. Related review areas (backup information only)
- IEB 79-02, base plate flexibility:
for those supports where loads increase above original design, base plate flexibility must be addressed.
- IEB 79-04, Velan valves (26 installed in plant) - Ifcensee's reanalysis will incorporate actual valve weight (licensee letter 4/30/79)
- Requests for additional information:
. pipe break criteria
. seismic and/or motion
. system function questions w
vc-----
. .=.. -
PIPING REANALYSIS STATUS REPORT AS OF 5/29/79 BEAVER FITZPATRICy SURRY SURRY VALLEY 1 2 REANALYSIS TO PERFORM 83 96 72 CO PLETED WITHIN ALLOW 6BLE 80 46 29 QA ACCEPTED RESULTS) COMPLETED ABOVE ALLOWABLE ** 3 0 0 (HARDWARE CHANGE REQUIRED) PIPING SUPPORTS TO EVALUATE 741 1156 873 COMPLETED WITHIN ORIGINAL DESIGb 623 0 118 - (QA ACCEPTED RESULTS) COMPLETED ABOVE ORIGINAL DESIGN 15 0 0
.(HARDWARE CHANGE REQUIRED)
REANALYSIS ESTIMATED COMPLETION 6/10/79 6/15/79 DATE (LICENSEE ESTIMATE) ' il
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*PLUS 18 ADDITIONAL PROBLEMS WHICH WERE ANALYZED FOR WATER HAMMER
+ SSE BUT NOT OBE,
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~ U PIPING REANALYSIS STATUS REPORT AS 0F 5/29/79 BEAVER FITZPATRICp SURRY SURRY VALLEY 1 2 REANALYSIS TO PERFORM 83* 96 72 COMPLETED WITHIN ALLOWABLE '80 46 29 (OA ACCEPTED RESULTS)
COMPLETED ABOVE ALLOWABLE ** 3 0 0 EIARDWARE CHANGE REQUIRED) unL
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PIPING SUPPORTS TO EVALUATE 741 :1156 '873 COMPLETED WITHIN ORIGINAL DESIGb g'23 0 H8 (QA ACCEPTED RESULTS) COMPLETED ABOVE ORIGINAL DESIGN 15 0 0
-(HARDWARE CHANGE -REQUIRED) - . . -
REANALYSIS ESTIMATED COMPLETION gfjy79 6/10/79 6/15/79 DATE (LICENSEE ESTIMATE) x h/r s,h f a d h h k S b -[fi est) O(, p
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ZED FOR WATER HAMME
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DRAFT
~WRUSSELL/ja 7 /kQ'A (b.
5/2/79 A. Last Week's - Briefed on the status of the five shut-down plants I (4/26/79)
- 1. Advised that Maine Yankee review was nearing completion B. Today's briefing is on Maine Yankee
- 1. Objective is to present the staff basis for a recommendation to restart C. Show Cause Order of 3/13/79
- 1. Addressed Algebraic Summation Required licensee to show cause why:
- a. 'should not reanalyze
- b. should not modify as necessary
- c. shutdown during reanalysis and modification
- 2. Computer code addressed by Show Cause Order - SHOCK 2
- a. 19 safety-related piping problems
- b. reanalyzed using acceptable methods
! 1. 6 done by SHOCK 3
- 2. 5 done by NUPIPE-SW
- 3. 8 done using hand calculations
- c. staff has reviewed reanalysis methods and results and found acceptable (i.e., within allowable)
- 3. Verification of computer codes used for reanalysis (SHOCK 3 and NUPIPE-SW)
5/2/79
- a. code listings were provided and reviewed to determine that algebraic summation was not used and that other obvious errors were not present
- b. 3 sample problems with documented results from EPIPE were run on NUPIPE-SW and SHOCK 3 (Problems were " Coffee Table," "Hougard's Bend" and "Two Loop Reactor Coolant System.")
9~
- c. independent analysis by NRC consultant (BNL) of sample problems from 4 plants (MY, BV, F, S-1)
- results between NUPIPE-SW and SHOCK 3 agree + 10% with EPIPE
- d. conclusion - even though sample problem verification is not yet fully complete, based upon independent analysis, check of code listings and review of some sample problems, there is a high level of confidence that SH0CK 3 and NUPIPE-SW are acceptable codes.
- 4. Pipe supports associated with 19 SHOCK 2 piping problems
- Loadings increased above original design valves on 2 supports such that detailed support reanalysis was required
- Minor modifications (gusset plates) to account for baseplate flexibility were performed
- 5. -Strict interpretation of requirements of Show Cause Order have been satisfied
5/2/79 D. Staff felt that it was appropriate to go beyond order
- 1. hand calculation methods - reviewed against Standard Review Plan criteria and found acceptable
- a. used on some larger piping where simple geometry was involved s
- b. method of. pipe support determines acceptability of use of hand calculation -- economic trade off between fewer supports and more expensive calculations
- c. method used by Maine Yankee acceptable independent of pipe size
- 2. Balance of plant dynamic computer analysis
- a. SHOCK 1 - 4 versions a 1 version found acceptable before last week's meeting
~ o 3 other versions were identified as remaining issue to be resolved before restart
~
- b. SHOCK 1 is 1-D code o X direction input generates X response o Y direction input generates Y response o Since 1-D cannot compare directly to current 3-D codes
- c. comparison SHOCK 1 piping design by reevaluation of 10 of 7j[ problems has been completed satisfactorily. Earlier
. versions of SHOCK 1 have also been cross-checked to the latest version of SHOCK 1.
_-~. 5/2/79 E. Conclusion -
- 1. Maine Yankee has complied with the Show Cause Order
- 2. High level of confidence that balance of plant piping would be acceptable if reanalyzed. (For example, . hand calculation and SHOCK 1 methods result in acceptable pipe designs and supports
- 3. Licensee has agreed to reanalyze all SHOCK 1 analyses as part of a longer-term program. Sufficient confidence in the results from reanalysis performed to date to allow restart.
F. Back Up Material
- 1. IEB 79 base plate flexibility - for those supports where loads increased above original design base plate flexibility was addressed
- 2. lEB 79 Velan Valves - worst case (100% weight increase) results in only 30% stress increase. Not significant enough to prevent restart
- 3. IEB 79 responding to 3/13/79 Show Cause Order
- 4. IEB 79 TMI fixes
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'u- km-SEISMIC - PIPING S PLANT SHUTDOWN (Slide 1) I. STATUS BRIEFING - 5 PLANT SHUTDOWN - AREAS II. BACKGROUND A. March 13 -
Show Cause Orders
- Surry 1/2 (822 Mw each)
- Beaver Valley (852 Mw)
- Fitzpatrick (821 Mw)
- Maine Yankee (790 Mw)
- 8. Oct. '78
- Duquesne Light Informed by S&W
- S&W Doing System Mod Work - Learned of _W info From May 78 re: Check Valve Weights
- S&W Found Misapplication of Hand. Calculation Method ,
- Correction Found Areas of Local Overstress (Added one Snubber & Modified One Support)
C. Codes
- Hand Calculation Misapplied
- 1974 As-Built Verification - PIPESTRESS ,
- 1978 Recheck - NUPIPE
- NUPIPE Stresses Higher than PIPESTRESS
- Oct. 26, 1978 DLC Notified IE of Differences
- NRC - DLC Discussed Differences D. March 8,1979
- DLC Notified NRC Differences Attributable to SHOCK 2 Subroutine in PIPESTRESS
~ - SHOCK 2 Uses Algebraic Summation of Loads Calculated Separately for Horizontal and Vertical e.g. Components
2-E. Algebraic Summation
- OK Only if Use Time History Approach
- SHOCK 2 Not Conservative For Response Spectrum Analysis. (No Time Phasing)
- Load Combinations
- Horiz. EQ Gives H and V Direction Piping Movement Vertical EQ. Gives H and V Direction Piping Movement F. March 8 Mtg.
- S'ign I'ncreases in Support Stresses
- Lesser Increases in Pipe Stresses G. Meetings at S&W - March 10 - 12 ,
- Many Piping Systems Stresses Over Allowables Even for OBE
- Some Overstressed Piping - RCPB, Therefore Could Cause LOCA
- Other Overstressed Piping in Safety Systems Needed to Handle LOCA ,
- Analyses for BV; Inference for Others H. March 13 Orders
- Respond in 20 Days (Slide 2) - Why Not Reanalyze with Appropriate Code
- Why Not Make Mods. Per Reanalysis
- Why Not Be Shut Down in Interim
- Opportunity for Hearing - No, Interventions e , e - m - e., am--
e -
---,a , -.,,--*,-------p y .o , m , ,, - y,y, - , - - -. - - - - ,---- --w-- - - -. - - - - .. -., - - , - - -
+
III. STATUS SINCE MARCH 13 A. NRC Set-Up 4 Review Teams 5 Visits to Site 8 Visits to S&W (2 by DSS on SSI) 2 Meetings in DC s (Slide 3) B. Status of Licensees Evaluation'
- Maine . Yankee Analyses Completed - -
- No Modifications Necessary eA. cept 4< % kt ' ohs IV. STATUS OF STAFF's REVIEW (Maine Yankee Only - Others'Not Submitted) ,~-
A. SHOCK 2 TO SHOCK 3/NUPIPE COMPLETE l B. SHOCK 2 - Reanalyzed with SHOCK 3/NUPIPE ; Code Verification _
- Staff Reviewed Code Listing g
.[
Benchmarked, Code (Not Done) M ud-th Se # wt 1 o{ 19 W 4 E'AC / os Ac.rd pele .e - . Comparison ed - . w.(jabre A %_ i. e.im A % M M ' $ " C. Staff Looked Into Other Evaluation Practices bO O - Beyond Strict Order Interpretation, Letter Dated 4-2-79, Information Request o l. A ,( b,A , m,j
\
1 G%c.v.I wA m "A ' - Code Listings for All Computer Codes 41- ec*/.etS1 *Ted .
- u.u A. b . 6, 4 Re.cac .,
c Hand Calculation Methods [toML2.NdvA. Cakes M9 Status & Schedule for IE Bulletin 79-02
\h*h boev (suat bu)
- N M N P d-{ f . -
- Identify All Safety Systems & How Analyzed Code Verification for All Codes Used
- - - - -. .,e..- - _ ,_ _ , _ _ . _.
~
sheog-T_. 9C lo Mkb@-
- maw
-4 W.%iess
, mot eq. b 4chp GCM
- gw3 - Status '. D " *P* E h D
- ABS lU Ns.L4. + 5Ess est Surry Using SSI . . Mog c..9 -mW M Beaver Valley Using SSI l- b _ __ _ __
}# Nep Maine Yankee Complete Except for SHOCK I44 Sof 3,
~
and Code Verification uAtk N tP P C - O L Swee.r "I. caus , D. Expected Completion A Few Days to a Week (Slide 5) V. RELATED ACTIVITIES A. Maine Area Earthquake April 17,1979 Fet4- Iso we. a.a.
- Epicenter # 10 Km Magnitude 4.0 Richter ST siu u. FM3 b7 " '*'d A 30 Aftershocks ,
S- bad MMM ovc 36hQ c,.. A h . s . L r Ep%4va.Q. Intensity V (Still Under Review) N= W t.Hu amu 4 g , T' "^ H ' / - Seismographs at Plant Mm % pq. po y,s }' One 00C R4 g, - Other Didn't Trip on bh MO (<0.01 Vertical at Site) 4 r. % W - Li. A g , @ M k.4 . 4. .' . C4 pu.k h B. New Hampshire Earthquake
- April n ,1979
-99 2 9 tuhqE
- P_+ to- i s 4
. - saa.. A7.1,.ul - d o e t , < 4 ,
c,sAAc s a, Eiwp -K4 .h H '*f* % do W 1 $ P" ' # "- - k eude 1.) NA5% ra yta. H ok. - can re v.p q mm
- b. SSMRP tcP ..
..._.a_, . . .
E, tG
- , L.,caoluVt..f,aEp A -6Lr eO Ed.E.e4Gve pa %
~
- Sy(J 5'wlmeof4 SM hw M. cok ,ak pr z.,. wa (r% - %s:y')
r.) dew WAR . . . --. t s #c. % f, (S1ide 6) VI. MIT'CELLANEOUS OTHER ACT VITIE A. Velan Valves
- DLC Problem Originated with Valve Weight Problem (g e's.w. 2.)
IE Bulletin 79-04 (All ors) (3-30-79)
- Responses due 5-01-79 (30 Days) 1
- Mostly in Small Piping
,7 Bounding Effect o
y,5/.
.. * -m
,. - - 6" Line, if 100% weight increase; 30% stress
..e
, q l'i' ' increase
' ' '/* '*" -
Supports Generally Near Valves c gr uer % Oguk pe te-- lide 7) - B. Base Plates IE Bulletin 79-02 (All ors) (3-08-79)
- Response due 7-6-79 (120 Days)
- Owners Group Meeting (4-26) - D;;..;g
- Flexing of Base peekh8cs Concrete Anchors QA Check i
&~ Check, if no QA
- For sample, test with
-- Torque
--Pull (Li]ft) Test l
l I
.-m-w~r. - , - , _- - ,- , - - . , , , , - -
, w m ,._,, , - - ,
~ (Slide 8) C. Alegbraic Sum Method
- IE Bulletin 79-07
- General Search for Algebraic Sum If Alg. Sum Used.
Identify Codes and Systems Analyzed Provide Safety Bases For All Codes Used Provide Verification Means Response due 4-24-74 (llpNM) ) k allt f-tah (Slide 9) D. Status with Alg. Sum to Date
- Used Widely in Early 1970's
- Locations Identified to Date
. SHOCK 2
. WESTDYN
. DAPS
- ADLPIPE (Slide 10) E. Criteria -
- Since a Number of Uses Found Developing Disciplined Criteria to Decide on Next Step
- Operative Part is 50% of Allowable for SSE
- 50% not Fixed - Under Development
- Used for Discussion Purposes O
~
*d M-
(Slide 11) - Uses Two Typical Pipe Stresses
- If These are as Previously Designed For 6". Note SSE >50% Allowable For 30", Note SSE 450% Allowable (S1ide 12) -
" Allowabl es" Change e
= "*
- e a
C. 3/13/79 SHOW CAUSE ORDER u O 1.1: Is 1 5 PLANTS
-+' . .u d.oh .v.w.m :-
' C ALGEBRAIC SUMMATION RELATED" ACTIVITIES CWUv'ocap S4.3 WFW
"^~.
CODE VERIFICATION REVIEW HAND-CA.LCULATION METHODS
- -VakuSun. tac-t.-
.f iv AA- ~ OTHER ACTIVITIES P qu i-
-cu ch.a_ k too of -
IE BULLETIN 79-02 (3/8/79) e,ig,gobec,
-T= dub .q'M a c,2 m.go BASE PLATES AND CONCRETE ANCHORS DEmb to-S lo so-+ RESPONSE DUE 7/6/79 b TM- -
IE BULLETIN 79-04 (3/30/79) VELAN VALVE WEIGHTS . RESPONSE DUE 5/1/79 IEBULLETIN79-07(4/i4/79)
! SURVEY OF USES OF ALGEBRAIC
'! SUMMATION: RESPONSE DUE DATE 4/24/79 IE INVESTIGATION t
ORDER TO SHOW CAUSE I
~
e WHY THE LICENSEE SHOULD NOT REANALYZE THE FACILITY l PIPING' SYSTEMS FOR SEISMIC LOADS USING ANJPPROPRIATE PIPING CODE; f e l ]; WHY THE LICENSEE SHOULD NOT MAKE ANY NECESSARY MODI-I FICATIONS FOLLOWING REANALYSIS; k l:
- WHY FACILITY OPERATION SHOULD NOT BE SUSPENDED.PENDING SUCH REANALYSIS AND COMPLETION OF ANY REQUIRED MODI-FICATIONS r
)
I i r% s ' ' 4 - PIPING REANALYSIS STATUS REPORT AS OF 4/24/79 i
! MY BV F S-1 S-2 i
i i 0.NAby!]!_Jg_Dg_fP PEJ [._UNS}_____ _ _ _ hf_,,_ _ _ f_2 __ ___pp__,,___pp__ _ _ _ _ _ _ _ , . tW ** h?;> COMPLETED WITHIN ALLOWABLE 19 s 12 8
! uc- r\y gp '. ; p (QA ACCEPTED RESULTS) 2./
t.. - s 3 **
!. COMPLETED ABOVE ALLOWABLE O- ..0 0 0 i
l (HARDWARE CHANGE REQUIRED) L'S ' l PIPINGSUPPORTSTbEVALUATE 2 729 1156 673
- .s COMPLETED WITHIN ORIGINAL 0,T d'.15 151 20 DESIGN (QA ACCEPTED RESULTS)
COMPLETED ABOVE ORIGINAL 2 2 0 0 DE51bN (MANDWARE CHANGE R5 QUIRED) LICENSEE ESTIMATED SUBMITTAL 4/14 5/6 6/20 5/28
.s .?
f ad .p v .s . , t#' yp, y~- .
.. 9 .s V
. ;;. ~ ,
. (uW' 3.
. ~ . . - . _ . ..
rm. .., INFORMATION REQUESTED BY NRR LETTER OF I4-2-79
- FOR COMPUTER CODES USED; VERIFY WHERE ALG. SUM USED PROVIDE CODE LISTING
- FOR HAND CALC METHODS DESCRIBE / JUSTIFY METHODS
- )
- PROVIDE STATUS OF RESPONSE TO IEB 79-02 IDENTIFY ALL SAFETY SYSTEMS AND ANALYSIS METHODS
- FOR CODES USED FOR PREVIOUS EVAL. OR REANALYSIS PROVIDE INFORMATION ON VERIFICATION 2 %t N
l[Q , sg* CO'O / amme-DK
~
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s '. h,e A
-en f Ve.u
! c/ >y <- j 1cf .
l -
'i O EARTNQUhkE"fN"MAYNE(dI4-17-79Y I ;
- EPICENTER 10KM (6 MILES) i FELT 150 MI AWAY l
.i ' ! , MAGNITUDE 14 ,' O (30 AFTER SH0CKS)
EPICENTRAL INTENSITY V (STILL UNDER REVIEW) l
!
- DETECTORS AT PLANT i.
I - ONE 00C
+
0
! - OTHER DID NOT TRIP l
< l (LESS THAN 0.01NERTICAL AT SITE) e LITTLE SIGNIFICANCE TO REGIONAL ANALYSIS . UNLESS ASSOCIATED WITH SPECIFIC FAULT OR STRUCTURE I i e RE-EMPHASIZED BELIEF THAT FELT EARTHOUAKES
; CAN M!D D0 OCCUR IN AREA i
s M E.
l . 'n, r i: IE BULLETIN 79-OLI i r
}
VALVE WEIGHT PROBLEM [ i ' i { CONSERVATIVE HAND CALCULATION OF t 6 INCH PIPE WITH 100% INCREASE IN h, VALVE WEIGHT q.> ~ APPROXIMATELY 30% INCREASE IN STRESS SUPPORTS USUALLY NEAR VALVES SIMILAR RESULTS FOR OTHER SIZES . i I I
)
I
i f i . lD .. IE BULLETIN 79-02
.-
- MEETING WITH OWNERS GROUP ON 4/26/79 I
.[ e CLARIFY BULLETIN WITH RESPECT TO MINIMUM ! REQUIREMENTS f
i 1. FLEXIBILITY OF BASEPLATES .
~ !
- USE STAFF 2:1 CRITERIA;
- JUSTIFY RIGID BY ANALYSIS; OR, i
- RECALCULATE BOLT LOADS c.
') 2. DESIGN' MARGINS -
f.-
- USE MARGINS SPECIFIED OR l
- JUSTIFY LOWER SAFETY FACTOR BY DYNAMIC
'I '
TEST DATA
?:
i 3. IDENTIFY DESIGN REQUIREMENTS WHICH ADDRESS t ' - i VIBRATORY LOADING I:
!: 4. TEST REQUIREMENTS
[ .
- LIFT-0FF TEST
- TORQUE TEST f
' ULTRASONIC EXAMINATION y 8 SAMPLE SIZE l.
O-IE BULLETIN 79-07 FOR ALL ors
.
- IDENTIFY COMPUTER CODES WHICH USED ALGEBRAIC SUMMATION AND AFFECTED SAFETY SYSTEMS e PROVIDE COMPUTER CODE LISTINGS IF ALG, SUM USED
.
- VERIFICATION OF ' COMPUTER CODES USED -
- IF ALG. SUM, USED EST'IMATE CAPABILITY [0 WITHSTAND SEISMIC EVENT p#
.)
r t
,\
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-*-- -- -w-+---,--,-,e.,-gs ,,a- e e,,,.m n . ,---y-,,--,--- - m,vm,w, g-n, - - -w------ - - -~-----,- -
g- e ,.,,g
--.,,--m-
c . . j C' . l CODES USING ALGEBRAIC SUMMATION PLANT / CODE REANALYSIS COMMENT I SH0CK 2 MAINE YANKEE YES, COMPLETE UNDER REVIEW I' BEAVER VALLEY FITZPATRICK IN PROGRESS IdPROGRESS i SURRY 1 & 2 IN PROGRESS POINT BEACH * . IN PROGRESS RADWASTE
-f
't i ADLPIPE C$1.'.:;t... . ' b>
+' RHR, MS
! GINNA, ,
IN PROGRESS 3 (y W f
!- WESTDYN ! TURKEY POINT 3 8 4** YES, COMPLETE RCS YES, COMPLETE RCS ! H. B. ROBINSON **
D. C. COOK 1* IN PROGRESS i' DAPS YES, COMPLETE RECIRC,MS PILGRIM *
!; BRUNSWICK 1, 2* YES, COMPLETE RECIRC f
p BASED UPON TELEPHONE REPORT BASED UPON MEETING WITH NRC STAFF 4/13/79 l . t
[ . (- . 1 DECISION CRITERIA FOR REVIEW 0F IE BULLETIN 79-07
- 1. EXAMINE METHODS USED FOR PIPING DYNAMIC ANALYSIS:
DETERMINE IF ALGEBRAIC SUMMATION WAS USED AND HAS NOT
' BEEN REANALYZED *
- L #" T E N h .
-sp Wyuu n A& c. rwa,:
l .- _. 5 v
- 2. EVALUATE FUNCTION OF EACH SYSTEM AND PERFORM SAFETY EVALUATION TO DETERMINE IF OPERATION IS ACCEPTABLE
~
DURING REANALYSIS. I : .. . . . . . ,
+' +
COMPLETE SHUTDOWN UNTIL REANALYSI'S IN REANALYSIS COMPLETE 120 DAYS, AND NECESSARY
}
I - MODIFICATIONS IMPLE-I MENTED. l
$ sSt fra.Elt le r.',#cA*L*Mt d'h' b & -f( x 5":% Ci 62 re' A
# X.'), , '
'O. l l . :
l
c . l~ [ . PIPE STRESS'(. ILLUSTRATIVE EXAMPLE) LOAD 6 INCH PIPE 30 INCH' PIPE
~
DEADWEIGHT (D) 2j000 1,500 PRESSURE (P) 5,'000 3!000 SSE .1'6'!000 9,000 l i TOTAL (D+P+SSE) 23',000 13,500 e
't <7 ALLOWABLE (SSE) 27,000 27,000
[ . .
- OBE 11,000 6,000
~
TOTAL (D+P+0BE) 18l000 10,500 ALLOWABLE (0BE) 18,000 18,000 LOCA N/A 13,500 TOTAL (D+P+SSE+LOCA) N/A 27,000 ALLOWABLE (SSE) N/A 27,000 l L. l'
- r-i h
I PIPE-STRESS' ALLOWABLE SSE OBE i .. PRE-1977 1.8 Sg 1.2 4 POST-1977 2.4 S g 1.2 S-H
= 5/8 YIELD OR 1/4 ULTIMATE e
l 6 9 r i
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2.
* ~- - - n o- e
, _ . _ - . - - - - - - - - . - . , _ _ . . . . _ _ . _ _ . . . _ . _ . _ . . _ _ - . ...---.._..,___.-_...,__._....--,,,_.m____.__, - - . _ _ . _ . - _ _ , , _ . - ~ . , _ _ .
SEISMIC DESIGN AT MalNE YANKEE pt d
- PRESENT DESIGN - HOUSNER SPECTRO ANCHO..ED AT 0.le I
e IFLICENSEDTODAY,REGULATORYGUIDk1.60 SPECTRA i MIGHT BE ANCHORED AT 0.13e TO 0.20e
. PROBABILISTIC ESTIMATES OF EARTHOUAKE HAZARD
- GREAT CAUTION MUST BE EXERCISED IN USING THESE .
IN DECISION MAKING j - RELATIVE ESTIMATES MORE RELIABLE THAN ABSOLUTE - ESTIMATES
- e INITIAL ESTIMATES OF RELATIVE EARTHQUAKE HAZARD
- DIFFERENCE BETWEEN CHANCES OF:
EXCEEDING PEAKS OF 0.1 VS 0.2e ABOUT A FACTOR OF 5
- DIFFERENCE BETWEEN CHANCES OF:
~
EXCEEDING EXISTING DESIGN VS. REG. GUIDE 1.60 ANCHORED AT 0.2e ABOUT A FACTOR OF 2
- ESTIMATES BASED UPON EXTRAPOLATION OF PREVIOUS S IN OTHER AREA.
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PIPE: 6" SCll. 160, INSULATED, CONIAINING WATER t = 17', B31.1.0 SUGGESTED MAX. SPAN ! LENGTH: VALVE % INCREASE FUNDAMENTAL SEISMIC ! WEIGilT IN WEIGilT EREQllENCL_ SIEESS f a 225 - l 259 15 .988f 1.05o 1.33o j 115 0 100 .878f
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WEIGHT IN WEIGHT FREQUENCY STRESS i l a j f 225 - 259 15 .995f 1.03flo : i 450 100- .9fi5r 1.22a l l ( ' i
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o 3/9/79 CONTACTED FITZPATRICK, MAINE YANKEE AND SURRY 1, 2 fj!' REQUESTING BY 3/12/79: . ii!!il .'. fili
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IDENTIFY AFFECTED SYSTEMSJ SCHEDULE FOR REANALYSISJ BASES FOR
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- 1 REGULATORY GU DE 9, , DIRECTORATE OF REGULATORY STANDARDS REGULATORY GUIDE 1.48 DESIGN LIMITS AND LOADING COMBINATIONS FOR SEISMIC CATEGORY l FLUID SYSTEM COMPONENTS M "'
i @A A. INTRODUCTION loading combinations for design or for identifying Seismic Category I fluid system components. The lack of Q
- h. .-M/
General Design Criterion 2, " Design Bases for adequate guidance for selecting loading combinations is Protection Against Natural Phenomena,"of Appendix A apparent from a review of recent construction permit to 10 CFR Part 50." General Design Criteria for Nuclear applications which reflect design requirements as Power Plants," requires,in part, that the design bases for contained in the code design specifications. For structures, systems, and components important to safety essentially identical components designed for the same reflect appropriate combinations of the effects of plant conditions (i.e., operating conditions of the plant normal and accident conditions with the effects of categorized as normal, upset, emergency, and faulted natural phenomena such as earthquakes. This guide plant conditions) and specified seismic events (i.e., delineates acceptable design limits and anernnriate one half the Safe Shutdown Earthquake (SSE) and the combinations ut loadme mocu ted with normal SSE) the loading combinations and associated design operation, postulated accidents, anit spectf red seisnyp limits vary considerably among applications for events for the dwen .4 sentmc t_'atecorv 1 timd system construction permits. Regulatory Guides 1.26 and 1.29 components (i.e., water- and steam <ontaining (Safety Guide 26 and 29) entitled " Quality Group components). This guide applies to light watercooled Classifications and Standards" and " Seismic Design rea c tors. The Advisory Committee on Reactor Classification," respectively, provide acceptable bases for Safeguards has been consulted conceming thisguide and classifying fluid system components in relation to has concurred in the regulatory position, applicable national codes (e.g., Section ill of the ASME - Code) and for identifying those structures, systems and B. DISCUSSION . components that should be designed to remain functlpnal under the effects of the SSE (i.e., Seismic The design conditions and functional requirements Category I structures, systems, and components). of fluid system components important to safety in nuclear power plants should be reflected in the To further provide a consistent basis for design of application of appropriate design limits (e.g., stress or fluid system components important to safety, this guide strain limits) for the most adverse combination of delineates acceptable design limits and appropriate loadings to which these components may be subjected in combinations of loadings associated with applicable service, plant conditions and specified seismic events. The approach set forth in this guide is directiv tabtei to For components that are constructed in accordance . Section ill of he Wf F Code. Design hmits as spect!)ed with Section !!! of the American Sostety of Mechanical in secuon lli are extensnely utillied to provide Engineers (ASME) Ikuler and Pressure Vessel Code, assurance of the pressure retairung integrity of vessels, provision of a design specification which stipulates the pipmg. non active pumps, and non active valves of each de3n requirements for the component (i.e., the Code class; however, for the particular case of actim i mechanical and operational loadings) and the Code rium s and valves (i.e., pumps and valves that must classification of the component (e.g., Code C!n. I,2, or ~pertorm a meciianical motion during the course of
- 3) is required. Ilowever, neither Section til not any accomplishing a system safety function), special design cther pohtished n."~ui code or standard provides limits and supplemental requirementa are m. aiico tu adm" *, guidance for selecting code classifleations and primd., en us of operehty. ihese special deugn usAsc asoutnonf ouross eg. eye ,,
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! [ Standard or alternative design rules for Code Class 1 Active ASME Code Class 1 Velves (Designed by I , . , . ,
valves are specified by NB-3512 and NB-3513 of Section Standard or Alternative Design Rules) j ) 111 of the ASME Code.These design rules encompass the
; \
use of pressure temperature ratings of valves.The design To provide greater assurance of operability, the limits specified in this guide are in terms of P, which primary pressure rating (P,) for Code Class I active
. differs from the definition given by Section ill in that P r valves designed by standard or alternative rules should is related to maximum transient temperature in lieu of not be exceeded when the valve is subjected to the the design temperature. Pr is defined in this guide as the combination of loadings delineated in regulatory primary pressure rating corresponding to the maximum positions 5.a41),5.a42), and 5.a.(3).This design limit is j transient temperature for each plant condition as selected on the same basis as that designated for active j specified in Tables NB 35311 to NB 35317 of Section pumps and valves that are designed by analysis and is
- Ill. Therefore, the maximum transient temperature for analogous to design limits specified for the normal each plant condition should be determined before the operating condition category of Section ill of the ASME pressure rating of the valve is selected (e.g., Class 600, Code. Note 6 to the regulatory position also applies.
] 900, or 1500). In order to provide assurance of However, in the case of pressure.related valves, Note 6 pressure retaining integrity, Pr should not be exceeded states that the primary pressure ratings (P r ) I0' l by more than 10,20, and 50 percent when the valve is non active valves designed by standard or alternative ! subjected to the combination of loadings delineated in design rules may be used for the applicable loading I
regulatory positions 3.a.,3.b., and 3.c., respectively. One combinations if appropriate testing demonstrates that
. hundred ten percent and 120 percent of Pr, respectively, operability is not impaired when the valve is so rated.
are analogous to the upset and emergency operating Since detailed analytical techniques are not used to condition category limits of NB-3200 of Section !!!.One design pressurc temperature rated valves, demonstration hundred fifty percent of Pr is analogous to the of operability by test is indicated. { hydrostatic test pressure specified for Code Class I valves in Section 111. ASME Code Class 2 and 3 Components Active ASME Code Class 1 Pumps and Valves (Desiped With one exception, no distinction is made between by Analysist Code Class 2 and 3 components since the design requirements of Section 111 of the ASME Code are the The normal operating condition category design same for both classes of components. The design rules _' T limits given by NB 3222 of Section ill should be applied for Code Class 2 and 3 components do not provide for (' ") to design active pumps and valves for the combination of loadings delineated in regulatory positions 4.a.(l), design by analysis (except for Code Class 2 vessels designed in accordance with Section Vill Division 2, of 4.a42), and 4.a43). The design limits of NB 3222 are the ASME Code) and do not yet provide any design rules j selected besause the primary stressintensities associated for pumps. Furthermore, no design limits for other than with thm,limist na in the clastic range and thus provide the normal plant condition are available (the one greater assurance of operabdity for pumps and valves exception to this is piping). Generally. Class 2 and 3 (i.e., less probability of unacceptable deformations that components are of somewhat lower quality as related to would impede or prevent operation) than the design material, fabrication, and nondestructive examination limits for the upset, emergency, and faulted operating requirements than Code Class I components. Because of condition categories of Sestion 111. Secondary effects less stringent design requirements and a lower quality (stresses and deformations) in components whose only level in comparison to Code Class I components, the function is pressure retention are not usually evaluated design limits selected for Ccde Class 2 and 3 non sctive for the loading combinations delineated in regulatory components are, on a comparable basis, lower for the position 4.a.12) and 4.a.(3), llowever, these effects combination of loadings associated with the emergency should be considered for active Class I pumps and valves and faulted plant conditions than for Code Class I so that unacceptable deformations do ngt result. Local non active components. The same considerations that effects (peak stresses) need not be evaluated for these apply to Code Class I active pumps and valves apply to loading combinations, in addition to compilance with Code Class 2 and 3 active pumps and valws. the design limits specilled, demonstration of operabdity - as outlined by Note ci to the regulatory position should ASME Code Class 2 and 3 Venels (Deelped to Division also be provided. Note 6 auggests appropriate testing, analysis, or combinations of those measures that should 1 W seem Vml be implemented to demonstrate the operabdity of active pumps and valves under all design loading combmations. To provide assurance of pe.e retaining integrity flowever, Note 6 states that the design limits for for Code Class 2 and 3 vessels,the allowab ,t'ess value non active pumps and valves designed by analysis may be S should not be exceeded by more than 10 pers fat used if assurance is provided by detailed stress and the combination of loadings delineated in regulatory deformation anslyses that operabdity is not impaired positions 6.a.(l), and 6.a.(2), and S should not be t , when designed to these limits. esceeded by mere than 50 percent for the combination 1.48 3
allowed by Note 11, if the design limits for non active b. Pr should not be exceeded by more than 20 ' O valves are used, appropriate testing should demonstrate percent when the component is subjected to the loadings operability in lieu of analysis since detailed analytical associated with the emergency plant condition. techniques are not a pp!!e d to design c. P, should not be exceeded by more than 50 pressure temperature rated valves. percent when the component is subjected to concurrent loadings associated with the normal plant condition, the C. REGULATORY POSITION vibratory motion of the SSE, and the dynamic system Seismic Category I Duid system components should
, be designed to withstand the following loading 4. Active ASME Code Class I pumps and valves
- that combinations within the design limits8 specified. are designed by analysis:
- a. The design limits' speciGed in NB 3222s As or
- l. ASME Code 8 Class I vessels and pining:
the ASME Code should not be exceeded when the
- a. The design limits specified in NB-3223 and component is subjected to cither (I) concurrent loadings NB 3654 of the ASME Code for vessels and piping. associated with either the normal plant condition or the respectively, should not be exceeded when the upset plant condition and the vibratory motion of 50 com ponent is subjected to concurrent loadings percent of the SSE, or (2) loadings associated with the associated with either the nnrm,f *nt condition or the g upset plant condition3 and the uhratory monon of 50 percent of tb We h' Anu n F c'M ' % U.
emergency plant condition, or (3) concurrent loadings associated with the normal plant condition, the Vibratory motion of th .SSE. 6 and the dynamic system
- b. The design limits specified in NB 3224 and loadings associated with the faulted plant condition.
NB 3655 of the ASME Code for vessels and piping.
, respectively, should not be exceeded when the 5. Active ASME Code Class I valves that are designed component is subjected to loadings associated with the by standard or alternative design rules: ; emergency plant condition. a. The primary pressure rating Pr ' should not be
- c. The design limits speciGed in NB 3225 and exceeded when the component is subjected to either (1)
NB 3656 of the ASME Code for vessels and piping, concurrent loadings associated with either the normal respectively, should not be exceeded when the plant condition or the upset plant condition and the
; _ component is subjected to concurrent loadings vibratory motion of 50 percent of the SSE, or (2)
{ associated with the norm 11 nimt enndition, th, loadings associated with the emergency plant condition, vihr,tnre mn' inn nf the ssF, and the dynamic svstem or (3) concurrent loadings associated with the normal loadmes attociated with the huited plant condition. plant condition, the vibratory motion of the SSE, and
' $$h' the dynamic system loadings associated with the faulted
- 2. Non active ASME Code Class I pumps and valves
- plant condition.
that are destened by analysis:
- a. The design limits specified in NB 3223 8 of the ASME Code should not be exceeded when the 6. ASME Code Class 2 and 3 vessels designed to co m ponent is subjected to concurrent loadings Division I of Section Vill of the ASME, Code:
associated with either the normal plant condition or the a. The allowable stress value S should not be upset plant condition and the vibratory motion of 50 exceeded by more than 10 percent when the component percent of the SSE. is subjected to either (1) concurrent loadings associated
- b. The design limits specified in NB 3224 of the with enthu the normal plant condition or the upset plant ASME Code should not be exceeded when the e ndition and the vibratory motion of 50 percent of the component is subjected to loadings assoetated with the SSE. or (2) loadings associated with the emergency plant emergency plant condition. c ndition.
- c. The design limits specified in NB 3225 of the b. S should not be exceeded by more than 50 ASME Code should not be exceeded when the percent when the component is subjected to concurrent component is subjected to concurrent loadings 1 adings associated with the normal plant condition,the associated with the normal p8am condition, the vibratory motion of the SSE, and the dynamic system vibratory motion of the SSE, and the dynamte system I adings associated with the faulted plant condition.
l loadings associated vth the faulted plant condition. l 7, ASME Code Class 2 vessels designed to Division 2 of
- 3. Non active ASME Code Class I valves that are Section Vill of the ASME Code:
designed t y standard or alternative design rules: a. The design hmits specified in NB 3223 of the
- a. The primary. pressure rating Pr should not be ASME Code should not be exceeded whm the exceeded by more than 10 percent when the component com ponent is subjected to concurrent loadings is subjected to concurrent loadings associated with either associated with either the normal plant condition or the the normal plant condition or the upset plant condition upset plant condition and the vibratory motion of 50 and the vibratory motion of 50 percer t of the SSE. percent of the SSE.
l.48 5 L
DEFINITIONS q' Active Pumps and Valves. Components that must and shutdown other than upset, emergency, or faulted perform a mechanical motion during the course of plant conditions. accomplishing a system safety function. Plant Conditions. Operating conditions of the plant Allowable Stress Value (S). As specified in Appendix I categorized as normal, upset, emergency, and faulted of Section ill of the ASME Boiler and Pressure Vessel plant condtions. Code.
; Design by analysis for Class 1 Pumps and Class i Valves. Primary Pressure Rating (Pg ). The primary pressure ; For Class I pumps, the design procedures specified in rating corresponding to the maximum transient i
NB-3200 of the ASME Boiler and Pressure VesselCode, temperature for each plant condition, as specified in Section III. For Class I valves, the requirements of Case Section ill of the ASME Boiler and Pressure Vessel 1552 of Interpretations of ASME Boiler and Pressure Code, Tables NB-3531 1 to NB-35317, for Code Class i Vessel Code. valves or as specified in NC-351I and ND-351: for Code Class 2 and 3 valves, respectively. Dynamic System 1.madings Associated with the ' Plant Condition. Refers to those dynamie loadings w us i
' Safe Shutdown Earthquake (SSE). That earthquake result from the occurrence of a nostuhted runture (e.g.,
complete severance or equivalent longitudmal break - which produces the vibratory ground motion for which area)of any reactor coolant pressure boundary ninine or structures. systems, and components important to safety of any other ntnine not a turt ni the certor contant are designed to remain functional, Seismic Category I. Those structures, systems. and 1 components that are designed to remain functionalif the l g Emergency Plant Condition. Those operating conditions SSE occurs' , which have a low probability of occurrence. Standard or Alternative Design Rules for Class ! Valves. Faulted Plant Condition Those operating conditions As specified in NB 3512 and ND 3513 of the ASME
, ' associated with extremely low probability postulated Boiler and Pressure Vessel Code, Section 111.
events. I s Upset Plant Condition. Those deviations from the Normal Plant Condition. Those operating conditions in normal plant condition which have a high probabtlity of , the course of system startup, operation, hot standby, occurrence. NOTES Arplies to all components (vessels, piping, pumps, and b. fuu-scale prototype testing. valves) that are relied upon to cope with the effects of specified c. reduced-scale prototype testing. plant conditions.
- d. detailed str:ss and deformation analyses emeludes experimental stress and deformation analyses).
8 Sectbn 111 of the Arneracin Society of Mechanical in the performance of tests or analyses to demonstrate Engineers Boiler and Pressure Vessel Code includmg the 1972 operability, the structural interaction of the entre auembly Winter Addenda thereto. (8 8. Valve *perator assembly and pump motor anembly) should be considered. If superposition of test results for other than the
' Identification of the specific transients or events to be combined loading condition is proposed, the applicabihty of considered under each plant condition will be addressed in a such a procedure should be demonstrated. The design limits for future regulatory guide, non-active pumps and valves designed by analysis may be used for the applicble loading combinations if assurance is provided
'The requstements of the Case 1552 (Interpretations of by dertiled stress and deformation analyses that ntserability is ASME Doiler and Pressure Vessel Code) should be met for all n impaired M.a. b.gm t.,
the:t um.... sk.!:c.rn . B.: stres of Code Class 1 valves designed by analysis. Pnmary pressure ratings P, for non-active valves designed by standard or alte native design rules may be used for the
'The provisions of ND 3411 and ND-3413 may be applied applicable loading combmations if appropriate testing for au stres of Code Class I pumps designed by analysis. demonstrates that operability is not impaired when the vahe is so tated.
'In addition to compliance with the design limits specified, assurance of operability under all design load;ng combinations should be provided by an appropriate combination of the ' Secondary effects (stresses and deformation,) should be fogowing suggested measures:
evaluated for the loading cnmbinations designated by regulatory
- a. In stu testirig (e.g., preoperstbnal testing after the positions 4.a.(2) and 4.4.(3). Local effects (peak stresses) need component is installed in the plant). not be considered for these loading combmations.
l.48 7
Pev,seon 1 7 gG' ( December 1973 19 k5U5 o
,4 ,es o' DIRECTORATE OF REGULATORY STANDARDS YbRY GUIDE REGULATORY GU. IDE 1.60 DESIGN RESPONSE SPECTRA FOR SEISMIC DESIGN OF NUCLEAR POWER PLANTS A. INTRODUCTION extensiw study has been described by Newmark and Blunie in references 1, 2, and 3. After reviewing these Criterion 2. " Design Bases for Protection Against referenced documents, the AEC Regulatory staff has Natural Phenomena," of Appendix A," General Design determined as acceptable the following procedure for Criteria for Nuclear Power Plants," to 10 CFR Part 50. deGning the Design Response Spectra representing the "Licensmg of Production and Utthzation . Facilities," effects of the vibratory motion of the SSE,1/2 the SSE, requires, in part, that nuclear power plant structures, and the Operating Basis Earthquake (OBE) on sites systems, and components imporIant to safety be underlain by either rock or soil deposits and covering all designed to withstand the effects of earthquakes. frequencies ofinterest. However, for unusually soft sites, Prop > sed Appendix A " Seismic and Geologic Siting modiScation to this procedure will be required.
Critena " to 10 CFR Part 100," Reactor Site Cnteria," would require, in pa rt, that the Safe Shutdown in this procedure, the conGgurations of the Earthquake (SSE) be denned by response spectra horizontal component Design Response Spectra for each au respmding to the expected maximum ground ~o f the two mutually perpendicular honzontal axes are a ccelerations. This guide describes a procedure shown m Figure I of this guide. These shapes agree with acceptable to the Alic Regulatory staff for deGning those developed by Newmark, Blume, and Kapur in iespmse spectra for the seismic design of nuclear power icference 1. In Figure I the base diagram consists of plants. The Adviory Committee on Reactor Safeguards three parts: the bottom line on the left part represents has been omsulted conceming this guide and has the maximum ground displacement, the bottom line on concurred in the regulatory position. the nght part represents the maximum acceleration, and
.the middle part depends on the maximum velocity.The B. DISCUSSION horizontal component Design Response Spectra m Figure I of this guide correspond to a maximum in order to approximate the intensity and thereby /rurizontal ground accelemtion of 1.0 g. The ruumum estimate the maximum ground acceleration of the 8 ground displacement is taken proportional to the expected strongest ground motion (SSE) for a given site, maximum ground acceleration, apd is set at 36 inches proposed Appendix A to 10 CFR Part 100 spec Ses a for a ground acceleration of 1.0 g.The numencal vslues number of required investigations. It does not,however, of design displacements, velocities, and accelerations for pve a method for defining the response spectra l the horizontal component Design Response Spectra are conesponding to the expected maximum ground obtamed by multiplying the correspondmg values of the accelera non. maximum ground displacement and acceleration by the factors given in Table I of this guide. The displacement The recorded ground accelerations and response region knes of the Design Response Spectra are parallel spectra of past eartliquakes provide a basis for the to the maximum ground displacement hne and are rational design of structures to resist earthquAes. The shown on the left of Figure 1. The velocity region hnes Design Response Spectra.' specified for cesign purposes, slope downward from a frequency of 0.25 eps (co,ttrol can be developed statistically from response spectra of Point D; to a frequency of 2.5 cps (control point C)and past strong-motion earthc.uakes (see reference 1). An are shown at the top. The remainmg two sets of hnes between the frequencies of 2.5 cps and 33 eps (control
'Se stinsuons i the end of the guide. point A). with a break at a frequency of 9 eps (control
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. DEFINITIONS
%- ; --- Spectrum means a plot of the maximum relationship obtamed by analyzing. evaluating, and response (acceleration, velocity, or displacement) of a statistically combining a number ofindividual response 4 Tamily of idealized single. degree <>f. freedom damped spectra derived from the records of significant past osc Hators as a function of natural frequencies (or earthquakes.
periods) of the oscillators to a specified vibratory motion inout at their supports. When obtamed from a Mozamam (peak) Grour.d Actaleration specifled for a recorded arthquake record, the response spectrum given site means that value of the acceleration which
, tends to be irregular, with a number of peaks and corresponds to aro period in the design response spectra valleys. for that site. At zero period the design response spectra acceleration is identical for all damping values and is equal to the maximum (peak) ground. acceleratior.
Desip Response Spectrum is a relatively smooth specified foe that site. l f l r TABLEI
. HORIZONTAL DESIGN RESPONSE SPECTRA RELATIVE VALUES OF SPECTRUM AMPLIFICATION FACTORS FOR CONTROL POINTS Arrgilification Factors for Control Points g Accelerationss Despinasemtaa "8 A(33sumi B(9 cos) Cl2.5 comi D00.25 asal O.5 1.0 4.96 5.95 3.20 2.0 1.0 3.54 4.25 2.50 1 5.0 1.0 2.61 3.13 2.05
! 7.0 1.0 2.2". 2.72 1.88 10.0 1.0 1.90 2.28 1.70
' Maximum ground dispisament is taken proportional to mamansen ground acceleration, and is 36 in. for ground amelerasson of 1.0 gravity.
s Acceleration and displacement ampufication facsors are taken front recomunendations given in reference t.
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O.1 0.2 0.5 1 2 5 10 20 50 100 FRr IUENCY, ces FIGURE 1. HORIZONTAL DESIGN RESPONSE SPECTRA - SCALED TO 1g HORIZONTAL GTIOUND ACCELERATION 9 . _ _ _ _ , _ _ - - . , . - - - - . - -- -- - - - - ~~ ~~ ' ~~~'~~~
1 1 I Revision 1 ' p aeew U.S. NUCLEAR REGULATORY COMMISSION October 1978 l I 4' o, P,@, REGULATORY GUIDE
*.... OFFICE OF STANDARDS DEVELOPMENT i
Regulatory Guide 1.130 SERVICE LIMITS AND LOADING COMBINATIONS FOR CLASS 1 PLATE-AND-SHELL-TYPE COMPONENT SUPPORTS A. INTRODUCTION plants because they retain components in place during General Design Criterion 2, ,' Design Bases for loadings associated with normal and upset plant con-Protection Agamst Natural Phenomena,, of Appen- ditions under the stress of specified seismic events, dix A, " General Design Criteria for Nuclear Power thereby permiiting system components to function Plants," to 10 CFR Part 50, " Domestic Licensing of properly. They also prevent excessive component Production and Utilization Facilities, , requires that movement during the loadings associated with emer-the design bases for structures, systems, and compo- gency and faulted plant conditions combined with a nents important to safety reflect appropriate combi- specified seismic event or other natural phenomena, nations of the effects of normal and accident conds- thereby helping to mitigate system damage. Compo-tions with the effects of natural phenomena such as nent suppons are deformation-sensitive because large carthquakes. The failure of members designed t deformations in component supports may signifi-suppon safety-related components could jeopardize cantly change the stress distribution in th'e suppon the ability of the supported component to perform its system and its components. safety function. In order to provide a consistent level of safety, the This guide delineates acceptable levels of service ASME Boiler and Pressure Vessel Code classifica-
- . limits and appropriate combinations of loadings as. tion for component supports should, as a minimum, sociated with normal operation, postulated accidents, be the same as that of the supported components. " and specified scismic events for the design of Class 1 This guide delineates levels of service limits and plate-and-shell-type component supports as defined loading combinations, as well as supplementary in Subsection NF of Section III of the American So. criteria, for Class I plate-and-shell-type component ciety of Mechanical Engineers (ASME) Boiler and supports as defined by NF-1212 of Section III of the Pres'sure Vesses Code. This guide applies to light. Code. Snubbers are not addrassed in this guide.
water-cooled reactors. Subsection NF of Section III permits the use of The Advisory Committee on Reactor Safeguards three methods for the design of Class I plate-and-has been consulted concerning this guide and has shell-type component supports: (1) linear elastic concurred in the regulatory posinon. analysis, (2) load rating, and (3) experimental stress analysis. For each method, the ASME Code de-B. DISCUSSION lineates allowable stress or loading limits for various Load-bearing members classified as component Code service levels, as defined by NF-3113 and gg'(Q NCA-2142.2(b) of Section III, so that these n,u - supports are essential to the safety of nuclear power g can be used in conjunction with the resultant loadings .i
- Lines indicate substantive changes from previous issue. or stresses from the appropriate plant conditions. @YI
, 8 American Society of Mechanical Engineers Boiler and Pressure Since the Code does not specify loading combina-Vessel Codes Section !!!. Divtsion I,1977 Edition, including the tions, guidance is needed to provide a consistent 1977 Winter Addenda thereto. Copies of the Code may be ob- basis for the design of component supports.
trined from the American Society of Mechanical Engineers. United Engineenng Center, 345 East 47th Street. New York. Component supports considered in this guide are N.Y. 10017. located within Seismic Category I structures and are USNRC REGULATORY OUIDES comments snuuid be sent to tre sec.etary o' trw comin.as.on. U S Nucear ae .,, c-es are - to de.Cnbe .r,d m e .vs m ine ub P"L9*"- " *""*"a- " ""'- * """* memods accectabie m the NRC start of empemenong speedc parts of me commessen s reguianons. m omneste techn.auen used av ene staff m eveau. The g-es are asued a the foilowes ten broad dnnsas: Gu de tJtu es or fogula as ynd Co 1. Power aceCtors
-C. aeguiaeer . not ,0.s..d. ue~s - so.- od,e.eet ror, mose z a-ch .ad rest a- 6. Products .=n 7 T-so-t set out in the guides ata be acceoreone d they prov.de a bases for the fadirtgs 3. Fuses and Matence Fecdotes 8 oCCuDatior al Heafth the esuance or Con 0nvence of a per*mt ar hcenes by the 4E nd rust and Fmanc.at Rowew
.J c - ts- -o.e.t_s, - em.m.m - se oo.e.,.o, - 4,0 u,ag.d.: til times and guedes .H be reveed. as appr*lloriste, to SCCommodate Comments
. esta o.n
-em.nt . .s ma dee_.t .or s _h m., be so ,com, .o.uced,
.e g-.sor ,or si sostdiC diviesons should be made in wntag to the d s. Nuclear Regulatory a bats oc pu addt sta f Ird' and Document Control hl dh NY
tion of ECCS during faulted plant conditions) will Service Limits. Stress limits for the design of com- ) operate properly regardless of plant condition, the ponent supports as defined by Subsection NF of Sec-n Code level A or B service limits of Subsection NF tion III.
/
(which are identical) or other justifiable limits pro- Specified Seismic Events. Operating Basis Earth-vided by the Code should be used. quake and Safe Shutdown Earthquake. 5 System Mechanical Loadings. The static and l
- 6. Deformation Limits dynamic loadings that are developed by the system l Since component supports are deformation- Operating parameters, including dead weight, pres-sensitive load-bearing elements, satisfying the serv- sure, and other external loadings, but excluding ef-ice limits of Section III will not automatically ensure fects resulting from constraints of free-end move-their proper function. Deformation limits, if specified ments and thermal and peak stresses.
by the Code Design Specification, may be the con- Ultimate Tensile Strength. Material property based trolling criterion. On the other hand, if the function on engineering stress-strain relationship. of a component support is not required for a particu-lar plant condition, the stresses or loads resulting UPfer Plant Condition. Those deviations from the from the loading combinations under the particular normal plant condition that have a high probability of plant condition do not need to satisfy the design lim. occurrence. its for the plant condition. C. REGULATORY POSITION
. SME Code Class I plate-and-shell type compo-Y. Definitions nent supports except snubbers, which are not ad-Critical Buckling Strength. The strength at which dressed in this guide, should be constructed to the lateral displacements start to develop simultaneously rules of Subsection NF of Section III of the Code, as with in-plane or axial deformations, supplemented by the following:3 Design Condition. The loading condition defined 1. The classification of component supports by NF 3112 of Section III of the ASME Boiler and should, as a minimum, be the same as that of the Pressure Vessel Code. supported components.
s Emergency Plant Condition. Those operating con. 2. Values of Su at temperature, when they are not ditions that have a low probability of occurrence. listed in Section III, should be estimated by
> ethod I, Method 2, or Method 3, as described Faulted Plant Condition. Those operating condi- below, on an interim basis until Section III includes tions associated with postulated events of extremely low probability. such values. Values of S, at temperature listed by Tables I-2.1, I-2.2, and I-13.1 of Appendix I and , Levels of Service Limits. Four levels (A, B, C, and Table 3 of the latest accepted version' of Code Case D) of service limits defined by Section III of the 1644 of Section III may be used for the interim Code for the design of loadings associated with dif. calculation.
ferent plant conditions for components and compo-
- a. Method 1. This method applies to component nent supports in nuclear power plants.
support materials whose values of ultimate tensile Normal Plant Condition. Those operating condi- strength at temperature have not been tabulated by tions in the course of system startup, operation, hot their manufacturers or are not available. standby, refueling, and shutdown other than upset, emergency, or faulted plant conditions. Su = S,3'-- l , Sn
- Operating Basis Earthquake (OBE). As defined in ,y,7, l Appendix A " Seismic and Geologic Siting Criteria for Nuclear Power Plants,, to 10 CFR Part 100, S = ultimate tensile strength at temperature t l ,
to be used to determine the design , , Reactor Site Criteria. limits Operating Condition Categories. Categories of de- S , = ultimate tensile strength at room tem- , sign limits for component supports as defined by perature tabulated in Section III, Ap-NF-3113 of Section III of the ASME Code, pendix I, or the latest acepeted version 8 Plant Conditions. Operatins conditions of the plant l categorized as normal, upset, emergency, and faulted 8 If the function of a component support is not required during a l ~~nlat conditions. plant condition, the design limits of the support for that plant con. ! dition need not be satisfied, provided excessive deflections or l Safe Shutdown Earthquake (SSE). As defined in failure of the suppon win not result in the loss of function of any Appendix A to 10 CFR Part 100. other safety.related system. 1.130-3 b 9 -
'e-. *"
c:mponent supports d:signtd by the linrar-clastic. condition should be design:d within the limits d:- analysis method. scribed in Regulatory Position 4 or other justifiable S b* The value of T*L* x 0 7 I should not be limits such as the level C or level D service limits 3, provided by the Code. These limits should be defined exceeded, where T.L. and S are defined according by the design specification so that the function of the to NF-3262.1 of Section III and S', is the ultimate Supported system will be maintained when the sup-tensile strength of the material at service temperature Ports are subjected to the loading combinations de-for component supports designed by the load-rating scribed in Regulatory Positions 5 and 6. D. IMPLEMENTATION
- c. The collapse load determined by 11-1400 and . .
divided by 1.1 should not be exceeded for component The purpose of th.is section is to provide guidance supports desigried by the experimental-stress-analysis to applicants and licensees regarding the NRC staff's method. Pl ans for using this regulatory guide.
- d. If plastic methods are used for the design of Except in those cases in which the applicant pro-component supports, the combined loadings of Reg. Poses an acceptable alternative method for complying ulatory Position 6 should include all loads such as with the specified portions of the Commission's reg-thermal loads and constraints of free displacements, ulations, the method desen, bed herein will be used in which contribute to expansion stress intensities, and the evaluation of submittals for construction permit the service limits of F-1324 and F-1370(c) of Sec- applications docketed after October 31.1978.1f an tion !!! should not be exceeded. applicant wishes to use this regulatory guide m, de-velop,ng i submittals for construction permit applica-
- 7. Component supports in systems whose normal tions docketed on or before October 31,1978, the function is to prevent or mitigate the consequences of pertinent portions of the application will be esaluated events associated with an emergency or faulted plant on the basis of this guide.
E d O 1 O C v 1.130-5
R .visi:n 1 [g urug'o U.S. NUCLEAR REGULATORY COMMISSION January 1978
~'
Qgf) ,,g f OFFICE OF STANDARDS DEVE DE REGULATORY GUIDE 1.124 e SERVICE LIMITS AND LOADING COMBINATIONS FOR CLASS 1 LINEAR-TYPE COMPONENT SUPPORTS A. INTRODUCTION with the specified seismic event, thus helping to General Design Criterion 2, " Design Bases for mitigate the consequences of system damage. Com-Protection Against Natural Phenomena," of Appen- Ponent supports are deformation sensitive because dix A, " General Design Criteria for Nuclear Power large deformations in them may s,gmficantly i change Plants," to 10 CFR Pan 50, " Licensing of Produc. the stress distribution in the support system and its tion and Utilization Facilities," requires that the de- supported components. sign bases for structures, systems, and components In order to provide uniform requirements for con-important to safety reflect appropriate combinations struction, the componen: supports should, as a of the effects of normal and accident conditions with minimum, have the same ASME Boiler and Pressure the effects of natural phenomena such as earthquakes. Vessel Code classification as that of the supported The failure of members designed to support safety- components. His guide delineates levels of service related components could jeopardize the ability of the limits and loading combinations, in addition to supported component to perform its safety function. supplementary criteria, for ASME Class I linear-type This guide delineates acceptable levels of service component supports as defined by NF-1213 of Sec-l limits and appropriate combinations of loadings as- are not addressed in this guide. tion Ill. Snubbers sociated with normal operation, postulated accidents, Subsection NF and Appendix XVII of Section ill and specified seismic events for the design of Class I permit the use of four methods for the design of Class linear-type component supports as defined in Subsec- I linear-type component supports: linear elastic anal-tion NF of Section III of the American Society of vsis. load ratine. experimental mA ,navw and Mechanical Engineers (ASME) Boiler and Pressure limit analysis. For each method, the ASME Code de-Vessel Code. This guide applies to light-water-cooled Tineates allowable stress or loading limits for various l remun. The Advisory Committee on Reactor Code levels of service limits as defined by NF-3113 Safeguards has been consulted concerning this guide of Section III so that these limits can be used in con-and has concurred in the regulatory position. junction with the resultant loadings or stresses from the appropriate plant conditions. Since the Code does _? B. DISCUSSION not specify loading combinations, guidance is re- h l Load-bearing members classified as component quired to provide a consistent basis for'the design of ' M " supports are essential to the safety of nuclear power component supports. l plants since they retain components in place during Component supports considered in this guide are the !c:dtngs asse:::ts, with normal and upset plant located within Seismic Category I structures and are conditions under the stress of specified seismic therefore protected against loadings from natural events, thereby permitting system components t phenomena or man-made hazards other than the spec-
~
function properly. Eey also prevent excessive com-
, ified seismic events. Thus only the specified seismic ponent movement during the loadings associated with , events need to be considered in combination with the emergency and faulted plant conditions combined loadings associated with plant conditions to develop
)
- Lines indicate substantive change from previous issue, appropriate loading combinations. Loadings caused l
USNRC REGULATORY GUIDES comm=s cooid tie sem e. ,he sweew, os the co,aan.on. u s Noc.e, am
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J lar plant condition, the stresses or loads resulting A complete and consistent design is possible only from the loading combinations under that plant condi-when system / component / component-support interac- tion do not need to satisfy the design limits for the o tion is properly considered. When all three are j evaluated on an clastic basis, the interaction is usu-ally valid because individual deformations are small. plant condition.
- 7. Definitions However, if plastic analysis methods are employed in the design process, large deformations that would re- Design Condition. He loading condition defined sult in substantially different stress distributions may by NF-3112 of Section Ill of the ASME Boiler and
***" Pressure Vessel Code.
When component supports are designed for load' Emergency Plant Condition. Those operating con-g
+ ings associated with the faulted plant conditions Ap-ditions that have a low probability of occurrence.'
pendix F of Section III perm,ts i the use of plastic Faulted Plant Condition. Those operating condi-analysis methods in certain acceptable combinations tions associated with postulated events of extremely ' for all three elements. Dese acceptable combinations are selected on the assumption that component sup- low probability. ports are more deformation sensitive (i.e., their de- Levels of Service La. mars. Four levels, A, B, C, and i formation in general will have a large effect on the D, of service limits defined by Section III for the de-stress distribution in the system and its components.) sign of loadings associand,with different plant condi-Since large deformations always affect the stress dis- nons for components and component supports in nu. tribution, care should be exercised even if the plastic clear Power plants. i* analysis method is used in the A,nendix F-approved Normal Plant Condition. Those operating condi-
- methodology combination. This is especially impor- tions in the course of system startup, operation, hot tant for identifying buckling or instability problems where the change of geometry should be taken into standby, refueling, and shutdown other than upset,
, emergency, or faulted plant conditions. ! account to avoid erroneous results. Operating Basis Earthquake (OBE). As defined in
- 5. Function of Supported System Appendix A to 10 CFR Part 100.
l In selecting the level of service limits for different Plant Conditions. Operating conditions of the plant loading combinations, the function of the supported categorized u normal, upset, emergency, and faulted system must be taken into account. To ensure that
' Pl ant conditions.
systems whose normal function is to prevent or miti-gate consequences of events associated with an emer- Safe Ntdown Earthquake (SSE). As defined in i ' gency or faulted plant condition (e.g., the function of A;,pendix A to 10 CFR Part 100. ECCS durmg faulted plant conditions) will operate Service Limits. Stress limits for the design of com-properly regardless of plant condition, the Code level ponent supports as defined by Subsection NF of Sec. 1 A or B service limits of Subsection NF (which are tion III. identical) or other justifiable limits provided by the Code should be used. Specified Seismic Events. Operating Basis Earth-quake and Safe Shutdown Earthquake. Since Appendix XVil derived all equations from AISC rules and many AISC compression equations , System Mechanical Lcadings. The static and have built.in constants based on mechanical prop- dynamic loadings that are developed by the system
*rties of steel at room temperature, to use these equa- operating parameters, including deadweight, pres-uons indiscriminately for all NF and the latest ac-sure, and other external loadings, but excluding ef-cepted version of Code Case 1644 materials at all fects resulting from constraints of free.end move-temperatures would not be prudent. For materials ments t.nd thermal and peak stresses, other than steel and working temperatures substan- Ult: mare Tensile Strength. Material property based tially different from room temperature, these equa- on engineering stress-strain relationship.
tions should be rederived with the appropriate mate-rial properties. Upser Plant Conditions. Rose deviations from the normal plant condition that have a high probability of
- 6. Deformation Limits occurrence.
Since component supports are deformation- C. REGU:.ATORY POSITION sensitive load. bearing elements, satisfying the serv-ice limits of Section III will not automatically ensure ASME Code8 Class I linear-type component sup. their proper function. Deformation limits, if specified by the Code Design Specification, may be the con
- 8 American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Section III. Division 1.1974 Edition, including the trolling criterion. On the other hand. if the function 1976 winter Addenda thereto.
of a component support is not required for a particu. N* 1.124-3
.r'Wt---" -w---r--w' + -t--- - -- -< =- -' =&-----w-= w r"-%'mr-tw=------v'------- w- '-"---"-------"-'------'---"v--= * '
or B servica limits provided by F-1370(a) for level D Section III divided by 1.7 should not be exceeded for service limits should be the smaller factor of 2 or component supports designed by the experimental
,. -'g .
1.167SJS,, if S. > 1.2S, or 1.4 if S.
- 1.2S,, stress analysis method.
i where S, and S, are component support material propernes at temperature. 6. Component supports subjected to the system mechanical loadings associated with the emergency However, all increases (i.e., those allowed by plant condition should be designed within the follow. NF-3231.l(a), XVII-2110(a), and F-1370(a)] ing design limits except when the normal function of should always be limited by XVII-2110(b) of Section the supported system is to prevent or mitigate the 111. The critical buckling strengths defined by consequences of events associated with the emer. XVII-2110(b) of Section III should be calculated gency plant condition (at which time Regulatory using material propenies at temperature. This in. Position 8 applies):* 8 cnase of level A or B service limits does not apply t , e limits for bolted connections. Any increase of limits a. He stress limits of XVII-2000 of Section III and Regulatory Positions 3 and 4, increased accord-for shear stresses above 1.5 times the Code level A . service limits should be justified. ing to the provisions of XVII-2110(a) of Section III
' and Regulatory Position 4 of this guide, should not If the increased service limit for stress range by be exceeded for component supports designed by the ' , NF-3231.1(a) is more than 2S, or S., it should be linear elastic analysis method.
limited to the smaller value of 2S, or S, unless it can . be justified by a shakedown analysis. b. The emergency condition load rating of NF-
; 5. Component supports subjected to the combined 3262.3 of Section Ill should not be exceeded for component supports designed by the load. rating loadings of system mechanical loadings associated method, with (1) either (a) the Code design condition or (b) ,
the normal or upset plant conditions and (2) the vib- c. He lower bound collapse load determined by ratory motion of the 0,BE should be designed within WM00 @M m% m h pih M the following limits. XVII-4110(a) of Section III should not be exceeded
' for component supports designed by the limit analysis
- a. The stress limits of XVII-2000 of Section 111 method. '
and Regulatory Position 3 of this guide should not be exceeded for component supports designed by the d. The collapse load determined by 11-1400 of
, linear elastic analysis method. These stress limits Section III divided by 1.3 should not be exceeded for t may be increased according to the provisions of component supports designed by the experimental
.k NF-3231.l(a) of Section III and Regulatory Position stnss analys.is m Wiod.
4 of this guide when effects resulting from constraints 7. Component supports subjected to the combined of free.end displacements are added to the loading loadings of (1) the system mechanical loadings as-combination. sociated with the normal plant condition, (2) the vib-
- b. The normal condition load rating or the upset rat ry m tion of the SSE, and (3) the dynamic system condition load rating of NF-3262.3 of Section III I adings associated with the faulted plant condition should not be exceeded for component supports de- should be designed within the following limits except 4 signed by the load. rating method. when the normal function of the supported system is .
to prevent or mitigate the consequences of events as-
- c. He lower bound collapse load determined by sociated with the faulted plant condition (at which XVII-4200 adjusted according to the provision of time Regulatory Position 8 applies):
i XVII-4110(a) of St.ctioa til should not be exceeded
~
- a. The stress limits of XVII-2000 of Section III r component supports designed by the hmir analysis
' and Regulatory Position 3 of this guide, increased ac-cording to the provisions of F-1370(a) of Section III l d. The collapse load determined by 11-1400 of and Regulatory Position 4 of this guide, should not l be exceeded for component supports designed by the i
- since component supports are deformatien sen:itiv. in the linear elastic analysis method.
performance of their service requirements. satisfying these entens does not ensure that their functional requirements will bc fulfilled. s,.. aller value of T.L. x 2S/Su or T.L. x Any deformation limits specified by the design specification may 0.7S.b. De not > meeeded, where T.L., S. and dSu should be controlling and should be satisfied. S, are defined according m NF-3262.1 of Seniort
- Since the design of component suppe.as is an integral part of the UI, and S'u is the minimum ulum.c WL & ngth
~
( design of the system and the design of the component. the de- of the material at service temperature for compC' signer must rrake sure that r ethods used for the analysis of the supports designed by the load-rating method. system. componeur, armi component support are compatible (see Table F-1322 2-1 in Appenda F of Section IID. Large deforma. c. De lower bound collapse load determined by
, tion. .ne system or components should be considered in the usn of component supports.
XVII-4200 adjusted according to the provision of F-1370(b) of Section 111 should not be exceeded fcr 1.124-5 j
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NE$UEI 65" GU DE OFFICE OF STANDARDS DEVELOPMENT REGULATORY GUIDE 1.92 COMBINING MODAL RESPONSES AND SPATIAL COMPONENTS IN SEISMIC RESPONSE ANALYSIS A. INTRODUCTION 2. Combining the mnimum vahres (in the case of time.historv dynamic analvsis) or the representative Cnterion 2, " Design Bases for Protection Against Tmaximum values (m the case ot~ spectrum dynamic Natural Phenomena," of Appendix A," General Design an alysM of the response or a given element of a Criteria for Nuclear Power Plants," to 10 CFR Part 50, structure, system, or component, when such values are "I.icensing of Production and Utilization Facilities," calculated independently for each of the three orthoes requires, in part, that nuclear power plant structures, nal soauai comnonents (two honzontal and one wrtical) systems, and components important to safety be de. of an earthquake. The combined value will be the siened to withstand the effects of earthounkes without_ representative maximum value of the combined response loss of capahtitty to perform their safetv 6m etions. of that element of the structure, system, or component Paragraph (a)(1) of Section VI, " Application to Engi. to simultaneous action of the three spatial components. neering Design," of Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 10 CFR The Advisory Committee on Reactor Safeguards has
- Part 100," Reactor Site Criteria," requires,in part,that been consulted conceming this guide and has concurred structures. systems. and components important to safety in the regulators position. m remain nunctional m the event of a Sate Shutdown MIb*
,L Earthquake ISSE). It specifies the use of a suitable dynamic analvsis as one method of ensuring that the B. DISCUSSION structures, systems, and components can withstand the keg
- 1. Combining Modal Responses seismic loads. Similarly, paragraph (a)(2) of Section VI 1 of the same appendix requires, in part, that the To find the values of the response of different T , t st. _. .. _
<....... . . components necessary for con- ,y, gg , g .
g, g tmued operation witnout unuuc m o e nr.M th and safety of the public remam functional in an Operating necessary to calculate the mode shapes and freauencies 7 (, Basis Earthauake tOBla. ag.uu, the uw or a suitable of the stmeture, system, or component. This is done oy g dynamic analysis is specified as one method of ensuring solving the following equation for the eigenvectors and that the structures, systems, and components can with-eigenvalues: stand the seismic loads. [K] - wn (MI (1) This gQ describes methods acceptable to the NRC f4nf=0 where (Kl is the stiffness matrix, wn is the natural frequency for the nth mode, [M] is the mass matrix, and
- 1. Combining the values of the response ofindividual
, modes in a response :,pectrum . nodal dynamic analysis to {4n} is the eigenwetor for the nth mode.
find the representative maximum value of a particula. response of interest for the design of a given element of a nuclear power plant structure, system, or component.
- Lines indicate substanuve cn g . f-<>m previous issue.
USNRC REGULATORY GU DES c.a.m.m. e.s. .. h. su,m.,, .e .h. c n. 2 . . ~ .. r..-c.........-.........
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, ;; ,,c --.- *- ..~ o e == a
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- 1. Combination of Modal Responses C mpon:nt should then be obtained by taking the Aquare root of the sum of the souares of correspondinq 1.1 With No Closely Spaced Modes representative maximum values of the response of the element attributed to each closely spaced group of In a response spectrum modal dynamic analysis,if the m des and the remaining modal responses for the modes modes are not closely spaced (two consecutive modes that are not closely spaced.
are defined as closely spaced if their frequencies differ Mathematically, this can be expressed as follows: from each other by 10 percent or less of the lower frequency), the representative maximum value of a N P j J
~%
particular response of interest for design (e.g., com- R= ponents of stress, strain, moment, shear, or displacu R[ + RpqRm9 tem (4) ment) of a given element of a nuclear power plait . k= 1 q=1 R=i m=i , structure, system, or component subjected to a single independent spatial component (response spectrum) of a whem Rgq and Rmq are modal responses, Re and R m three. component earthquake should be obtained by within the 9th group,respectively;iis the number of the taking the square root of the sum of the souares (SRSS) mode when a group starts;j is the number of the mode of corresponding maximum values of the response of the whem a group ends; R, Rk, and N are as defined element attributed toindividual significant modes of the previously in regulatory position 1.1 of this guide;and P structure, system, or component. Mathematically, this is the number of groups of closely spaced modes, can be expressed as follows: excluding individual separated modes.
-N -
g 1.2.2, Ten Percent Methoo R= R[ (3)
~
N "M _k=1 _ R= R[ + 2 Rj Rj i*j k=1 (5) where R is the. representative maximum value of a - - particular response of a given element to a Eiven component of an earthquake, Rkts the peak value of the where R, R g, and N are as delined previously in response of the element due to the kth mode, and N is regulatory position 1.1 of this guide. The second the number of significant modes considered in the modal summation is to be done on all i and j modes whose response combination. frequencies are closely spaced to each other. Let w and 3 wj be the frequencies of the ith and jth mode. In order 1.2 With Closely Spaced Modes to verify which of the modes are closely spaced, the following equation will apply: In a response spectrum modal dynamic analysis,if some or all of the modes are dosely spaced, any of the Wj - Wi (6) following regulatory positions (i.e., 1.2.1, 1.2.2, or w $ 0.1 1.2.3) may be used as a method acceptable to the NRC staff to combine the modal responses. also I$ i <j $ N (7); 1.2.1 Grouping Method 1.2.3 Double Sum Method Closely spaced modes should be divided into
- N N 5
groups that include all rnodes having frequencies lying between the lowest frequency'in the group and a R- - { RR k s eks (8) l frequency 10 , percent higher. The representative . k=1 s=1 . maximum value of a particular response ofinterest for the design of a given element of a nuclear power plant where R, Rk, and N are as defined previously in structure, system, or component attributed to each such regulatory position 1.1 of this guide. R 3is the peak value f the response of the element attributed to sth mode. l group sum of of themodes absolute should values of first be obtained bylltaking the ~ the correspondine vTIues of the response of the element attnhuted to_ neak
, g+
(4 gg e [ .i (9) indmdual modes m tnat croup. The representative maxunum value of tius particular response attributed to f Wk + Cs' Ws} j all the significant modes of the structure, system, or in which !
' Groups should be formed starung from the lowest frequency and working towards successively higher frequenc:cs. No one wf = Wk 1 - $( (10) frequente is to be in more than one group.
% M-_ _s R
v-
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gn p agg 4- $0$ V I'91'3 6t4*s- O- l*%
- m
REFERENCES ( h.
- 1. R. L. Wiegel, editor, Earthquake Engineering, 4. E. Rosenblueth and J. Elorduy, " Response of Englewood Cliffs, N.J., Prentice. Hall, Inc.,1970, chapter Unear Systems to Certain Transient Disturbances,"
by N. M. Newmark, p. 403. Proceedings. Fourth World Conference on Earthquake
" "# "#' ' *E ' #'
- 2. A. K. Singh, S. L Chu, and S. Singh," Influence of Closely Spaced Modes in Response Spectrum Method of Analysis," Proceedings of the Specialty Conference on 5. N. C. Tsai, A. H. Hadjian et al., " Seismic Analysis Stmetural Design of Nuclear Plant Facilities, Vol. 2, of Structures and Equipment for Nuclear Power Plants,"
Chicago, December 1973. (Published by American Bechtel Power Corporation Topical Report 4.A, Revi. Society of Civil Engineers, New York, New York.) sion 3, November 1974
- 3. S. L Chu, M. Amin, and S. Singh, " Spectral Treatment of Actions of Three Earthquake Components 6. C. Chen, "Def'mition of Statistically Independent on Structures," Nuclear Engineering and Design,1972, Time Histories," Journal of the Stmetural Division, Vol. 21, No 1, pp.126136. ASCE, February 1975.
M e e 1.92 5
JUNE 1973 (8'k -N5U5.IibRY
,, ,/ DIRECTORATE OF REGUULTORY STANDARDS GUIDE l REGULATORY GUIDE 1.57 DESIGN LIMITS AND LOADING COMBINATIONS FOR METAL PRIMARY REACTOR CONTAINMENT SYSTEM COMPONENTS A. INTRODUCTION functional under the effects of the Safe Shutdown l Earthquake [SSE]). This conclusion is supported by General Design Criterion 2 " Design Bases for B.1223.4(a) of Appendix B to Section 111. " Owner's Protection Against NaturalPhenomena,"of Appendix A Design Specification" which states, in part, "The to 10 CFR Part 50," General Design Criteria for Nuclear system's function, the environmental conditions under Power Plants," requires,in part, that the design bases for which these functions are performed, and the loading structures, systems, and components important to safety combinations must be evaluated from the system reflect appropriate combinations of the effects of standpoint. nis Section [III] does not provide guidance M%
normal and accident conditions with the effects of natural phenomena such as earthquakes. His guide in the identification of these system functions, conditions, and loading combinations." It is apparent { yi y' myy delineates acceptable design limits and appropriate from a review of recent applications for construction %id b combinations of loadings associated with normal permits in which ASME Code design specifications are operation, postulated accidents, and specified seismic reflected that adequate guidance for selecting loading events for the design of components of metal primary combinations is not presently available. For essentially reactor containment systems. This guide applies to identical components that perform a containment light. water cooled reactors. He Advisory Committee on function, the loading combinations and associated design Reactor Safeguards has been consulted concerning this limits are not consistent among different applications for guide and has concurred in the regulatory position. construction permits. However, components that perform a primary reactor containment function are N- B. DISCUSSION identified as Category I for seismic design purposes by Regulatory Guide 1.29 (Safety Guide 29), " Seismic ne design conditions and functional requirements Design Classification." of components which provide a pressure boundary for the primary reactor containment function should be To further provide a consistent basis for the design reflected in the application of appropriate design limits of metal containment system components, this guide (e.g., stress or strain limits) for the most adverse delineates acceptable design limits for appropriate combination of loadings to which these components combinations ofloadings. The intent of this guide is to might be subjected. For components constructed in address only the most adverse combinations ofloadings accordance with Subsection NE (Code Class MC) of resulting from those events or conditions identified Section til of the American Society of Mechanical herein (e.g., those combinat ons of loadings that result in Engineers (ASME) Boiler and Pressure Vessel Code, the limiting or controlling design condition). Rese provision of a design specification is required which loadings are associated either with conditions for which stipulates the design requirements for the components the containment function is required in combination I ( e.g., the mechanical and operational loadings). with specified seismic events (i.e., one-half the SSE and i However, neither Section 111 nor any other published SSE) or with other conditions (appropriately combmed code or national standard provides adequate guidance with specified seismic events) producing possible for selecting combinations ofloadings for design or for mechanisms for failure that could affect the function identifying Seismic Category I components (i.e., and/or structural integrity of structures, systems, and components that should be designed to remain components important to safety. Included in the latter usAEC REGut.ATORY GUIDES C .e g. n., ny r .,
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" ',, L".*.*'r.'0".,*:::'L%"; .. :.'" '71.:". '" LA'*"C","MA.':::TJC;;.'*"'"""" *'"'*" "
- "TI., " ::*#' ,,'.?.r!" '.".,"," *. ,7" ' ,, .f,'::',';:'.:";*.~": n.. - n.. ,,
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- 2. R re w.sT tRe r.
a
- 7. Tr.,uport.reen arW ,et n en no .,uf e.e.n i G.sur.s I
Y
as delineated in regulatory position C.I.c., the design 2. Bellows-Type Expansion Joints that Are Parts or
,m limits of NE 3131(cXI) or (2)are specified.NE 3131(c) Appurtenances of ASME Code Class MC Vessels of the Code distinguishes in the application of design a. For the tests stipulated by NE 6000 of the limits between areas of the containment structure that Code as delineated in regulatory position C.2.a., the are integral and continuous and those that are not (e.g., applicable design limits of NE 6000, as supplemented by bolted flanges and mechanical joints). For the' integral NE 3810(b) of the Code, are specified to provide and continuous regions of the containment, an increase assurance of both pressure retaining integrity and in allowable stress intensity is permitted by functional performance. For tests in addition to the ten NE 3131(cX2) of the Code to accommodate the effects tests permitted by NE 6000 of the Code, the design of the SSE. However, NE 3131(cXI) of the Code limits for Testing Conditions are specified. Note 9 to the permits no increase in allowable stress intensity for regulatory position also applies since Testing Conditions noncontinuous and nonintegral areas of the containment should be evaluated in accordance with the cyclic design under earthquake loadings. requirements of NE 3810 of the Code.
- d. Jet impingement and associated reactions may b. The applicable design limits of NE 3810 of the occur on the containment structure as a result of the Code are specified for each of the following loading occurrence of postulated piping ruptures within the combinations as delineated in regulatory position C.2.b:
reactor coolant pressure boundary. When the impact (1) design loadings combined with loadings associated forces fromjet impingement and associated reactions are with the vibratory motion of 50 percent of the SSE,(2) considered in combination with design loadings and concurrent loadings associated with flooding the loadings associated with the vibratory motion of the containment for accident recovery and the vibratory SSE, as delir eated in regulatory position C.I.d., the motion of 50 percent of the SSE, or(3) design loadings allowable stress intensities local to the jet and reaction combined with loadings resulting from the occurrence of forces are limited to the values specified in either an SSE and impact forces resulting from jet NE.3131.2(a) or NE 3131.2(b) of the Code. These impingenet and associated reactions. Loadings design limits are applied to accommodate the extreme associated witn the vibratory motion of 50 percent of loadings local to the jet impingement or associated the SSE should be evaluated in accordance with the reactions without loss of pressure. retaining integrity. cyclic design requirements of NE 3810 of the Code as NE 3131.2(a) of the Code restricts the allowable stress stated in note 10 to the regulatory position. Note 11 to intensities to the values of NE 313I(cX2)in regions of the regulatory position provides consistency between the the containment structure that are not integral and design limits inherent in using the procedures of continuous and in regions where partial penetration NE 3810(e) I or 2 and NE.3810(e) 3 of the Code. In c welds form part of the containment system boundary in addition, for the reasons given in note 12 to the the immediate areas of penetrations and access openings. regulatory position, the requirements of NE-3810(c) of NE 3131.2(b) of the Code permits the use of 85 percent the Code should be met by testing the major structural of the stress intensity values of Appendix F of Section assemblies in which bellows. type expansion joints are lit for areas local to jet impingement and reaction installed. loadings not excluded by NE 3131.2(a),
- e. The loading combination delineated in regulatory position C.I.2. encompasses those loadings C, REGULATORY POSITION that produce the greatest potential for shell instability (buckling) of containment pressure retaining ASME Code' Class MC components of primary components. The design limits of NE 3131,1 of the metal containment sys te ms' that are completely Code are specified for this loading combination; enclosed wittun Seismic Category I structures should be however, if a detailed analysis is performed, note 7 to designed to withstand the following loading the regulatory position set forth in this guide applies. combinations within the design limits specitied.
The factor of 2 between the critical buckling stress and the apphed suess a4 specified in note 7 is based on I. ASME Code Class MC vessels, electrical and generally applied margins used where shell buckling is a mechanical penetration assemblies, and other design consideration. Design loadings (as combined with penetration assemblies (excluding bellows. type loadings associated with the vibratory motion of the expansion joints) that are parts or appurtenances
- of the SSE) include design external pressure,if applicable,and vessel:
all other concurrent loadings that induce compressive 2. The design limits specified in either NE 6222 or stresses as outlined in note 8 to the regulatory position. NE 6322 of the Code, as applicable, should not be In reference to design external pressure, the condition of exceeded when the component is subjected to a concern is the maximum net differential external hydrostatie test, a pneumatic test, or a leak test,and the pressure that occurs across the containment vessel. This design limits of NE 3226(a), (b), and (c) plus loading should be evaluated for all containment designs, NE-3131(d) of the Code should not be exceeded when but may be significant only for cases in which a the component is subjected to a hydrostatic test, a limited-leakage concrete slueld building with annular pneumatic test, or a leak test in addition to the ten such space surrounds the steel containment vessel. tests permitted by NE 6222 and NE 6322 of the Code. 1.57 3
0
, DEFINITIONS O
ASME Code Class MC Components. Metal containment connections to pass through the containment vessel shell
% vessels including parts and appurtenances thereof that or head and maintain leaktight integrity while are constructed in accordance with the rules of compensating for such things as temperature and Subsection NE of Section 111 of the ASME Boiler and pressure fluctuations and carthquake movements.
Pressure Vessel Code. Parts or appurtenances of the containment vessel that perform a containment pressure Primary Metal Containment System. Includes the boundary function may include mechanical penetration following components: assemblies (including personnel or equipment hatches), I. The containment vessel or vessels; electrical penetration asemblies, piping penetration 2. All penetration assemblies or appurtenances not a assemblies, and bellows. type expansion joints. part of the vessel;
- 3. All piping systems attached to containment vessel Design Leadiny. Includes all static and dynamic nozzles or to penetration assemblies out to and including loadings used to design the containment vessel such as all pumps, instrumentation connections, and the valves design loadings associated with specified seismic events required to isolate the containment system and provide a (e.g., 1/2 SSE and SSE), design loadings that are pressure boundary for the containment function.
superimposed from other systems or components, and design pressure and temperature loadings (excluding, for Safe Shutdown Earthquake (SSE) That earthquake the purposes of this guide, jet impingement and which produces the vibratory ground motion for which associated reactions) from loss.of. coolant accidents due structures, systems, and components important to safety to the occurrence of postulated piping ruptures within are designed to remain functional. the reactor coolant pressure boundary. Seismic Category I. Those structures, systems, and Penetration Assemblies. Parts or appurtenances required components that are designed to remain functionalif the to permit piping, mechanical devices, and electrical SSE occurs.
-~,
9* 1.57 5 _. - _ . ~ .
QM . E iSk W_4 NUREG-75/087 NN Yna afog h o e
, ~' g U.S. NUCLEAR REGULATORY COMMISSION f -
i 4 STANDARD REVIEW PLAN OFFICE OF NUCLEAR REACTOR REGULATION SECTION 3.7.2 SEISMIC SYSTEM ANALYSIS REVIEW RE5p0NSIBILITIES Primary - Structural Engineering Branch (5ES) Secondary - None I. AREAS OF REV!EW The following areas related to the seismic system analysis described in the applicant's safety analysis report (SAR) are reviewed.
- 1. Seitmic Analysis Wethods_
Fo* all Category I structures, systems, and components, the applicable seismic analysis methods (response spectra, time history, equivalent static load) are reviewed. The manner in which the dynamic system analysis method is perfonned, including the modeling of foundation torsion, rocking and translation, is reviewed. The method chosen for selection of significant modes and an adequate number of masses or degrees of freedom is reviewed. The manner in which consideration is given in the seismic dynamic analysis to maximum relative displacements between supports is reviewed. In addition, other significant effects that are accounted for in the dynamic seismic analysis such as hydrodynamic effects and nonlinear response are reviewed. If tests or empirical methods are used in lieu of analysis for any Category I structure, the testing procedure. load levels, and acceptance basis are also reviewed.
- 2. Natural Frequencies and Response Leads For the operating license review, significant natural frequencies and response loads -
for major Category I structures are reviewed. In addition the response spectra at critical major Category I equiprient elevations and points of support are reviewed.
- 3. Procedures Used for Analytical Modeling Tht criteria and procedures used in moceling for the se'*.mic system analyses are reviewed. The criteria and bases for determining whether a component or structure is analyzed as part of a system analysis or independently as a subsystem are also reviewed.
USNRC STANDARO REVIEW PLAN
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- 8. Interaction of Non-Category I Structures with Category I Structures
~*
The design criteria to account for the seismic motion of non-Category I structures or portions thereof in the seismic design of Category I structures or portiens thereof am reviewed. The procedures that are used to protect Category I structures from the structural failure of non-Category I structures, due to seismic effects, are reviewed.
- 9. Effects of Parameter Variations on Floor Responses The procedures that are used to consider the effects of the expected variations of structural properties, dampings, soil properties, and soil-structure interaction on the floor response spectra and time histories are reviewed.
- 10. Use of Constant vertical Static Factors Where applicable, justification for the use of constant static factors as vertical response loads for designing Category I structures, systems, and components in lieu of the use of a vertical seismic system dynamic analysis is reviewed.
- 11. Methods used to Account for Torsional Effects ,
The method employed to consider torsional effects in the seismic analysis of Category I structures is reviewed. The review includes the evaluation of the conservatism of any approximate methods to account for torsional accelerations in the seismic design of Category I structures. ! 12. Comparison of Responses For the operating license review, where applicable, the comparison of seismic responses
]
for major Category I structures using modal response spectrum and time history approaches is evaluated. 11
- thnrtt for Seismic Analysis of Category I Dams The analytical methods and procedures that will be used for seismic analysis of Category
. I dams are reviewed. The assumptions made, the boundary conditions used, and the procedums by which strain-dependent soil properties are incorporated in the analysis are reviewed. 14 Detemination of Category I Structure Overturning Moments The description of the dynamic methods and procedures used to detemine design over-turning moments for Category I structures are reviewed.
, 15. Analysis Procedure for Camping The analysis procedure to account for the damping in different elements of the model of a coupled system is reviewed. . .
!!. ACCEPTANCE CRITERIA The acceptance criteria for the areas of review described in Section I of this plan are given below. Other approaches which can be justified to be equivalent to or more conserv-ative than the stated acceptance criteria may be used.
..)
3.7.2-3 11/24/75
~ T (3) To obtain an equivalent static load of a structure, equipment, or component " which can be represented by a simple model, a facter of 1.5 is applied to the peak acceleration of the applicable floor response spectrum. A factor of less than 1.5 riy be used if adequate justification is provided.
In addition, for equipment which can be modeled adequately as a one-degree-of-freedom system, the use of a static load equivalent to the peak of the floor response spectra is acceptable. For piping supported at only two points, the use
~
of a static load equivalent to the peak of the floor response spectra is also acceptable.
- 2. Natural Frequencies and Response loads To be acceptable for the operating license review, the following information should be provided.
- a. A sumary of natural frequencies, response loads, mode shades, and modal responses for a representative number of major Category I structures, including the contain-ment building.
- b. A time history of acceleration (or equivalent parameters) or response spectrum at the major plant equipment elevations and points of support.
- 3. Procedures Used for Analytical Modelino
~
A nuclear power plant facility consists of very complex structural systems. To be acceptable, the stiffness, mass, and damping characteristics of the structural systems should be adequately incorporated into the analytical models. Specifically, the following items should be considered in analytical modeling:
.. hssi :6tien of Systems Versus Subsystems Major Category I structures that are considered in conjunction with foundation media in forming a soil-structure interaction model are defined as " seismic systems." Other Category I stnactures, systems, and components that are not designated as " seismic systems" should be considered as " seismic subsystems."
- b. Decouplino Criteria for Subsystems It can be shown, in general, that the absolute frequencies of systems and suo-systems have negligible effect on the error due to decoupling. It can be shown that the mass ratio, R,, and the frequency ratio, Rf, govern the results where R, and Rg are defined as:
g , Total mass of the ' supported subsystem - m Mass tnat supports tne suosystem Fundamental frecuency of the supported subsystem R f = Frequency of the dominant support motion The following criteria are acceptable: 3.7.2-5 11/24n5
Table 3.7.2-1 Acceptable Methods for Soil-Structure Interaction Analysis Soil Foundation ** Shallewly Emoecded Case Method of Soil-Structure Deeply Deep Soil Deep Soil Shallow Interaction Rocktt Embedded Found, w/Unifom Found, w/ Layered Soil Analysis Foundation Caset Properties Properties Founcation Single Lumped Mass-Spring Approach x x Multiple Lumped Mass-Spring Approach
- x x x x Finite Element Approach
- x x x x x
*0r equivalent.
+0eep embethent: actual embedded depth >l5% of the least base width or other appropriate value to be justified, ttA medium for which the soil-structure interaction effect is negligible or alternatively, a medium with a shear wave velocity greater than or equal to 3500 fps.
** Soil foundation means the depth of soil between the bottom of the foundation slab and the base rock.
The lumped mass-spring method or " compliance function method" may be used for cases where the depth of embe<hent is shallow and the soil foundation is relatively unifom and of sufficient depth that it can be considered as a half-space. The acceptability of the procedures used to account for the effects of adjacent . structurts on structural responses in soil-structure interaction analyses will be reviewed on a case-by-case basis. Other techniques which give an equivalent degree of conservatism as the appropriate acceptable technique and which are justified are also acccotaole. Since the finite element and the lumped mass spring approaches are the most cormnonly used in current - practice, these two approaches are discussed below. Finite Element Approach The finite element approach may be used for all cases where soil-structure interaction is involved. The acceptance criteria for different aspects of the finite element technique are as follows: . . .
- a. Boundary Conditions (1) Bottom Boundary Wherever possible the base of the model is placed at the rock level. However, if the bedrock is too deep, the bottom boundary can also be placed at a 3.7.2-7 11/24/75
variation in the earthquake motion itself and also the inertia effort of the soil cannot be taken care of adequately by the lumped parameter method. (3) In the lumped parameter technique the entire soil medium is assumed 'o be homogeneous and elastic. However, in any stratum of soil deposits there are generally many different types of soils present. The properties also are strain-dependent. These factors could have a significant effect on the structural response, which cannot be accounted for by a single set of stiff-ness and damping values. (4) As stated earlier, if the site situation can be approximated as an elastic half-space, the design earthquake can be directly input in this approach. This is a relative advantage of the method. However, as Whitman (Ref. 8) points out, for a structure sitting on a stratum whose thickness is less than twice the width of the foundation, the effects of seil amplification and soil-structure interaction cannot be separated. So, he notes, the input motion at the spring support in such cases should be the rock motion, nn the design motion specified at the foundation level. It is , therefore, obvious that in cases of shallow overburden, an uncertainty about the input motien exists. Thus, the actual site conditions for a particular plant should be carefully reviewed before accepting the lumped spring approach.
- 5. Development of Floor Response Spectra To be acceptable, the floor response spectra should be developed taking into consideration the three components of the earthquake motion. The individual floor response spectral values for each frequency are obtained for one vertical and two mutually perpendicular horizer.tal earthquake motions and are combined according to the " square root of the sum of the squares" method to predict the total floor response spectrum for that particular frequency.
In general, development of the floor response spectra is acceptable if a time history approach is used. If a modal response spectra method of analysis is used to cevelop the floor response spectra, the justification for its conservatism and equivalency to that of a time history method must be demonstrated by representative examples. Depending upon what basic methods are used in the seismic analysis, i.e., response spectra or time history method, the following two approaches are considered accepuble for the combination of three-dimensional earthquake effects. (Ref. 3, 4, and 5.) i N 5' Yed If8[9 N 8I8Jt N
--- 3.7.2-9 11/24/75
The most probable system response, R, is given by N R = (I K=1 R2+2I lR R,[ g (2) where the second suonation is to be done on all t and a modes whose frequencies are closely spaced to each other. Other approaches which give an equivalent degree of conserystism,to the above methods, and which are adequately justified are also acceptable.
- 8. Interaction of Non-Category I Structures with Cattoory I Structures To be acceptable, the interfaces between Category I and non-Category I structures and plant equipment must be designed for the dynamic loads and displacements produced by both the Category I and non-Category I structures and plant equipment. In addition, a statement indicating the fact that all non-Category I structures meet any one of the following requirements should be provided. *** ,
- 4. The collapse of any nea-Category I structure will not cause the non-Category I l
structure to strike a seismic Category I structum or component. O
- b. The collapse of any non-Category I structure will not impair the integrity of seismic Category I structures or components. .
- c. The non-Category I structums will be analyzed and designed to prevent their failure under $$E coneitions in a menner such that the margin of safety of these structures is equivalent to that of Category I structures.
- 9. E**ec*? Of Daremeter Variations on Floor Response Spectra To be acceptable, consideration should be given in the analyses to the effects on floor response spectra (e.g., peak width and period coordinates) of erpected variations of structural properties, despings, soil properties, and soil-structure interactions.
An acceptable method for determining the amount of peak widening associated with the structural fmquency is described below. t Let f be the j th mode structural frequency which is determined from the structure y model. The variation in the structural frequency is determined by evaluating the individual frequency variation due to the variation in each parameter that has significant effect, such as the soil shear modulus, dessing, and meterial density.
.The total frequency variation, af j, is then determined by taking the squart root .
of the sum of squares of a minisman variation of 0.05f) and the individual frequency [ ! variation (afj )n, that is, ; i af)==#(0.05f)2+:(af)f j y A value of 0.10 f) is used if the actual computed value of af) is less than 0.10f,. [
-- 3.7.2 11 i
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- 14. Determination of Categorv i Structure overturnina Moments
% To be acceptable, the determination of the design moment for overturning should incorporate the following items:
- a. Three cogonents of input motion,
- b. Conservative consideration of vertical and lateral seismic forces.
- 15. Analysis Procedure for Dancina Either the composite modal damping approach or the modal synthesis tachnique can be used to account for element-associated damping.
For the composite modal damping approach, two tecnniques of determining an equiva-lent modal damping matrix or composite damping matrix are cosmonly used. They are based on the use of the mass or stiffness as a weighting function in generating the composite modal damping. The formulations lead to: I j= (el [k](*) (3) (e)T ({j g,) s) = (e)T[g) g,3 (4) where (K) e assembled stiffness matrix, th g, I) = equivalent modal daging ratio of the j [k). (k] = the modified stiffness or mass matrix constructed from element matrices formed by the product of the damping ratio for the element and its stiffness or mass attrix and, Ih normalized modal vector. (e) = J For models that take the soil-structure interaction into account by the Imped soil spring approach, the method defined by equation (4) is acceptable. For fixed base models, either equation (3) or (4) may be used. Other techniques based on model synthesis (Ref. 9) have been developed and are particularly useful when more, de-tailed data on the damping characteristics of structural subsystems are available. Themodelsynthesisanalysisprocedureconsistsof(1)extractionofsufficient modes from the structure model. (2) extraction of sufficient modes from the finite l element soll model, and (3) performance of a coupled analysis using the modal synthesis ! technique, which uses the data obtained in steps (1) and (2) with appropriate damping ratios for structure and soil subsystems. This method is based upon satisfaction of (
, 3.7.2-13 1
11/2a/75
e response spectrum method of analysis is used to develop the floor response spectra, its N conservatism compared to that of a time history approach is reviewed. The applicant is requested to provide additional technical justification for any procedure considered not adequately justified.
- 6. Three Components of Earthquake Motion
- The procedures by which the three components of earthquake motion are considered in determining the seismic response of structures, systems, and components are reviewed to detemine compilance with the acceptance criteria of Section !!.6. Any other pro-cedures that are considered not adequately justified are so identified, and the applicant is asked to provide additional justification.
- 7. Combination of Modal Responses The procedures for combining modal responses (shears moments. Stresses, deflections, and accelerations) for closely spaced modes are reviewed to detemine compliance with the acceptance critierta of Section II.7 when a response spectrum modal analysis method is used.
- 8. Interaction of Non-Category I Structures with Category I Structures The design and analysis criteria for interaction of non-Category I structures with Category I structures are reviewed to ensure compliance with the acceptance criteria of Section !!.8.
- 9. Effects of parameter Variations on Floor Response Spectra The seismic system analysis is reviewed to determine whether the analysis considered the effects of espected variations of structural properties, dampings, soil properties, and soil-structure interaction on floor responses spectra (e.g., peak width and period coordinates) and to detemine compliance with the acceptance criteria of Section !!.9.
- 10. Use of Constant Vertical Static Factors Use of constant static factors as response loads in the vertical direction for the seismic design of any Category I structure, system or component in lieu of a detailed dynamic mothed is reviewed to detemine that constant vertical static factors are used only if the structure is rigid in the vertical direction.
- 11. Methods Used to Account for Torsional Effects The methods of seismic analysis are reviewed to detemine that the torsional effects of vibration are incorporated by including the torsional degrees of freedom in the dynamic model. Justification provided by the applicant for the use of any approximate method to account for torsional effects is judged to assure that it results in a conservative design. If such justification is deemed inadequate it is identified and tne applicant requested to provide additional justification.
- 12. Comparison of pesconses Where applicable the responses obtained from both response spectrum and time history methods at selected points in major Category I structures are compared to judge the 3.7.2 15 11/24/75
for the analyses of all major Category I structures.. systems, and components. When the modal response spectrum method is used. governing response parameters are comoined f-- , by the squart root of the sum of the squares rule. However, the absolute sum of the modal responses are used for modes with closely spaced frequencies. The square root of the sum of the squares of the maximum codirectional responses is used in accounting for three components of the earthquake motion for both the time history and response spectrum methods. Floor spectra inputs to be used for design and tes'. verifications of structures, systems, and components are generated from the time history method. taking into account variation of parameters by peak widening. A vertical seismic system dynamic analysis will be employed for all structures, systems, and components where analyses show significant structural ampitffcation in the vertical direction. Torsional effects and stability against overturning are considered.
"The finite element (or the lumped soil spring) approach is used to evaluate soil.
structure interaction and structure-to-structure interaction effects upon seismic responses. For the finite element analysis, appropriate nonlinear stress-strain and damp'ngi relationships for the soll a-e considered in the analysis.
"For the analysts of Category I dams, a finite element approach which takes into consideration the time history of forces, the behavior and deformation of the dam due to the earthquake and applicable stress-strain relations is used.
"We conclude that the seismic system and subsystem analysts procedures and criteria proposed by the applicant provide an acceptable basis for the seismic design."
V. REFERENCE $
~
- 1. P. K. Agrawal. " Comparative Study for Soll-Structure Interaction Effect by the Soll Spring and Finite Element Model." Report No. SAO-082. Sargent & Lundy Engineers (1973).
- 2. P. Schnabel. H. 8. Seed, and J. Lysmer. " Modification of Seismograph Records for Effects of Local Soil Conditions." Bulletin of the Setsmological Society of America. Vol. 62 No. 6. pp. 1649-1654 (1972).
3, 1. M. Ne w ark. J. A. 81ume, and K. K. Kapur " Design Response Spectra for Nuclear Power Plants.* Journal of the Power Division. American Society of Civil Engineers, pp. 287 303 Novereer 1973. 4 S. L. Chu. M. Amin, and S. Singh. " Spectral Treatment of Actions of Three Earthquake Components on Structures." Nuclear Engineering and Design. Vol. 21, pp.126136 (1972).
- 5. N. M. Newark and E. Rosenblueth. " Fundamentals of Earthquake Engineering " Prentice Hall,op. 236-237 (1971).
- 6. H. 8. Seed. K. L. Lee. I. M. Idriss, and F. Handist. " Analysis of the $lides in the San Fernando Dams During the Earthouake of Feb. 9,1971." Report. Earthquake Engineering Research Center. Univ, of Calif., Berkeley, March 1973.
3.7.2-17 11/24/75
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c '# " % UNITED STATES 7.
" i1 NUCLEAR REGULATORY COMMISSION 1I g WASHINGTON. D. C. 20656 y
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MEMORANDdM FOR: Chaiman Hendrie Commissioner Gilinsky Commissioner Kennedy Comissioner Bradford
- h j 'Comisstoner Ahearne FROM: Harold R. Denton, Director j Office of Nuclear Reactor R ulation
.C rd.
THRU: , V Go k, Executive o
SUBJECT:
STAFF REQUIREMENTS FROM THE MAY 3, 1979 COPei!SSION BRIEFING ON THE STATUS OF MAINE YANKEE , l 1. The staff completed its review of the final set of PSTRESS/ SHOCK 1 i to NUPIPE-SW code comparisons of May 3, 1979 and concluded that j all requirements of the show cause order of March 13, 1979, has been met with respect to the Maine Yankee facility. l The staff has also reviewed the licensee's response to IE j Bulletin No. 79-068 (Review of Operational Errors and System ' j Misalignments Identified During the Three Mile Island Incident) , and finds this response satisfactory.
- 2. Secretary Chilk's memorandum to Lee V. Gossick dated May 4,1979,
; requested that staff provide responses to two questions raised during the May 3,1979 briefing. The Conunission requested:
1 1
- a. A list of the seismic design margir.s built into the Maine I
Ya ' :e facility that~ leads the staff to conclude that the - Ma.ne Yankee facility structures could absorb a .2g level - of ground motion.
- b. An estimate of the time and effort that would be required j
to complete an evaluation of the Maine Yankee seismic design under the NRC's current seismic design criteria set forth in Regulatory Guide 1.60. ; Enclosure (1), Discussion of Conservatisms in Maine Yankee Seismic Design, and Enclosure (2), Estimate of Manpower and Cost for
- . Seimic Reanalysis, are the staff's responys to these requests.
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s- I The Commission . 3. On the basis of the review discussed in paragraph 1 above, the staff concluded that the requirements of the show cause order had been met and that continued shutdown of the facility was not required. On May 9,1979, however, the licensee reported the earlier versions of SHOCK 1 known as SHOCK 0 may have computed piping natural frequencies incorrectly. The lic~ensee is reviewing whether or not this latest information is significant and will report the results of its review. The staff will review the information submitted and inform the Commission if this
'information significantly effects our evaluation and our recommendation regarding restart of the facility.
- 4. Also enclosed is a draft of the Order which I propose to issue, upon satisfactory resolution of the SH0CK 0 piping natural frequencies computation matter. The order, which will terminate the show cause order of March 13, 1979, which led to the current suspension of operation of the Maine Yankee facility, together with the NRR Safety Evaluation of the actions taken by the licensee in response to the show.cause order, are forwarded at this time for Comission review and connent. I will inform the Connission when the additional review of the SHOCK 0 piping natural frequency matter is complete and I am
. prepared to issue the order.
The order contains the staff's conclusion that the facility could withstand seismic events in excess of the current design basis seismic event, even though it is beyond the scope of matters addressed in the original show cause order. While the Office of the Executive Legal Director has no legal objection to the inclusion of this finding in the order, it points out that the discussion of seismic conservatism is legally separate from the issues addressed in the show cause order, the responses to that order, and the proposed order permitting resum t, ion of opera on of Maine Yankee.
/ /
7 j/ arold H
' p.S_,
R. Denton, Director hv Office of Nuclear Reactor Regulation
- 1. Discussion of Conservatisms in Maine Yankee Seismic Design
- 2. Estimate of Manpower and Cost for Seismic Reanalysis
- 3. Proposed Order Terminating Suspension of Operation of Maine Yankee i.
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The Comission DISTRIBUTION Docket (50-309) NRC PDR SECY GC Local PDR ORB #4 Rdg NRR Rdg EDO Rdg - VStello HRDenton EGCase DGEisenhut - CNelson RReid BKGrimes VNoonan JKnight RDenise LVGossick TJCarter WTRussell RIngram RVoll GCunni ham l l
DISCUSSION OF CONSERVATISMS IN MAINE YANKEE'S SEISMIC DESIGN While increasing the SSE seismic input from 0.lg with a Housner spectrum to
~
between 0.13 to 0.29 with a Regulatory Guide 1.60 spectrum may seem to be a large percentage increase in seismic input, the inherent resistance of a facility properly designed to 0.1g should, in general, provide adequate resistance to For example, the relatively low seismic input of between 0.13 to 0.29 th: impact of increasing from 0.19 to 0.29 is much less severe than going from 0.25g to 0.59 This is because nuclear plant designs are based on various combinations, of loads with seismic loads as only one part. As an example, of the 85 piping runs analyzed at Maine Yankee, all of the peak stress points would be less than 50% of ultimate strength even if the seismic . stresses are doubled from the 0.1g level. Only six of the runs sould have peak stresses greater than current allowable stress limits. even though eleven runs would have peak stresses exceeding the more conservative criteria in the FSAR. Of the six runs with peak stresses over current allowable stress limits, it is likely that these stresses would be less than the actual material yield stress. Seismic design of nuclear power requires interaction between these principal
~
endeavors: (1) definition of the seismic hazard, in terms of intensity and ( characteristics of shaking, and (2) design of structures, systems and components to resist the defined seismic shaking. The definition of seismic hazard invloves consideration of the geology and sefs-mology of the region, observed ground motion, and observed effects of earthquakes. The information available for historic records, measurements recorded in more recent years, and insights that can be gained from analyses and damage assessment
~
_ 2_
' 11owing earthquakes have been synthesized to arrive at the engineering methods we use to define the seismic hazards for nuclear power plants, dams and other public structures. __
The seismic input, once defined, is used in a mathematical process to determine how the structure would vibrate in response to the seismic shaking. Throughout this process very couplex natural phenomena and the response of complex structures and equipment are idealized so that the principles of applied mechanics and mathematics can be employed to determine,the response of each of the major portions of the structures and equipment. To compensate for these idealizations the engineering practices involved in the seismic design for nuclear power plants establish a conservative design quantity at each stage in the analytical process (see the attached list of conservatisms). The final design, resulting from
. ,3mpounding of the conservatisms in each step, is therefore also conservative.
For plants of the Maine Yankee vintage, conservatisms in the seismic analysis and design for strpetures, systems and components are generally found in the following areas: (1) Elastic dynamic analyses are performed using conservatively low damping values. (2) Multiple-directional seismic input, with each horizontal component having equal intensity, is considered in design of plants. Actual earthquakes are typically stronger in one direction. l
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) The OBE is selected at one-half of the SSE and controls the design in many cases, rather than the SSE, due to the substantially lower allowable stresses for the OBE. __
(4) Loading combinations consider other loadings (dead weight, live loads, pressure loads, etc.) in addition to the seismic loadings. Seismic loading is therefore only a part of the total loading and in fact loadings other than seismic may govern designs. A sizable increase in seismic stresses may be only a small addition to the. total stresses. K (5) In' the design of structures and equipme'nt, it is convenient to assure that all elements of the structure or equipment are designed to stress levels well below the actual strength of the materia's so that any permanent deformation is very small. This approach obviates the need for complex and costly inelastic analyses. Inelastic behavior would significantly reduce structural response prior to failure. (6) Stress limits, whether elastic or inelastic, are based upon material behavior under static loading conditions. Since dynamic loads contain a limited amount of energy, the margin (between the stress limits and failure) under dynamic loads is grehter than under static loads if elastica 11y calculated peak response is compared to the stress limits with strain rate effects neglected.
(7) The design of the structural elements is such that their capacity usually exceeds the seismic requirements called for by the analyses. In Maine . Yankee, orthogonally spaced reinforcing steel was used in the containment wall with additional diagonal reinforcement at large penetrations. Much of the actual structural design is controlled by the availability of standard structural members, such as beams and piping sections, so that larger, sizes than are needed are often used. In-situ (8) Engineering codes specify " code minimum strength" for materials, strengths are usually higher. Additional conservatisms for major mechanical components and piping can be found in: 1) When the floor response spectra are developed for the design of components located at different locations in the structure, the peaks in the individual flour a esi,cesa spectra are broadened in order to predict conservative responses. (2) Where the system has multiple supports, maxiritum response spectra is usually applied to all support points. When calculating the seismic loads for components, conservatively established , (3) values are applied several times (first, to major structures, then to the intermediate structures and finally the equipment themselves). (4) Even identically designed redundant systems may not always experience similar seismic excitation due to different mounting locations, with different structural filtering effects. Thus, a loss of a redundant component may not mean a loss of function for the system. E
- 5-h The end result of the conservatisms employed in the analyses, followed by t e conservatisms resulting from standard design practices, is structures and c:mponents with seismic capability well in excess of the established design goal. ,
This is the reason that the record is replete with cases where well-engineered structures, even those for which no specific seismic design standard was invoked, A number of have withstood major earthquakes while remaining fully functional. The Esso re-plants of various kinds have been subjected to large earthquakes. Another example is the pump finery in Managua, Nicaragua is a good example. ' stations in the Exxon pipeline in Italy.. subjected to the Friuli earthquakes. These are structures.that were designed by ordinary codes, with perhaps the The earthquakes that seismic design coefficient of the order of .05 to .089 occurred had accelerations that were measured of the order of.35g in Managua and perhaps more than that in Friuli. The Esso refinery was able to continue
,perating with no damage to any of the equipment while the pump stations on the Exxon pipeline were able to continue operating without damage to the equipment.
For these reasons and taking into account the use of orthogonally spaced rainforcing steel in the containment wall, the staff judgment is that the major structural components of the Maine Yankee facility will likely remain functional even for an increased range of seismic input of from 0.13 to 0.2g. - Even at the 0.29 level, it is unlikely that the seismic event would initiate
,a serious accident. For minor mechanical and electrical equipment, where the fragility.is likely lower, loss of function is not expected to be sufficient to prevent plant shutdown when all plant systems and available corrective actions are considered.
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...o likelihood of the SSE is presently judged to be on the order of 10-3 10-4 per year for the 0.13 to 0.2g range, decreasing with the higher values.
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The confidence in the judgment that major structural components will likely .
~
remain functional increases at the lower SSE range. Th2 NRC will be further reviewing the subject of seismic design capability of all operating reactors within the next few months. That effort will assist the staff in determining whether and when additional seismic re-evaluation is needed at any operating facility, including Maine Yankee. a
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CONSERVATISMS IN SEISMIC DESIGN I. Seismic Design for Ground Motion Enveloping response spectra and time histories , Conservative OBE (usually controls design) II. Seismic Analysis and Design Method
- a. Structures, systems and components Elastic dynamic analysis (inelastic behavior can significantly reduce response spectra)
Damping valpes Multi-directional earthquakesLoading combinations (seismic o
- b. Additional conservatisms for piping and major components Peak widening of floor response spectra
. System Redundancy Generic Qualification for Many PlantsUse of maximum and supported systems Multiple applications of damping values -
III. Structural and mechanical resistance factors
. Allowable stress from Code Dynar.ic resistance of materials 28 day concrete strength Ductility to failure Minor attachments absorb energy Redundancy in structural elements Use of standard size pipe and equipment
. Quality Assurance .
IV. Seismic Experience to Date Inherent resistance shown for large industrial facilities Nuclear plant resistance shown in Japan Other loads (wind and pressure) influence design
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ESTIMATE OF MANPOWER AND COST FOR SEISMIC REANALYSIS
~
An estimate of the amount of resources needed to perform a seismic reanalysis for a nuclear power plant is directly dependent on the - extent of the reanalysis, i.e., the requirements for the reevaluation. A first approach would be to have the licensee reanalyze all safety-related features of the plant and have the staff review the reanalysis. The staff estimates that this would take up to 21/2 years or even longer. The range of time depends primarily on the actual scope of
= items to be analyzed, the number of modifications to be designed, efficiencies, and resource levels. For example, for Diablo Canyon, the entire plant was essentially analyzed in 21/2 years after the basic criteria was s'et, including about 10 manyears of staff review, under considerable cs' hedule pressure to get the plant started. Other examples of reanalyses, such as Humb~oldt Bay, Unit 3, and San Onofre, Unit 1, have taken much longer without the same schedule pressure.
A shorter time, perhaps one year, can be considered an optimistic schedule for a more abbreviated approach similar to that being done by the staff for certain of the SEP plants. It should be noted, however, that the result of the staff's seismic review of SEP plants may result in the requirement for a more detailed seismic reanalysis ~ i by the-licensee. Such an approach could be similar to the SEP review
< of Dresden 2. There the staff's consultants are reviewing the existing design over a period of several months and will determine what areas need further detailed reanalysis. The detailed analyses could then take approximately 1 year. For Dresden 2, where the previous design bases is not con:idcrably changed by presen't requirements, this approach will reduce the amount of detailed reanalysis. However, this may not follow for other plants, where the seismic input might be increased more significantly.
Based on these limited experiences, several additional months follow 4ng. reanalysis could be required to complete any needed physical modifications !- (i.e., if there are numerous modifications but installation begins j early in the process as the needs are identified.) The staff's cost estimates are approximate, but it appears that the ~ licensee's total analysis cost would be between $5 million and $15 i million depending on the approach used the staff's resources could vary from'a couple of manyears of effort to 10-15 manyears of effort. i
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UNITED STATES NUCLEAR REGULATORY COMMISSION f 'Q(') ,- j WASHINGTON, D. C. 20555
%...D.s SAFETY EVALUATION BY THE OFFICE OF NUCLEAR _ REACTOR REGULATION MAINE YANKEE AT0*11C POWER COMPANY MAINE YANKEE ATOMIC POWER STATION DOCKET NO. 50-309 Introduction On March 13, 1979 the Commission issued an Order to Show Cause to Maine Yankee Atomic Power Company (licensee) requiring that Maine Yankee (facility) be placed in cold shutdown and the licensee show cause:
'(1) Why the licensee should not reanalyze the facility piping systems for s'eismic loads on all potentT811y
.affected safety systems using an appropriate piping analysis computer code Which does not combine loads algebraically; (2) Why the licensee should not make any modifications to the facility piping systems indicated by such reanalysis to be necessary; and
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(3) Why facility operation should not be suspended pending such reanalysis and completion of any required modifications. The licensee's response to the Order, dated April 2, 1979, stated that a ll ofrected safety systems have been reanalyzed using an appropriate piping analysis method, and that no modifications are necessary as a result of these reanalyses. Therefore, the ifcensee requested that the Order be modified or rescinded such that the facility could be restarted. In support of this request the licensee provided information by letters dated April 2, 3, 12, 13, 19, 27 and May 2, 4 and 5, 1979. In the letter of April 13, the 1tcensee indicated that two piping res$raints needed to be modified as a result of the reanalyses to account for base plate flexibility. On April 19, the licensee reported that these modifications had been completed. Discussion The Stone and Webster (S&W) PSTRESS/SH0CK 2 computer code for pipe stress analyses sums earthquake loadings algebraically and is unacceptable for reasons set forth in the March 13, 1979 Order to Show Cause. This code was used in the seismic analyses of certain safety and nonsafety related systems at the facility. The licensee has identified the seismically analyzed (Seismic Category I) systems at the facility including those analyzed with SliOCK 2. It has also identified the other methods of seismic analysis used for other Seismic Category I systems. Furthermore, the licensee has sumarized the results of the reanalyses of SHOCK 2 safety systems and has provided support for the acceptability of the analysis methods used on the remaining Seismic Category I systems.
.P L -- ._. __
lle have evaluated the facility's safety related systems, the results of seismic reanalysis,and the methods of pipe stress analysis currently in effect for the facility. Evaluation
- 1. Systems .
The licensee has stated that the response to Question 1.3 of the Maine Yankee Final Safety Analysis Report (FSAR), submitted February 9,1971, is the complete list of structures, systems and components that were designed to the Seismic Category I requirements. Verification has also been provided by the licensee that the Seismic Category I piping systems identified in response to Question 1.3 of the Maine Yankee FSAR include all of the piping systems required to assure: (a) The integrity of the' reactor coolant pressure boundary; (b) The capability to shutdown the reactor and maintain it in a safe shutdown condition; and (c) The capability to prevent or mitigate the consequences of accidents.which could result in potential offsite exposures ' comparable to the guideline exposure of 10 CFR Part 100. Portions of the following systems were identified by the licensee as having been either analyzed with SH0CK 2 or analyzed by static seismic methods which were verified by SH0CK 2. H;W P essure Safety Injection Residual Heat Removal Containment Spray Low Pressure Safety Injection Primary Component Cooling Water Stear Generator Feedwater Chemical and Volume Control Primary Vents and Drains
- Waste Gas Disposal .
Baron Recovery Fuel Pool Cooling Fire Protection Auxiliary Steam Auxiliary Condensate Return High Pressure Drains (Secondary) Piping A total of 39 SHOCK 2 analyses (Computer runs) v.are performed. associated with these analyses and the methods cf rear *1ysis are identified in the Enclosure to this Safety Evaluation (SE). Y
* - - - . - - . - . ~ . . . . . .
I . i Nineteen of these 39 analyses have been identified by the licensee l l lle have reviewed the ! as pertaining to safety related piping, information submitted and agree with the licensee's identification ' of piping which is safety related. The ifcensee has completed the reanalysis of all 39 SHOCK 2 analyses. l 2. Verification of Analysis Methods We have reviewed the acceptability of the analytic methods which
- are currently a basis for the facility piping design. The licensee l
has identified the following computer codes / analysis r.ethods as applicable: l PSTRESS/ SHOCK 1 (4 Versions - Initial 3 Versions sometimes - referred to as SHOCK 0) STRUDL - SHAKE (Combustion Engineering) Static Analysis Methods PSTRESS/ SHOCK 3 NUPIPE - SW , l PSTRESS/ SHOCK 1 l The licensee has identified four (4) versions of the PSTRESS/SH')CK 1 l computer code. Documentation on only the last version of this code j was available for our review; however, earlier versions are expected i to use similar and definitely no more sophisticated methods of analyzing seismic loads. The licensee has stated that this version of SHOCK 1 ccmbines the intermodal responses by the so-called "flavy Method". This consists l l in taking the largest absolute modal response and adding the root- ! mean-square value of all other modal responses. Intramodal respenses due to multi-directional earthquake excitation were not calculated since the code only produced responses parallel to a given earthquake component excitation (i.e., the responses were considered uncoupled). A review of the code listing has confirmed these statements.
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Some safety systems of the facility were analyzed with each of the four versions of the SHOCK l Code. Because this computer code only - considers one direction earthquaEe excitation, it is not considered equivalent to current analysis techniques. A comparison of the results of each of the four (4) versions of PSTRESS/ SHOCK 1 and the NUPIPE Code was conducted by the licensee using " typical" piping problems. The problems consist of different size piping, elbows, tees and reducers. The licensee reported that the general stress distribution of both codes was similar and PSTRESS/ SHOCK 1 gave comparable results. The licensee concluded that although the PSTRESS/SH0CK 1 is not equivalent to current practice, it is suitably conservative to insure that the piping systems meet the allowable stress levels. - We have
- reviewed the piping configuration and results of the comparative analyses of NUPIPE and each version of the SHOCK 1 code.
We have determined that the problems analyzed produce , respresentative comparisons. We have also determined that although SH0CK 1 is not equivalent to current practice, the resulting stresses are at least consistent with the results as obtained from NUPIPE and in many cases are conservative . In addition the code comparison did not take credit for the alternative t' application of the " Robinson Fix" (i.e., adjusting the response spectra peak instead of increasing all analysis results) which would provide additional conservatism to the SH0CK 1 stresses in this ccm-parison. The " Robinson Fix" was described in Amendment 35 to the , Maine Yankee FSAR. Therefore, we conclude adequate assurance has been provided that systems analyzed with SHOCK 1 will withstand the design basis earthquake. Although this assurance regarding SH0CK 1 systems has been provided, the licensee has, by letter dated May 2, 1979, stated its intent 6 to reanalyze all SH0CK 1 systems using a method of analysis verified l against current criteria. By doing this,the licensee will not only l upgrade the facility's seismic design analysis but also greatly ' improve the records and accuracy of numerical results associated with the facility's seismic design analysis for possible future use. The licensee has also stated that following its review of l the details of this program and by June 1,1979 a schedule for l completion of this reanalysis will be proposed. We agree with the licensee that this reanalysis is appropriate. S
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e STRUDL - SHAKE - The licensee has provided the following description of the analysis technique used by Combustion EngTneering (STRUDL - SHAKE Code):
'The dynamic seismic analysis of the reactor coolt.nt system main loop and pressurizer surge line piping was performed utilizing 3 dimensional mathematical models subjected to unidirectional support motion response spectra. The six components of force or moment at a particular piping location were determined separately for each significant rode of response for a single direction of excitation. The separate modal responses for each component of force or moment were then combined on a root-sum-square basis to define the total force or moment response to a single djrection of excitation. The loads due to each horizontal earthquake were added, manually, to the loads due to the vertical earthquake by the absolute sum method. The larger of the two loads thus calculated was employed in the stress analysis of the piping system."
We have reviewed the analysis technique of Combustion Engineer-ing. The procedures are in compliance with the plant FSAR and i conservatively combine (absolute sum) both the spatial com- . ponents from each of two independent earthquake directions and the contribution of each mode (SRSS). We find this technique acceptable. Static Analysis Some of the safety related systems at the facility were The licensee submitted analyzedusing(staticanalysestechniques. documentation letter dated April 12,1979) detailing the basis ' for static analysis technique use in the design. Generally - piping 6 inches in diameter and smaller was designed using the static rothods unless the criteria.for support placceent could not be met, then a more rigorous dynamic analysis was performed. Some piping larger than 6 inches in diameter was analyzed using the static methods if the geometry and support configurations were sufficiently simple to make the static analysis methods practical. The major constraint on applying static methods to larger piping was one of
- econonics in that a dynamic analysis typically would result in fcwor restraints at a more optimum spacing and supports for larger piping were sufficiently more costly to warrant less conservative but more expensive analysis techniques.
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The analysis technique used at the facility is outlined in kendment No. 35 to the FSAR and the procedure-was submitted in detail in the report, "Ncn-dynamic Seismic Analysis of Piping 12, and1979. Supports by Stone & Webster at F4Tne Yankee" submitted April The procedure states that the piping frequencies will be designed to be a minimum of 1.5 times the peak resonant frequency of the cmplified response spectra by locating seismic supports at appropriate span lengths. Orthogonal responses will be decoupled by including supports at elbows, tees and concentrated masses. The piping systems were designed considering a horizontal static load of (1.3) X (22 X peak ground acceleration) acting concurrently with a vertical static load equal to two-thirds the horizontal val ue. The nathod of equivalent analysis outlined in this procedure has been reviewed against the flRC's Standard Review Plan 3.7.2 and is .cceptable. PSTRESS/SH0CK 3 The licensee has stated that in this code the intramodal responses are calculated by adding the absolute value of the responses due to the vertical earthquake component to the root-mean-square of the responses due to the two horizontal earthquake components. The intermodal com-ponents are calculated by the root-mean-square method. A review of the code listing has confimed these statements. A confirmatory analysis was performed by an NRC consultant, Brookhaven National Laboratory (BNL), of a typical piping design problem in the Maine Yankee plant. A problem (no. 803) has been submitted by S&W t0gether with the corresponding solution obtained by using PSTRESS/ SH0CK 3. This problem has been analyzed by BNL using a different code (EPIPE), and the results have been submitted to the NRC staff. A comparison of the solutions indicat'es that various quantities of r interest such as frequencies, displacements, forces, and strasses, appea, to differ by not more than 10%.which is within the accuracy of the analyses. In addition, hand calculations were perfomed with the PSTRESS/SH0CK 3 results as 4 check on the modal combination methods. We find that the S&W results have been adequately confirmed by BNL and are therefore acceptable. NUPIPE - SW_
.. The licensee has stated that this code calculates intramodal and intermodal responses according to the provisions in Regulatory Guide 1.92. A review of the code listing by the staff has confirmed this
. to be the case. Additional documentation has also been submitted I by the originators of this code (Nucicar Services Corporation) providing detailed information on the methods of modal combinations. I This information has been reviewed and also provides reasonabic A confirmatory confirmation of the statements made by the licensee. analysis has also been performed by our consultants on the piping A comparison of the solutions again indicates problem listed above.that the various quantities of interest listed above again dif Therefore, the use of this code is acceptable. not more than 10%.
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7-
- 3. Reanalysis l'ethods and Results The safety related piping systems at the l'aine Yankee nuclear plant have been reviewed to determine the method of an'alyses. flineteen (19) computer stress problems of safety rSlated piping have been identified where the analysis used an algebraic intramodal sunnation of responses to earthquake loadings. The problems where an algebraic intramodal response combination technique was used in the design have been reavaluated using the criteria in the FSAR. The reevaluation included a static analysis technique, and a dynamic ccmputer analysis using either the PSTRESS/SH0CK 3 or fiUPIPE programs.
A static analysis technique was employed for reanalysis of scme lines The static design prog 6 inches in a report in d bmeter titledand smaller.
" tion-dynamic Seismic Analysis of g and Supports Pipin,g, pur by Stone & oUebster at Maine Yankee." submitted April 12, 1979. The g
acceptability of this procedure has been discussed in Section 2 of this SE. The dynamic analysis technique incorporated a lumped mass response spectra The floor modal analysis using the PSTRESS/ SHOCK 3 or fiUPIPE programs. response data used in the reanalysis included the " Robinson Fix" criteria. The " Robinson Fix" criteria required the peak resonant frequency acceleration values to be a minimum of (22) x (peak ground acceleration) and the peaks tn be broadened by i 10% of resonant frequency. The piping systems were modeled as three dimensional lumped mass systems which included consid-erations of eccentric masses at The valves and appropriate ficxibility and dynamic analysis procedures meet stress intensificaticn factors. the criteria specified in the plant FSAR and are acceptable. The piping support designs for affected system piping were inspected by the licensee to v(rify the "as built" configuration. As noted in fiRC Inspection Report 79-05 issued. April 12, 1979, differences were fcund to exist between the "as built" c'onfiguration and the support drawings. The differences noted resulted from the use of drawings which had not baen Subsequently the licensee has updated verified that to include installation changes. updated drawings which do reflect the s were used in the support design calculations. The support designs were reevaluated in cases where the original support The support design loads were exceeded as a result of piping reanalysis. reevaluation ' included the consideration of local stresses at regions of discontinuity and base plate flexibility considerations. Modification
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of two supports was detemined to be necessary to account for base plate flexibility. These modifications consist of adding a stiffener to the base plate of each hanger and have been completed. Loads on attached equipment nozzels were also checked and verified to be either below the initial allowable _ values or verified by the equip-ment manufacturers to be acceptable. The design and analysis of the supports and attached equipment are in accordance with the criteria specified in the plant FSAR. The pipe break criteria for Maine Yankee vere reviewed and determined not to be altered by this reanalysis. 'of the containment structurePipe break considera required for High Energy Lines outside and break locations were detemined by inspection a'nd their proximity to safety related systems. The pipe break considerations are outlined in a report titled " Supplementary Report on Effects of a Postulated Break in a. High Energy P.iping System Outside the Containment" dated September 1973. ~ The piping systems and supports were designed to the allowable limits of ANSI B31.1 for the gross properties and to the limits of ANSI 831.7 Appendix F for local stress considerations per the FSAR criteria. The safety related piping systems, supports and attached equipment, I .. where the original analysis used an' algebraic.intramodal response summation technique, have been reanalyzed with acceptable methods which do not use an algebraic intramodal response technique. The pro-cedures used in the reanalyses and their results have been reviewed against the criteria in the plant FSAR and found acceptable. 1 Conclusion
~
The licensee has demonstrated that PSTRESS/SH0CK 2 is the only method of analysis used for the facility's safety related systems which combines seismic loads algebraically. Safety related systems analyzed with Shock 2 have been reanalyzed with an acceptable dynamic code or with static analysis techniques as permitted by the FSAR criteria. The results of those reanalyses have shovin that the subject syster.s will withstand the design basis earthquake. The reevaluation of supports performed by the licensee for the subject piping considered base plate flexibility. As a result stiffeners were added on two supports in the containment spray system. We reviewed the acceptability of the analysis techniques which are currently a basis for the facility's piping design. We have determined that the application of these techniques, at Maine Yankee, assures that { safety related systems can withstand the design basis carthquake and that there is reascnable assurance that the facility can operate without j endangering the health and safety of the public. l t I l i c
. g.
"ased on the ..bove,'te conclude that the requirements of the ,
Order have been mat for I:aine Yankee and therefore the Order c.nd its restriction on facility operation should be terminated. Dated: 4 6 0 e 4 o
, 4 O
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,l APR 0 41950 l MEMORANDUM FOR: V. Stello, Director Office of Inspection and Enforcement FROM: Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
PENNSYLVANIA PUBLIC UTILITY COMMISSION REQUEST FOR HRC ASSISTANCE IN PROCEEDINGS y The enclosed letter from the Pennsylvania Public Utility Comission (PUC) requests that niembers of the NRC staff appear before the PUC to give testimony on matters relating to the Beaver Valley Nuclear Generating Station, Unit 1. The PUC has indicated that last year's shutdown of Beaver Valley as the result of concerns regarding the seismic design of safety related piping systems is one area where NRC testimony will be requested. Bill Russell of my staff plans to provide testimony covering the period from March 1979 to present; however, the rules of evidence followed by the PUC will not permit his testimony regarding the events prior to March 1979. Accordingly, I request that you make available the Beaver Valley resident inspector to provide testimony on this issue for the period prior to March 1979. Bill Russell will be making the necessary arrangements with the PUC. If you need any further information, please call him at 492-8041. gg. . Harold R. Denton, Director Office of Nuclear Reactor Regulation
Attachment:
As stated
. cc: D. Wigginton W.J. Dircks g q p 805 .
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Narold Danten, Director office of Nuclear Reactor Regulation l'.5. Nuclear Regulatory Commission ; l Wcshinston, DC 20555 s Kat ?.equest for NRC assistance in proceedings of the Pennsylvania Public Utility Comission
Dear Mr. Denton:
The Fannsylvania Public Utility Co= mission has currently before it an investigation into the prolonged outages of the Beaver Valley Nuclear C:nerating Station, t* nit 1, at its Investigation Docket No. 1-79070314, et al. I have spoken at some length with Mr. William T. Russell on various aspects of Isaver Valley's operation and conclude that it would be most helpful
embers ef the NRC staf f would appear to give testimony before our Commission
'en technical =atters.
I therefore request your help in furnishing the assistance of members
*.s *..ve special knowledge of Beaver Valley's operating characteristics of youz 5..!
cnd history. Very truly yours, n . f 0m - ohn A. Levin Assistant Counsel , JAL/jem ,
. ,cs:
- Willia = T. Russell 9
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u - COV.MONWEALTH OF PENNCYLVANIA 1C,%g ;:ENNCYLVAN!A PUZLIC UTILITY COMMISSION , A.ik30 . . P. O. COX 3225. HARRICCUAG Pa.17120 . I
- Harch 27, 1980
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H:rold Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 , Re Request for NRC assistance in proceedings
- of the Pennsylvania Public Utility Comission
Dear Mr. Denton:
1 The Pennsylvania Public Utility Commission has currently before it i an investigation into the prolonged outages of the Beaver Valley Nuclear Generating Station. Unit 1. at its Investigation Docket No. I-79070314, et al. - i " I have spoken at some length with Mr. William T. Russell on various ( aspects of Isaver Valley's operation and conclude that it would be most helpful
.. ,f' embers cf the NRC st. a ff would appear to give testimony before our Commission i
cn technical =atters. . I therefore request your help in furnishing the assistance of members of your staff who have special knowledge of Beaver Valley's operating characteristics ' i end history. Very truly yours, , Q . t l (
- chn A. Levin .
Assistant Counsel , { t JAL/jem ,
,cs:" Ni111a= T. Russell 4
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UAY 2 1980 Docket No. 50-334 Mr. John Levin, Assistant Counsel Commonwealth of Pennsylvania Pennsylvania Public Utilities Commission P. O. Box 3265 llarrisburg, pennsylvania 17120
Dear Mr. Levin:
During the meeting held in Bethesda, Maryland on April 22,1980 we discussed topics that would be covered in the forthcoming depositions to be given by HRC staff members to be entered as part of the record of the Pennsylvania PUC proceedings on the outages of the Beaver Valley Nuclear Generating
. Station. Unit 1. It is our understanding that the NRC staff will provide testimony on safety issues related to turbine cracking, net positive suction head for safety related pumps and last year's shutdown orders regarding the seismic design of safety related piping systems. NRC personnel are not able to co.nent on issues relatiM to utility management efficiency but can comment on the quality of the cnd product of the utility's effort from a nuclear safety perspective.
Please contact me if you have any questions on this matter. OrlBi nal signed by William T. Russell, Chief Technical Support Dranch Planning & Program Analysis Staff Office of Nuclear Reactor Regulation cc: Sec next page OlSTRIBUTION c P Docket 50-334 4g A p 4* NRC PDR 0 LPOR DWiggington ELD d ky Dross wha O, #7 / * [F h' ,/ y~7# da W ulson 07 TERA s r0Arr4th _ p MR .: T. ,S B,: C , , ,, ,,, ,,E LD ,,,,,,, ,, ,,,,, ,,,, ,
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cc: Gerald Charnoff, Esquire Mr. James A. Worling . Jay E. Silberg Esquire Plant Superintendent Shaw, Pittman, Potts and Trowbridge Beaver Valley Power Station 1800 M Street, N.W. P. O. Box 4 Washington, D. C. 20036 Shippingport Pennsylvania 15077 Karin Carter. Esquire Mr. John A. Levin Special Assistant Attorney General Public Utility Connission Bureau of Administrative Enforcement P. O. Box 3265 Harrisburg, Pennsylvania 17,120 5th Floor. Executive House Harrisburg, Pennsylvania 1712C Mr. J. D. Sieber, Superintendent Mr. Roger Ta san of Licensing and Cogitance Stone and We> ster Engineering Duquesne Light Cogag Corporation - Post Office Box 4 P. O. Box 2325 Shippingport, Pennsylvania 15077 Boston, Massachusetts 02107 Irwin A. Popowsky, Esquire Mr. F. Noon Office of Consumer Advocate 1425 Strawberry Square R & D Center Harrisburg, Pennsylvania 17120 Westinghouse Electric Corporation Building 7-303 Pittsburgh, Pennsylvania 15230
- 5. F. Jones Memorial Library
- 663 Franklin Avenue Aliquippa Pennsylvania 15001 Mr. John Carey, Director Nuclear Operations Duquesne Light Cogag 435 Sixth Avenue Pittsburgh, Pennsylvania 15219 Mr. R. E. Martin Duquesne Light Cogag 435 Sixth Avenue Pittsburgh, Pennsylvania 15219
! Marvin Fein Utility Counsel
- City of Pittsburgh 313 City-County Building Pittsburgh, Pennsylvania 15219 l
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Docket No. 50-334 April 15, 1980 MEMORANDUM FOR: William T. Russell, Acting Chief Technical Support Branch Office of Nuclear Reactor Regulation FROM: Walter A. Paulson, Research Analyst Technical Support Branch Office of Nuclear Reactor Regulation
SUBJECT:
FORTHCOMING MEETING WITH COMONWEALTH OF PENNSYLVANIA PUBLIC UTILITY CO MISSION ON NRC PARTICIPATION IN PVC PROCEEDINGS Date & Tir.e: Tuesday, April 22,1980 " 10:00 a.m. Location: Room P-422 Bethesda, Maryland .
Purpose:
To discuss the extent of NRC participation in the Pennsylvania PUC proceedings en the outages of the Beaver Valley Nuclear Generating Station, Unit 1
Participants:
NRC Pennsylvania PUC D. Beckman, Region I, IE J. Levin. Esq. W. Hazelton W.A. Paulson . W. Ross W.T. Russell D. Wigginton
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((( LL /~) l w:,.w, u*- Walter A. Paulson, Research Analyst Technical Support Branch Office of Nuclear Reactor Regulation ec: D. Beckman, Region I, IE W. Hazelton W.A. Paulson W. Ross W.T. Russell D. Wigginton E. lordan, IE f 00 ? )q) JS
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-~ May 1, 1980 Docket No. 50-334 EMORANDUM FOR: William T. Russell, Chief yIjge Technical Support Branch Planning & Program Analysis Staff, NRR FROM: Walter A. Paulson, Technical Assistant Technical Support Branch Planning & Program Analysis Staff, NRR
SUBJECT:
SUMMARY
OF MEETING WITH PENNSYLVANIA PUBLIC UTILITIES COMMISSION , On April 22, 1980 a meeting was held in Bethesda, Maryland at the request-of the Pennsylvania Public Utilities Commission. The purpose of the meeting was to discuss the extent of NRC participation in the Pennsylvania PUC proceedings on the outages of Beaver Valley Nuclear Generating Station, Unit 1. A list of attendees is enclosed. The NRC agreed to provide testimony on technical issues related to Beaver Valley turbine cracking, net positive suction head concerns, and last year's shutdown as the result of concerns regarding the seismic design of safety related piping systems. The NRC attendees indicated that they would
. be unable to comment on issues related to utility management efficiency but cculd cc=ent on the quality of the end product of the utility's effort.
It is expected that the depositions of the NRC personnel will be taken during the period May 7-9, 1980, depending upon the availability of the Pennsylvania PUC counsel (subsequent to the meeting, the PUC counsel ** - requested a change in dates to May 20-22,1980). A
- l. C0 k i lc:!.'s.-
Walter A. Paulson, Technical Assistant Technical Support Branch Planning & Progam Analysis Staff, NRR
Enclosure:
As stated y>6# Y v
1 ATTENDANCE LIST l 1 MEETING OF NRC AND PENNSYLVANIA PUBLIC UTILITIES COMMISSION APRIL 22. 1980 l 1 NAME ORGANIZATION W. Paulson NRC/NRR D. Beckman NRC/IE, Region I ! W. T. Russell NRC/NRR W. J. Ross NRC/NRR/ DOR S. Sohinki NRC/0 ELD John A. Levin Public Utility Commission
- D. L. Wiggington (Part-time) NRC/ DOR 4
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* ~ April lS', 1980 DISTRIBUTION Docket 50-334 ELD NRC PDR TERA LPDR CParrish HRDenton Docket No. 50-334 EGCase Mr. John A. Levin, Assistant Counsel gh Commonwealth of Pennsylvania .Eisen1u WHazelton Pennsylvania Public Utility Commission WRoss P. O. Box 3265 DWigginton Harrisburg, Pennsylvania 17128 D8eckman, Reg. I,IE EJordon, IE
Dear Mr. Levin:
You 1etter to me dated March 27, 1980 requested that members of the NRC staff appear before your Commission to give testimony on technicalWematters have regarding the Beaver Valley Nuclear Generating Station, Unit 1. considered your request and will make staff members available to testify on matters related to the Beaver Valley turbine cracking problem, last year's shutdown of Beaver Valley as the result of concerns regarding the seismic design of piping systems and other safety related matters. Mr. Warren Hazelton and Mr. William Ross will be available to provide testimony on the turbine cracking problem. Mr. William Russell plans to provid2 testimony regarding the seismic design of piping systems covering the period March 1979 to present. Mr. Donald Beckman, Office of Inspection and Enfon:ement, who is the resident inspector at Beaver Valley, will provide testimony on the seismic issue leading up to the shutdown of Beaver Valley. Mr. David Wigginton, NRC Project Manager for Beaver Valley will provide support in these and other areas. I request that we adhere to the following procedures to accommodate the tight schedules of the NRC staff. I propose that the initial examination of the NRC witnesses be in the fann of a deposition in lieu of prefiled testimony. This deposition could then be incorporated into the record and the NRC witnesses would be made available for cross-examination upon the information provided in the deposition. This procedure was utilized by the Maine Public Utilities Comission in a similar proceeding and was beneficial in reducing NRC staff resources. Please contact Mr. William Russell of my staff at (301) 492-8041 if you need any additional information. Sincerely, S k Act Harold R. Denton, Director
' Office of Nuclear Reactor Regulation (C > ff hwe d?jd)< g S
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l i Mr. 00hn A. Levin l l cc: Gerald Charnoff, Esquire Mr. James A. Werling Jay E. Silberg, Esquire Plant Superintendent Shaw, Pittman, Potts and Trowbridge Beaver Valley Power Station 1800 M Street, N.W. P. O. Box 4 - l~ Washington, D. C. 20036 Shippingport, Pennsylvania 15077 Karin Carter. Esquire Mr. John A. Levin Special Assistant Attorney General Public Utility Commission Bureau of Administrative Enforcement P. O. Box 3265 5th Floor, Executive House Harrisburg, Pennsylvania 17120 Harrisburg, Pennsylvania 17120 Mr. J. D. Sieber, Superintendent Mr. Roger Tapan of Licensing and Conpliance Stone and. Webster Engineering Duquesne Light Ccepany Cor Post Office Box 4 P. D.poration Box 2325 -- Shippingport, Pennsylvania 15077 Boston, Massachusetts 02107 Mr. F. Noon Mr. C. N. Dunn, Vice President R & D Center Operations Division Westinghouse Electric Corporation Duquesne Light Company Building 7-303 435 Sixth Avenue Pittsburgh, Pennsylvania 15230 ' Pittsburgh, Pennsylvania 15219 B. F. Jones Memorial Library 663 Franklin Avenue Aliquippa, Pennsylvania 15001 Mi. John Carey, Director Reclea.- Operations Duquesne Light Conpany ~ 435 Sixth Avenue P ittsburgh, Pennsylvania 15219 Mr. R. E. Martin Duquesne Light Conpany 435 Sixth Avenue Pittsburgh, Pennsylvania 15219 Marvin Fein Utility Counsel City of Pittsburgh , 313 City-County Building ' Pittsburgh, Pennsylvania 15219 s . s
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2 ! opriCs or Twe April 30,1979 i SECRETARY ? 5
- j. MEMORANDUM FOR THE RECORD j i2 t 4 FROM: Samuel J. Chilk, Secretit ll H
i
SUBJECT:
STAFF BRIEFING ON FIVE F L SHUTDOWN (SEISMIC DESIGN), 3:00 P.M.
! THURSDAY, APRIL 26,197P, COMISSIONERS' CONFERENCE ROOM, D. C. ! 0FFICE (OPEN TO PUBLIC ATTENDANCE) i l 'The Comission discussed the status and progress of the five plants shut down on March 13, 1979 to permit reevaluation of facility piping systems i for seismic loads. No final actions were taken and no requests were made i during the briefing.
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V(, REVIEW 0F THE(SEISMI[ DESIGN CRITERI A FOR THE BEAVER VALLEY POWER STATION (Docket No. 50-33.4) Revised April 17, 1970 JOHN A. BLUME & ASSOCIATES, ENGINEERS San Francisco, California Q 4 g-L 1190 o
l . i* . REVIEW OF THE SEISMIC DESIGN CRITERI A FOR THE BEAVER VALLEY POWER STATION (Docket No. 50-334) This report summarizes our review of the engineering factors pertinent to I the seismic and structural adequacy of the Beaver Valley Power Station. The plant will be located in Shippingport Borough, Beaver County, Pennsyl-vania on the south bank of the Ohio River approximately one mile from Midland, Pennsylvania. Immediately west of the plant ~Is the Shippingport Atomic Power Station. The design and construction of the plant will be performed by Stone & Webster Engineering Corporation under the direction of the applicant,'the Duquesne Light Company. The nuclear steam supply systems will,be supplied by Westinghouse. A Prel'iminary Safety Analysis t Report has been submitted in support of the application to show that the plant will be designed and constructed in a manner which will provide for safe and reliable operation. Our review is based on the information pre-l sented in the Preiiminary Safety Analysis Report and Amendments, and is directed specifically towards an evaluation of the seismic and structural design of Class I structures, systems, and components. The list of re- ! ference documents upon which this review has been based is given at the end of this report. DESCRIPTION OF FACILITY l .
-The Beaver Valley Plant site is located on gently sloping river terraces on the south bank of the Ohio River. The plant'will be about 1000 ft southeast of the river; ground surface of the plant will be about 40 ft above normal river leve). Steep bluf fs which comprise the margins of the river valley rise to elevations several hundred feet above river level south and east of the site. The site is located in an area of river de-posits which were removed for construction of the plant. Underlying these ,
I JOHN A. BLUME & ASSOCIA'rES. ENGINEERS (- ~
of well consolidated alluvial gra-sedimentary deposits is a terrace vels. These gravels have a maximum depth of about 100 feet and rest The directly on the sandstone and shale bedrock of Pennsylvanian age. bedrock is bedded horizontally and is essentially undeformed. Foundation materials vary from gravelly soils to highly silty and clayey soils. Gravelly soils are predominant on the upper bench where the reactor structure will be located and relatively weak silty and clayey soils pre-dominate near the rivers edge where the cooling water intake f acilities In the area intermediate between th.e upper bench and the will be located. river where some major structural components will be placed the uppermost soils wi11 be excavated and replaced with compacted fill. The containment structure will be reinforced concrete cylinder and dome The interior of the supported on a reinforced concrete foundation slab. structure will be lined with a welded steel plate to ensure leak tightness. The ver-The inside diameter of the containment structure will be 126 ft. tical wall thickness will be 41 f t and the dome thickness will be 21 ft. The foundation slab thickness is 10 ft. STRUCTURAL DESIGN CRITERI A AND LOADS . Class l structures, systems, and equipment are those whose failure could They are cause or increase the severity of a loss-of-coolant accident. designed to withstand the appropriate seismic loads simultaneously with Structure design loads other applicable loads without loss of function. ' are increased by load factors based on the probability and conservatism of the predicted design loads. Yield capacity reduction factors are applied to the stresses allowed by the applicable building codes. pressure of The containment structure has been designed for a design The containment 45 psig. The maximum design temperature is 280 F. Loads structure is designed for a 40 psf snow or ice load on the roof. due to flooding are a factor only during construction of the containnent building.
. JOHN A st.UME & ASSOCIATES. ENGWEEW
)
Wind loads are to be determined for an 80 mph wind using the methods out-Ilned by the Task Committee on Wind Forces, ASCE Paper No. 3269 " Wind [ Forces on Structures". The structure will be designed for tornado loading which corresponds to a design tornado with a 300 mph tangential velocity, a 60 mph forward velocity, and an atmospheric pressure drop of 3 psi. Tor-nado generated missiles considered in the design will be among others, a 35 ft long utility pole at 150 mph. ADEQUACY OF THE-SEISMIC DESIGN CRITERIA
\
Ve have reviewed the Preliminary Safety Analysis Report and Amendments No. 1 through 15. In addition, we have discussed the various aspects of the seismic design of.the plant with the applicant and ' members of the. staff of of the Division of Reactor Licensing at meetings on January 28, 1970 and
~
karch18,19'70. We have the following comments regarding the adequacy of the. seismic design criteria: ,. 1 '. An analysis of settlements in the founding soils has been presented by the applicant and based upon this data the predicted settlements In addi:lon, the can be accomodated by suitable de. sign procedures. applicant has considered the possibility of densification of granular soils under vibratory loading due to earthquake and has indicated , that suitable precautions will be taken to prevent vibratory densi-fication.
- 2. One of the areas of concern is possibility .of liquefaction in those soils which consist predominantly of sand. The applicant has shown that the natural in place density and grading of the sand are not within the range characteristic of sands which have been found to be d
subject to liquefaction during earthquakes. The applicant has presented slope stability analyses developed for 4 3 satur:ted conditions and under earth eake loading. Based upon the JOHN A. BLUME & ASSOCIATES. ENGN
- RS
data which was submitted the slope design should be adequate with
~
respect to safety and integrity of Class I structures. It should be noted that saturation of soils can be achieved under conditions of thawing snow or prolonged heavy rain. fall without necessity of flooding. Therefore the assumption of saturated conditions is not overly conservative and combining this condition with the DBE is appropriate. .
. 4. Geologic reports state that the sedimentary strata beneath the site are very gently folded which indicates that the region has probab.ly not been subject to strong stresses for at least since the Pennsyl-vanian period, and is not considered tectonically active. The re-ports further indicate that no faults are known to exist at the site or in the surrounding region. The nearest fault is said to be loca-ted about.60 miles southeast of the plant site trending in a north-easterly direction.
- 5. Historically, no earthquake of epicentral Intensity greater than V
( has occurred within 80 miles of the site. The nearest earthquake A occurred at Sharon, Pa., about 40 miles north of the site, and had an epicentral intensity of MM lit or IV.' It is estimated that the New Madrid (1812) and Charleston (1886) earthquakes may have caused intensities of up to MM V at the site. The nearest areas of moderate seismicity are near Attica, N.Y. and Anna, Ohio, both about 200 miles from the site; earthquakes in these areas have had a maximum inten- ! sity of MM Vill, and apparently none of these has been perceptible - at the site. It is estimated by the applicant that the maximum bed- l 1 rock acceleration under DBE conditions would be 0.0359, which results in.a peak ground surface acceleration of 0.125g for the Design Basis Earthquake (DBE). A peak ground acceleratidn of.0.069 has been selec-ted for the Operatin'g Basis Earthquake (OBE). We concur with the se-1ection of these ground accelerations for the DBE and OBE. W'e also concur with the site response spectra for the DBE and OBE and the appilcation of these site spectra as proposed by the applicant in AppenJ:x B, page B.1-1, Amendment 15.
. .;oss A. st.u.iE & ASSOCMES ENQNr.! 8.S l
- 6. Tha epplicant has stated that he will use thn response spectrum
, method of dynamic analysis for Class I structures, piping, and equipment, and that discrete-mass-multiple-degree-of-f reedom mathe-matical models will be developed for the structural system. Possible variations in foundation material properties will be considered in ' the seismic analyses. Structures will be analysed for response in both the horizontal and vertical directions, and horizontal and ver-tical response spectra will be developed at the points of support of piping and equipnent. Ve concur in general with this approach. The analytical techniques proposed by the applicant are satisfactory and if properly implemented will result in a conservative design.
- 7. The applicant has proposed to use approximate techniques for the development of response spectra at support points to be used as input in the design of Class I piping and equipment within structures.
These methods are based on assumed motions at the support points of piping end equipnent. We have concern about the conservatism of the proposed method as compared to the more coimonly accepted time-history method. The applicant has presented a comparison of the i two nethods in ' Amendment 15 in which an "r" factor has been intro-duced. The comparisons presented in Amendment 15 do not demonstrate that the proposed approximate method pro' duces conse'rvative results. The appilcant has also stated in Amendment 15 that he will develop comparisons of response spectra computed by his proposed method and the time-history method for the Beaver Valley Plant. He has stated that these comparisons will be developed for the appropriate piping . and equipnent damping ratios and will utilize an input time-history that produces the closest match to the response spectra presented in the PSAR. We understand, based on discussions with DRL scaff, that the applicant will utilize spectra in the design, which will envelope the spectra produced by the proposed method and the time-history method. The applicant should also specifically state what methods
*w ill be used in this regard for Class I structures other than the containment structure.
JOHN A. BLUME & ASSOCIATES. ENGINEER 4S
I ( In summary, the applicant has outlined a program to substantiate that .; the proposed approximate method to develop response spectra at the support points of Class I piping and equipment is conservative as applied to the Beaver Valley Plant. We feel that the general overall program is acceptable. We understand that the applicant will be required to submit his demonstration of the validity of the proposed method to the AEC for review and approval prior to utilization of the response spectra curves in the final design. CONCLUSIONS On the basis of the information presented by the applicant in thd Preliminary Safety Analysis Report and Amendments, it is our opinion that the seismic design criteria and approach to seismic design as out-lined in the PSAR'and Amendments I through 15, If properly implemented by the applicant, will result in a design that is adequate to resist the earthquake conditions postulated for the site. JOHN A. BLUME & ASSOCIATES, ENGINEERS 4 4Afil.- Roland L. Sharpe hWlb & [arrison Kost l l JOHN A. B: UME & ASSOCIATES. ENGINEC NG I J
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. REFERENCES BEAVER VALLEY POWER STATION DUQUESNE LIGHT COMPANY (Docket No. 50-344)
Preliminary Safety Analysis Report, Volumes 1 through 4 Amendments No. I through 15 Boring Logs a _7
. JOHN A. BLUME & ASSOCIATES. ENG'NECAS
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[pm trO\,, I. .: LEAR REGULATORY COMMISSIO [ ri WASHINGTON, D. C. 20555
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- e CNp# March 14,1979 OFFICE OF THE SECRETARY MEMORANDUM FOR: Lee V. Gossick 3
- ~ Executive Director for Oper 1 ns 1
FROM: SamuelJ.Chilk,SecretaryJh
SUBJECT:
STAFF REQUIREMENTS - DISCU S: MNOFSEISMICDESIGN PROBLEMS IN CERTAIN PLANTS, 8:55 A.M., TUESDAY, MARCH 13, 1979, COMMISSIONERS' CONFERENCE ROOM, D.C. OFFICE (Closed to Public Attendance) The Commission unanimously concurred in the recomendation of the Director of Nuclear Reactor Regulation that Orders to Show Cause should be issued to utilities operating Surry Units 1 & 2, Beaver Valley Unit 1, the Fitzpatrick Nuclear Power Plant, and the Maine Yankee Atomic Power Plant. The Orders would require the utilities to shut down these plants within 48 hours until it is determined that any needed modifications are made in safety-related piping systems to bring them into conformance i with NRC seismic requirements. (NRR) (The Orders were issued later in the day on March 13,1979) The Commission requested that:
- 1. the Director of NRR consult with the Commission prior to issuing any further Order permitting any of the plants to resume operations; (NRR) (SECYSuspense: As required by circumstances)
- 2. an estimate of the economic costs of the suspended operation of the lants be provided; five p(OPE) (SECYSuspense: March 27, 1979)
- 3. the Director of NRR should contact the licensees by telephone to inform them of the shutdown action prior to issuance of the Orders and to obtain from the utilities an oral analysis of impact of the shutdowns on area electricity supplies; (NRR) .(The utilities were contacted by telephone later in the day on March 13,1979)
- 4. the Governors of states affected by the shutdown orders be informed prior to issuance of the Orders; (SP) (State Liaison Officers in 5 states were informed of this action and requested to inform their respective governors later in the day on March 13,1979) -J E
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- 5. a full chronology of events leading up to today's action be provided.
The chronology should include a discussion of the circumstances leading to Westinghouse's discovery that certain valves weighed more than previously assumed in the Beaver Valley facility's piping system analysis, as well as the generic implications of that discovery. (NRR) (SECY Suspense: March 21,1979)
- 6. this particular matter should be carefully reviewed to determine potential generic implications for the treatment of seismic loadings
. in other plant designs; (NRR) (SECY Suspense: April 25, 1979)
- 7. a review of the circumstances through which the problem, its nature, and the origins were made known to the staff, and the implications for future technical review procedures of these cir-dumstances.
(NRR) (SECYSuspense: March 27,1979)
- 8. the transcript of this meeting be reviewed promptly to determine whether the entire text, or any portions thereof can be released to the Public Document Room.
(0GC/SECY) (SECYSuspense: March 14,1979) The Commission also reminded the Executive Director for Operations and the staff of their obligation to keep all Commissioners informed in a timely manner of similar problems of such potential magnitude. cc: Chairman Hendrie Commissioner Gilinsky Commissioner Kennedy Commissioner Bradford Commissioner Ahearne Acting Director, Policy Evaluation Director, Congressional Affairs Director, Public Affairs Director, Inspector and Auditor Osc-
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From: R. T. Kennedy /M.?i '
Subject:
SEISMIC EVALUAT10 tis OF 5 flVCLEAR PLANTS c n
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. )fl )I I would like a statement from the staff as to whet ef, at the time U that the operating license was issued for each of he 5 plants affected by the recent review of the safety imp) cations of a j .
4 varying chalytical code fEr 5eismic-induced p}g5fng stresses, the if k k,/ data presented understand and them, metthethe analytical llRC (AEC) techniques safety used, as we nowdquirements. If the fy answer to the question is "no", a full .exp anation of the matter
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a brief discussion of how our requirements have evolved since that time should be provided. l
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- A. Kenneke, OPE i L. Bickwit, OGC }ld.ed' Q h w.u; .
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(.2 x 2. p.7>) . a full chronology of events leading up to today's action be provided. The chronology should include a discussion of the circumstances ; leading to Westinghouse's discovery that certa.in valves wei more than previously assumed in the Beaver Valley facility'ghed s piping
)#b vp system analysis, as well as the generic implications of that discovery.
(NRR) 45ECY__c.ugr a*+ --4' = ' " 11- 1070i r - 6. this particular matter should be carefully reviewed to detemine potential generic implications for the treatment of seismic loadings I in other plant designs; (NRR) (SECY Suspense: A@ M - WM_.
- 7. the transcript of this meeting be reviewed promptly to determine ;
whether the entire text, or any portions thereof can be released to -
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the Public Document Rcom. (,' (OGC/SECY) (SECY Suspense: March 14, 1979) The Comission also reminded the Executive Director for Operations and the staff of their obligation to keep all Comissioners infomed in a timely manner of similar problems of such potential magnitude, cc: Chairman Hendrie Conrnissioner Gilinsky Commissioner Kennedy Commissioner Bradford Comissioner Ahearne Acting Director, Policy Evaluation ofrector, Congressional Affairs Director, Public Affairs ' Director, Inspector and Auditor A review of the circumstances through which the problem, its nature, and its origin were made known to the staff, and the implications for future technical review procadures of these circumstances. 1
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,, f UNITED STATES y 'I s NUCLEAR REGULATORY COMMISslON 5l, 's..Is-) a y e WASHINGTON, D. C. 20555
%;..eje'...f March 15, 1979 Docket No. 50-334 LICENSEE: Duquesne Light Company FACILITY: Beaver Valley Power Station, Unit No.1
SUBJECT:
MEETING
SUMMARY
- PIPE STRESS CALCULATIONS USING "PIPESTRESS" The licensee and Stone and Webster Engineering Corporation met with the staff on March 8,1979 to discuss the pipe stress calcu-lations and methods as reported by R. R. r,eimig in a memo to E. L. Jordan dated January 18, 1979. This memo and follow-up review had been transfered to the Division of Operating Reactors by E. L. Jordan in a memo to R. H. Vollmer dated February 23, 1979.
The attendance list is attached (Attachment 1). The Stone and Webster Engineering Corporation (S&W) discussed the .
! PIPESTRESS code and selected runs to demonstrate features of the code. They also presented a comparison of a number of calculations on selected pipe systems to compare PIPESTRESS and NUPIPE. PIPE-STRESS, a proprietary S&W code, was developed and used during the 1960's and early 1970's.
During the review of the PIPESTRESS code, S&W infomed the staff that the seismic stresses were computed in the SHOCK 2 subroutine using an algebraic intra-modal component summation. This summation, in effect, accounts for both positive and negative components but algebraically adds the two. The resultant component from which the seismic stresses are computed can be significantly less than is thought to be appropriate for that location. Regulatory Guide 1.92 published in 1974 lists the two acceptable methods as absolute summations or the square root of the sum of the squares method. The licensees and S&W were then advised that this matter deserves immediate NRR management attention due to the potential for over-stressed piping systems at Beaver Valley as well as other vintage pl ants.
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I 9 Meeting Sumnary for , Duquesne Light Company March 15,1979 . Following a break in the meeting at noon, at which time NRC management A) i was first notified of the putential problem, the meeting continued ~,i with the licensee, S&W, and NRC management. In view of not having ;.
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any specific re-analyses of the Beaver Valley Unit No.1 piping, - NRC management requested an immediate effort by the licensee and S&W to identify the piping systems involved and scope the inadequacies 4 by Wednesday, March 14 and reanalyze the effected systems by Friday, a March 16. S&W committed to identify the other plants using :5 PIPESTRESS - SH0CK2 by Monday, March 12 and to notify those :N utilities as soon as possible. The licensee was directed to inform 3 his Safety Review Committee as soon as possible to assure continued health and safety of the public. f ! 1 Dave Wigginton, Project Manager Operating Reactors Branch #1 Division of Operating Reactors Attachments: - As Stated cc: w/ attachments See next page I 4 1 A d l 4
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Meeting Sumary for March 15,1979 Duquesne Light Company Docket File NRC PDR Local PDR . ORB 1 Reading NRR Reading , H. Denton E. Case V. Stello D. Eisenhut B. Grimes D. Davis D. Ziemann P.-Check G. Lainas A. Schwencer R. Reid T. Ippolito V. Noonan J. McGough Project Manager OELD OI&E(3) . ACRS(16) C. Parrish NRC Participants TERA J. R. Buchanan Licensee
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. I MEETING ON STONE AND WEBSTER ENGINEERING CORPORATION j PIPE STRESS CALCULATIONS WITH DUQUESNE LIGHT COMPANY MARCH 8, 1979 Attendees Dave Wigginto'n NRC ORB 1 John Lynch DLC - ENG J. J. Carey DLC J. M. Cumiskey S&W BVPSl Project Engineer Don King S&W Engineering Manager !
Dana Shave S&W Supervisor i G. L. Harper S&W Lead Engineer R. G. LaGrange NRC D0R EB S. Hosford NRC 00R EB John Fair NRC 00R EB R. Vollmer* NRC DDR AD S&P A. Schwencer NRC D0R ORB 1 . V. Noonan NRC D0R EB K. Wichman NRC EB V. Stello NRC 00R , and others I
- Attendance at the afternoon management meeting, some part time.
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- April 13, 1979 MEMORANDUM FOR: Chairman Hendrie Comissioner Gilinsky Comissioner Kennedy
, Commissioner Bradford Comissioner Ahearne Al Kenneke, Acting Director, OPE [ _
A N-[ v FROM: TF
SUBJECT:
ANSWERS TO COMMISSIONER AHEARNE'S QUESTIONS: RE FIVE REACTOR SHUTDOWN Attached are the answers to questions posed in Comissioner Ahearne's memo- , randum of March 14. Progress in re-analysis work being done by Stone and Webster (S&W) for the licensee has been much slower than the schedule shown in that memo. High-lights of efforts to date are:
-- The licensees' initial analytical efforts have adhered closely to a narrow interpretation of the Comission's Show Cause Order; i.e. re-analysis using SH0CK 3 or NUPIPE of the specific safety-related piping systems originally run on SH0CK 2. The analytical work is aimed at determining whether or not the pipe stresses and restraints are within code-allowable levels.
-- Where stresses exceeding allowable are encountered, the licensees perform additional detailed analyses rather than comit themselves to any hardware changes. (More apparently is involved than costs of hardware changes per ,se. They are apparently concerned that the design changes for new hardware would constitute a plant modification which would require a hearing before re-start -- i.e. lengthening of the shutdown.)
-- Substantial effort by S&W and the licensees has gone to verifying that computer input data accurately represents the piping configurations as shown on the engineering drawings. -
-- Code verification efforts of the NUPIPE and SH0CK 3 programs by the NRC staff has been underway for several weeks. Among other things, this involves running several benchmark problems (done for NRC using EPIPE at Brookhaven) at S&W using their codes. This work should be done by the end of next week (around 4/20).
CONTACT: Dennis Rathbun (OPE) 634-3295 - s sp9n w .
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0 For the Comission :
-- Although the licensees have pressed hard on what results would be accept- i able, NRR's position has been (and continues to be) that the licensees bear the responsibility of proving to the NRC staff the validity of whatever analytical and/or technical fixes they may wish to propose.
-- Neither the licensees nor the NRC staff are examining the broader risk
-- cost / benefit question, i.e. analyzing in detail the risks of interim operation taking into acount both the probability and consequences of ,
seismic events weighed against economics and other costs associated with : continued shutdown. This is at least in part attributable to the extreme difficult of predicting the actual consequences of severe seismic events. It appears that there is littl5 that the Comission could do to expedite the process should it choose to -- since the pacing item is essentially i determined by the rate of re-analysis of the piping systems by S&W in Boston. .i According to the licensees' recent submissions, there are over 200 computer reruns involved (there could be more if NRC staff reviews flag additional safety related piping systems). Most of the mechanical operations in re-running the programs are probably done now. However, S&W engineers must review each run in detail; the analyses are also reviewed by the licensees before submission to the NRC staff. At this point, the NRC staff has received and is nearing completion of its reviews for five piping systems packages (those for Maine Yankee). Assuming no major problems in staff rev-iews, no hearings prior to startup, and no hardware changes (truly "best case"), the following might be the startup schedule: Maine Yankee -- end of April Beaver Valley -- mid/ late May Surry -- late May/early June Fitzpatrick -- late May/early June We understand that the NRC staff should be ready to discuss results to date (particularly for Maine Yankee) in a-briefing next Friday (4/20). As a 1 contingency that the rough estimates of startup turn out to be overly
- optimistic, the Comission may wish to discuss two possibilities with the staff at next Friday's meeting:
-- Whether there might be a point in the weeks ahead where staff confidence (based upon (a) results of the code verification work now in progress and (b) completion of a greater number of reviews) might justify recom-mending interim operation pending completion of the re-analysis efforts.
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r-For the Commission -- Whether site specific seismic work to estimate earthquake probabilities would be of utility for any of the plants. (Note: In particular, PASNY contends in its March 30 submission that the Fitzpatrick site is "... generally considered seismically inactive," but we understand that the
. licensee is not now pursuing technical analyses to support this position.
This would not necessarily be easy for either the licensees or the staff to do in a short time since we understand significant detailed on-site geotechnical work would be required -- and for this reason not much may be gained from pursuing this course.) OGC prepared the answers to questions four and five, in which ELD concurs. Enclosu s: As stated L - . h cc: Leonard Bickwit Sam Chilk - 6 9 e e
NN UNITED STATES [M t NUCLEAR REGULATORY COMMISSION .) i ,% WASHINGTON, D. C. 20656 '
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,/ April 18,1979 N..-!< ...
l Docket Nos.: 50-250 ; 50-251 LICENSEE: FLORIDA POWER AND LIGHT COMPANY FACILITIES: TURKEY POINT UNITS 3 and 4
SUBJECT:
SUMMARY
OF MEETING HELD ON APRIL 13, 1979, TO DISCUSS SEISMIC ANALYSIS OF TURKEY POINT REACTOR COOLANT PIPING On April 13,1979,the NRC staff met with representatives of Florida Power and Light Company (FP&L) and Westinghouse Electric Corporation (W) to discuss the seismic analysis of reactor coolant piping at Turkey' Point Units 3 and 4. A list of attendees is attached. Highlights of the meeting are summarized below. Introduction The meeting was requested by the NRC staff to gain a better understanding promptly of the original seismic analysis and recent reanalysis conducted by W for Turkey Point 3. Following the disclosure that Stone and Webster Engineering Corp. had used a computer code for seismic analysis of piping that combined intra-modal stress combinations by the algebraic summation technique, the NRC staff conducted a survey to find out if this technique had been used by others on plants other than those designed by Stone and Webster. Initial pre-Ifminary response from W was that they had not used such a method. How-ever, in checking further, it was discovered that a W computer code WESTDYNE) l used the algebraic sum technique until 1971 at which time the code was l modified to combine seismic stress combinations by absolute sum value technique, an acceptable method. l Since the NRC staff had earlier been informed that W had not used algebraic sumation, and more recently learned that they had Tn fact used this method and that a reanalysis had been done for Turkey Point, the NRC staff re-quested that this matter be discussed and resolved in a formal meeting prior to further operation of Turkey Point 3, which was conducting low-power physics tests in preparation for power operation. I WESTDYNE is a W proprietary code adapted from ADLPIPE. gp, r >
Discussion l As applied to Turkey Point, WESTDYNE with algebraic sumgtion is the code This was reported of record for only the reactor coolant main loop piping. to FP&L by W on April 9, 1979. On April 10, 1979, FP&L requested W reanalyzetTiereactorcoolant(largebore)pipingusingtheWESTDYiiEgo code, but corrected by using the absolute sum technique. The results were completed on April 12, and are sumarized in an attachment. The results show essentially no changes in stress levels from thost; reported and accepted in the Turkey Point FSAR(FSAR Table SA3, Ameidment 24, November 24, 1970). The main loop piping has very little coupling between the horizontal and vertical modes so the method of combination has an insignificant effect on the results. In addition for this piping, seismic loads are small when compared to LOCA loads (LOCA is the limiting load condition). P.ipe support loadings were also examined and found to be essentially unchanged in the reanalysis. The reanalysis applies to both units at Turkey Point. Conclusions and Followup Action At the conclusion of the meeting, the NRC staff told FP&L that the re-analysis was acceptable, and that Turkey Point 3 could continue operation. - FP&L was requested to send NRC a letter documenting the reanalysis results for both Turkey Point dockets 4 W is checking on other plants which may have been analyzed using the earlier version of WESTDYNE. Charles M. Trantnell Project Manager Attachments:
- 1. List of Attendees
- 2. Letter from W (Anderson) to NRC (CaseT dtd 4-12-79 2
The pressurizer and spray line piping have been analyzed with WESTDYNE, but with a post 1971 version using the absolute sum technique, khe current version of WESTDYNE has been revtewed and approved by NRC, although it has not been verified with benchmark problems. 4 This was done the same day, April 13, 1979. ,
Meeting Summary for Florida Power and Light Company Docket Files NRC PDR Local PDR ORB 1 Reading NRR Reading H. Denton E. Case V. Stallo D. Eisenhut B. Grimes R. Vollmer A. Schwencer D. Ziemann P. Check G. Lainas D. Davis B. Grimes T. Ippolito R. Reid V. Noonan G. Knighton - D. Brinkman Project Manager OELD OISE (3) C. Parrish ACRS (16) HRC Participants J. Buchanan TERA Licensee Short Service List i
I Robert E. Uhrig Florida Power and Light Company cc: Mr. Robert Lowenstein, Esquire Lowenstein, Newman, Reis and Axelrad 1025 Connecticut Avenue, N.W. Suite 1214 Washington, D. C. 20036 Environmental and Urban Affairs Library Florida International University Miami, Florida 33199 Mr. Norman A. Coll, Esquire Steel, Hector and Davis 1400 Southeast First National Bank Buildins . Miami, Florida 33131 Mr. Henry Yaeger, Plant Manager Turkey Point Plant Florida Power and Light Company P. O. Box 013100 Miami, Florida 33101 Mr. Jack Shreve Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304
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f LIST OF ATTENDEES FP&L MEETING APRIL 13, 1979 Florida Power and Light Westinghouse D. Whittier R. Sero
- 0. Pearson R. Brandon S. Brain NRC Lowenstein, Newman, Reis E. Case Axelrad & Toll R. Mattson N. Moseley M. Axelrad R. Bosnak
-V. Noonan A. Schwencer W. Russell C. Trammell K. Herring M. Hartzman E. Sullivan K. Wichman J. Fair R. LaGrange
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' April 12,1979 NS-TMA-2066 Edson G. Case Dsputy Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Comission 7920 Norfolk Avanue Bethesda, Maryland 20014 Dear Mr. Case.
TURKEY POINT SEISMIC ANALYSIS At the request of Florida Power and Light, a detailed file search was conducted of piping analyses performed by Westinghouse for the Turkey Point Units. Three anaylses were performed by Westinghouse including a seismic analysis of the reactor coolant loop, the pressurizer surge line, and the pressurizer spray line. The analysis of the loop was performed using the algebraic summation technique for intramodal responses. The analyses of the pressurizer surge and spray lines were performed using the absolute summation technique. A reanalysis of the Turkey Point loop has been perfonned incorpor-ating the absolute sumation of intramodal responses. The results are presented in the attached table. The results from the previous analysis, which were reported in Revision 9 of the FSAR, page SA-20, are also shown for comparison. As with the original analysis, both horizontal and vertical components of the seismic response spectrum were input simultaneously. Two different directions of the horizontal component were chosen and the results reported were for the most severe loading condition. l As can be seen from the table, the magnitude of the stresses are l essentially unchanged, and stresses are well below allowables. The method of combination has little effect on the pipe stresses due to - the lack of coupling beNeen the horizontal and vertical modes of the
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." COMPARISON OF SEISMIC STRESSES Location Maximum Stress , psi Previous Reanalysis Analysis Reactor Coolant Pump Inlet 4085 4100 Reactor Coolant Pump Outlet 3616 3700 10 Inch Accumulator Line 3201 3300 Steam Generator Outlet 2274 2300 Reactor Vessel Inlet 1289 1300 Reactor Vessel Outlet 182 200
- Pressurizer Surge Line Connection 78 100 Steam Generator Inlet 71 100 Maximum allowable seismic stress = 13,125 psi 1
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- ":.""la.t.L'O Mr. Harold Denton January 8, 1980 Director ,
Office of Nuclear Reactor Regulatica U. S. Nuclear Regulatory Commission Washington, D. C. 20555 .
Dear Sir:
This letter supplements.the information which was provided to you by our January 3, 1980 letter regarding the status of Shock 2 computer calculations which were performed for 6 inch and under piping for the Surry 1 and 2,
?bine Yankee and J. A. FitzPatrick Proj ects.
The information contained herein addresses only 6 inch and under piping configurations that were not analyzed as branch piping (branches from piping greater than 6 inch in diameter) as a part of the Show-Cause Reanalysis effort. Surry 1 and 2 - The review of all available pre-1979 records
.. disclosed no evidence that any Shock 2 computer '
calculations exist other than those which were reanalyzed as a part of the Show Cause reanalysis effort. t.aine Yankee - As previously reported, there are a number of Shock 2 computer calculations that were made for 6 inch and under piping. Al1 of these were reanalyzed as a part of the Show Cause reanalysis. J . A. Fit zPa trick - All of the 21s to 6 inch piping identified in compliance with IE Bulletin 79-14 is being re-evaluated and/or reanalyzed with only one piping problem remaining to be completed. We have verified that none of this piping was previously analyzed using the Shock 2 computer program. Based on this information, we conclude that there are no outstanding Shock 2 computer program related questions for the piping discussed above. Very ruly yours,
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- Merech 9, 1979 Docket No. 50-334 MEMORANDUM FOR: Richard H. Vollmer, Assistant Director for j Systems and Projects, D0R FROM: A. Schwencer, Chief Operating Reactors Branch #1, D0R
SUBJECT:
BEAVER VALLEY POWER STATION, UNIT NO.1 TELECON DISCUSSION REGARDING POTENTIAL PIPE SIRESS PROBLEM During the meeting on March 8 1979, Duquesne Light Company (DLC) h and Stone and Webster Engineering Company (S&W) committed to pro-viding on March 9 a schedule for development of further infonnation on the use of PIPE STRESS at the Beaver Valley Power Station, Unit No. 1 (BVPS). This schedule was provided by DLC and S&W just before noon today. DLC and S&W are both mobilizing staff to review the BVPS calculations on a high priority basis and intend to meet the schedule we proposed on March 8. S&W cautioned that this is a tough job and there could be problems as they get into it. As of now there have been no reasons to believe they cannot meet this schedule. However, since they are working over the weekend, they will give us a status report on Monday, March 12 if such problems develop. On Wednesday, March 14, DLC and S&W will provide a list of Q-1 and Q-2 piping that has been reviewed and identify by building and pipe run, those that have vector errors. A remarks column will identify the expected magnitude of the resultant error. On Frid6y, March 16, DLC and S&W will provide an evaluation of the BVPS safety and a proposal < for any required modifications. This evaluation will include revised stress results concluded to that point. Since S&W is assuming they will find few problems, they expect these calculations will be completed by Friday. In addition, on Monday, March 12, S&W will identify other S&W designed plants which have used PIPE STRESS. DLC stressed the point that they were actively pursuing this matter by sending staff to Boston to assist in the review and to have a first hand analysis of the problem. The offsite safety review committee has not reconvened on the matter, however, members have been contacted individually and a quorum has agreed to quickly consider the potential impact on safety of the additional information now being developed by S&W. 1 [39
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l 1 Mr. Richard H. Vollmer - 2- March 9,1979 The offsite safety review conmittee had met on this matter previously and had issued questions to S&W. 1 Or A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors e 6 e O 5 l t l
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QUESTIONABLE PIPE STRESS CODE USED BY STONE & WEBSTER ON FIVE PLANTS 4 In following up on a Licensing Event Report on Beaver Valley, D0R learned on March 8,1979 that Stone & Webster used an unconservative version of its PIPESTRESS code to calculate stresses on pipes and pipe supports. Algebraic summation was used instead of the accepted conservative techniques of summation of absolute values or taking the square root of the sum of the squares of the absolute values (SRSS). The code version used resulted in predicting stresses significantly lower than would be predicted by accepted conservative techniques. Use of a newer code, NUPIPE, considered to be acceptably conservative, yielded unacceptable stresses in two pipe runs. Because of the potential generic aspects of this matter, Duquesne Light Company and. Stone & Webster were asked to identify whether any other safety related piping systems could have unacceptable stresses when analyzed by NUPIPE or another code acceptable to us. Further, Stone & Webster was asked to identify any other plants which may have used the nonconservative version of PIPESTRESS to calculate stress levels. Four other plants are involved:
. Fitzpatrick, Maine Yankee, and Surry Units 1 and 2.
Efforts were initiated by Stone & Webster on March 9,1979 to obtain the requested information on an expedited basis. D0R dispatched technical staff to the Stone & Webster Boston offices, as did Duquesne Light Company, to monitor this effort. Also on March 9 D0R called the other licensees to inform j them of this potential problem and to request that they also inform the NRC i of any safety systems that could be adversely affected in addition to their bases for continued operation of the plants. A preliminary response from all licensees has been requested by Monday, March 12. At the present time three
. of the five plants are operating and two are shut down. (Surry Unit 2 is l shut down for steam generator replacement. Beaver Valley shut down March 9
- and Duquesne Light Company has indicated that' the plant will not be restarted j without prior notification of DOR.)
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The Honorable Morris K. Udall, Chaiman Subcomittee on Energy and the Environment Comittee on Interior and Insular Affairs -- _e. _ United States House of Representatives - -.
-~- 4 --
".".7 Washington, D. C. :20515
Dear Mr. Chaiman:
Your letter dated December 3,1979 requested clarification of previous NRC stataments regarding the use of the algebraic summation technique in the seismic design of nuclear power plants. Following receipt of your letter, Dr. Henry Myers of your staff contacted individuals in the NRC's Office of
. Nuclear Reactor Regulation to discuss this issue. In response to Dr. Myers' request, the enclosed letter from Darrell Eisenhut was sent on December 12, 1979. ,
The NRC's position is that from a mathematical and engineering standpoint, the use of the algebraic sumation of incoherent forces and displacements obtained by spectral analysis in the seismic design of nuclear power plants is incorrect. Mr. Denton's letter of Nove=ber 28,1979 to you contains' a more detailed di:::::icn cf the background of this matter and the results of our current analyses. - Sincerely.
/5/
l John F. Ahearne O
Enclosures:
- 1. D. Eisenhut letter to Dr. Henry Myers, December 12, 1979 g
d 'O 1 N er 8 97 cc: Rep. Steven Synns
- Clearec with all Cmrs. ' Offices by SECY C/R -,
. Typed in final in the Office of the Chairman ./ 1 ta incorcorete tws.' corrents .+/ b . .cc ) . .E..D. 0./.N. .R.R.
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- U.S. GOV ERNMENT P8tlNTING OF FICE: 1979 299 M9
- . _ _ _ . - - , _ _ _ _ _ . . _ _ . _ _ _ _ . ~ _ . _ _ _ . _ - . -
wne e e.w e e n . 3- ,, e 3 v. h NUCLEAR REGULATORY COMMISSION
~
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December 12, 1979 c 3 3 5 5 Dr. Henry Myers, Special Consultant i on Interior and Insular Affairs 1324 Longworth House Office Building Washington, D. C. 20515 i
, y
Dear Dr,
. Myers: - [ t Pcr our discussion following yo'ur request, the following paragraohs should f
- clarify the use of the term " error" .in the NRC's March 19, 1979 testimony. g E
"In~our March 19, 1979 testimony we stated that we considered the p use of the algebraic sumation method to have been in " error" as j used by Stone and Webster in a computer code used for the design L of piping systems under seismic events. Since presenting that +
testimony we have learned that the algebraic suct.ation was used in five different computer programs and that this technique _ was used in the design of 29 nuclear facilities. These facilities are both in operation and under construction. Because of these - findings it is unclear whether or not the designers of the facility should have recognized their mistake. . - Independent of whether this should be considered an " error", today's understanding of these calculational techniques clearly-indicate that the use of the alcebraic summation is incorrect. We cenclude this because the algebraic technique can in some cases sicnificantly uncerestimate the seismic resconse of piping, while in other cases it may produce results similar to acceptable methocs." 5 Sincerely,
\ .
t l, I I
- Ol h [ ./ .
Darrell G. _.tsenhu
/Mb.UM.
, Ac.ing Director
- Division of f0ceratinc c.eactors i Office of Nuclear Reictor Re;uiation l
l ga v Do 1
r' Arthur D Littkinc. ,conen<.u ance.c u,muo., sir, ~ s770.mn wce April 19, 1979 Mr. Vincent S. Noonan, Chief Engineering Branch Division of Operating Reactors U.S. Nuclear Regulatory Commission
- Washington, D.C. 20555
Dear Mr. Noonan:
98705 I am enclosing a memorandum which confirms the information furnished at a meeting with you and other members of the NRC Staff Monday afternoon, April,16, 1979. I am sending a copy of this letter (and its attachments) to John G. Davis, Acting Director, Office of Inspection and Enforcement, under cover of transmittal, a copy of which is attached for your information. A copy of this letter and its attachments are being sent to the organizations listed in Appendix II, who are ADLPIPE licensees. As discussed at our April 16 meeting, we will verify the five bench mark problem solutions (after receipt of the problems from NRC) published in ENL-NUREG 21241-RS and BNL-NUREG-23645 utilizing the present version of ADLPIPE, February 1977, Version 3C. If you desire any further information, do not hesitate to call. Very truly yours, O I. W. Din 11 sp Enclosures Menorandum Letter to John G. Davis from I. W. Dingwell of 4/19/79 [d enin mr.u. u n,s: Af t(P4S DT10STLS l ONIJON MALVMt3 PAFUS IPb 1 ; lAN' fM "A*4fflAfdJie t ?n*a*I ' 'l ii *J " ' A 'a *4 '"i a l e +a V,4 9 M 8'l
9 N. 04hN .9 Arthur D Littie,Inc. aconn eens.,- .,nn,oos .e . ,,:.,40.,c.,,, 2 stro.r<<<,22,43s April 19, 1979 Mr. John G. Davis Acting Director Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Mr. Davis:
I am enclosing a copy of letter and memorandum dated today which I have sent to Mr. Vincent S. Noonan, Chief, Engineering Branch, Division of Operating Reactors. The information contained in them confirms information which I furnished at a meeting Monday after-noon, April 16, 1979, at the NRC with Messrs. Noonan, Russell, and other members of, the NRC staff. I do not believe that the information contained in the enclosures is reportable under Part 21 of NRC's regulations. I recognize, however, that a contrary interpretation of Part 21 might be made. Accordingly, I request that the enclosed letter and its attachments be regarded as a report under Section 21.21 to the extent that they might be regarded as containing information reportable under Part 21. Very truly yours, 4; " " I. W. Dingw611 sp Enclosures Letter to Mr. Noonan from 1. W. Dingwell of 4/19/79
, Memorandum to NRC of 4/19/79 gfN A fl e PM stelt W' t % 4 DP414 N M AIacil I'Al W. li ~ti ' '**de 'l * * ** 1 'la . .
* -* '4" * '*3 68 *'*.
e
MEMORANDUM April 19, 1979 A Brief History of ADLPIPE (see table 1) Arthur D. Little, Inc., first prepared a program in 1952 to compute the flexibility and thermal deformations of piping systems for a private firm. An ASME paper was delivered in Apr*1 1956, "The 6X6 Matrix Method of Piping System Stress Analysis". Later during the liquid oxygen fueled ballistic missile program, Arthur D. Little, Inc., adapted this program to make dynamic analyses of missile fueling systems. A new program, ADLPIPE, was developed in the period 1967-1968, first for the static (deadweight, thermal, external force, applied displace-ment) analysis of elastic piping systems. The program was written in FORTRAN and designed to be independent of the particular computer system used. The second development--also in 1968--(modification one) was for the dynamic (modal) analysis of lumped mass piping systems. The transient loading was described as a response spectra. Following a prototype development period, a version was delivered in August 1970 which enabled the user to implement ANSI B31.7 " Nuclear . Pouer Piping". This version could not produce a full stress report but gave stresses for particular loadings. In 1972 a version was released which enabled the user to produce a partial stress report to meet the ~ requirements of ASME Section III. In 1972 this version was released to Control Data Corporation Cybernet. In 1973 the computation of fatigue usage factors was completed. In '1974 a version was released for the utilization of ASMC Section III, Class 2. In 1975 a force time history analycic wac included for the calculation of hydraulic transients. At the same time a one-dimensional thermal transient analysis was developed for the requirements of ASME Class 1. In 1976 the automatic computation of seismic analyses in accordance with Regulatory Guide 1.92 was developed and checked. The complete matrix analysis portion of the program was rewritten based generally on the . techniques of SAP IV with some improvements in the matrix storage methods. In addition, a post-processor was developed which allowed the user to make load set combinations for use in applications other than Regulatory Guide 1.92. This version was released in February 1977 and upgraded in December 1977 and September 1978. In the period 1968 to 1973, ADLPIPE was the only computer program (which was available to the public) for computing piping response to various static and transient loads. Other programs were in use, but to our knowledge, these were proprietary and not available for general use. From its inception, ADLPIPE could be utilized for a variety of stress calculations not involving nuclear power piping. In 1975 applications were extended to meet the requirements of chemical plant and refinery piping and petroleum transportation piping. Arthuri)latielnc
TABLE I DEVELOPMENT OF ADLPIPE 1967 Development of static load version 1968 Delivery of static version Delivery of prototype dynamic version 1969 ---- -- 1970 Delivery of static dynamic !!31.7 version 1971 ----------------- 1972 Inclusion of ASME Section III Class 1 . Inclusion of closely spaced modes 1973 Inclusion of AS!!E Section III Class i usage factors Inclusion of Metric units 1974 Inclusion of ASME Section III Class 2 and 3, B31.1 1975 Revised input organization
, 1976 Force time history analysis Transient thermal analysis (one-dimensional) 1977 Inclusion of 1.92 modal summation (group method)
Inclusion of post-processor for new 1.92 summation Revised matrix storage and solution 1978 - - - - - - - -- Ar thurl)l .ittleine
e Modal Analysis by ADLPIPE During the Period 1968 Through 1976 In this period of time, ADLPIPE was licensed to several clients and released beginning in 1972 to several nationwide computer service bureaus. A listing of ADLPIPE versions and documentation is given in Appendix I. The names of ADLPIPE licensees and the ef fective dates of the license agreements are given in Appendix II. The development of the seismic analysis method was guided by available Ilterature and the design requirements of our clients. A method of analysts was developed which was explained by two documents published in 1969. These are enclosed as Appendix III, " Modification One-- Response to Ground Shock Spectra" and Appendix IV, " Development of Modal Participation Matrix for General Three-Dimension Shock Input to Lumped Dynamic System". In Appendix III on page III-5, I state "the modal amplitude, q ,.is thus evaluated as a scalar summation of the products of the nt0 vector of the modal participation matg,Lx and the spectra amplitude (Dg )n". The " spectra amplitude" means the spectra displacement components in the principal coordinates of the piping S system. From these modal amplitudes, a set of displacements for each mode of response is computed. At each point in the piping system, three modal moment components are then computed, one of each principal axis. Each component was then squared and then the square root of the sum of squares was taken to combine the effect of all modes. This - concept was used to model earth motion along a vector which was not necessarily aligned with a principal axis but was skew and was decom-posed to three components. The reason for this development is shown in Figure 1 (page 4) where a structure is not aligned with a global coordinate system. An earthquake is assumed to act perpendicular to one wall of the structure. Mathematically, the skew axis of the earthquake is decomposed into two horizontal components in the global axes. fp; A user could calculate earthquake response with n vartical component and a single horizontal component if the two axes were decoupled com-
- $ df bining several such analyses to create a worst case effect. A user _
i could make three or more different analyses, one for each principal axis again combining the results. - 1 [ Users who made a single analysis usingit;ri-_ directional earthouake I \A)bgC A *~would have_ printed out a__sinile set of_ modal moments. If one isolated each response spectra component by a separate analysis and computed three
~
g@ > sets of individual moment components, the resultant from_the single tri_- directional analysis would be the algebraic sum of each individual component for each earthquake directional component. The upper level would be the j absolute sum of the intra-modal components. The lower level could be
~
_z_ero within a mode. However, it is my view that the inter-modal summation i using the square root ^* eha squares would not be zero and, in fact, wouldnotvarygreatly@sn33percen@fromasquarerootsumofsquares (SRSS) intra-modal summation. a numerical example is given in Appendix l} vy Mf'M
. Arthurl)littleinc
r^ FICllRE 1 UNIDIRECTIONAL EARTHQUAKE WITH SKEW COMPONENTS 1N y
-- E(+x) t x M 9, 4- + - -
\ .
p l Ta m k nen T I v b -
!3 (44)
Ar thur Diattleinc
- - . +-...--=~-e .mm,. . o.m-- - - . - - - + - .*
y___
-- V, " Dynamic Analysis by ADLPIPE" which I distributed in September 1974.
l Prior to 1971 any combination of loads or earthquakes had to be made { by hand or by another progrem. In 1972 I released a summation proce-dure which enabled users to combine loads in accordance with B31.7 and Section III criteria. In 1973 the computation of fatigue usage factors was released, which included the cyclic effects due to various earth-quake components. If these summation techniques were used, the user could input several transient (earthquake) loadings and combine these loadings, one by one, with a sustained loading (deadweight) to achieve a " worst case" stress calculation. Modal Analysis by ADLPIPE During the Period 1977 to the Present A new option was made available in ADLPIPE in February 1977 for the computation of earthquake response in accordance with Regulatory Guide 1.92, Revision 1 March 1976. In addition, a post-processor has been developed which enables the user to make a number of combinations of directional earthquakes effects not included in Regulatory Guide 1.92. Verification of ADLPIPE Verification of ADLPIPE was undertaken in a series of fundamental checks. In important modifications a supporting document was prepared as an ADLPIPE reference. The verification procedure was as follows. The thermal and deadweight loadings were checked by a Hovgaard Bend and hand calculated systems given in " Design of Piping Systems", M. W. Kellogg, Second Edition,1956, and " Formulas of Stress and Strain", R.J. Roark, McGraw-Hill. The dynamic analyses were checked by " Response of Structural Systems to Ground Shock", Shock and Structural Response, ASME, 1960, in "ADLPIPE Results of Model Given by Young (ADLPIPE Reference 4), and " Dynamic Behavior of a Foundation-Like Structure", Mechanical Independence Methods, ASME,1958, in " Experimental Verification of ADLPIPE Mod 1" (ADLPIPE Reference 3). The time history analysis was checked by a separate analytical solution of the problem given in " Analytical Methods of Vibrations," page 395, Leonard Meinovitch, "ADLPIPE Time History Response Compared with a Known So~ution for a Heavily Damped System (ADLPIPE Reference 14). A second check'was made using " Pressure Vessel and Piping 1972 Computer Progress Verification", ASME, 1972 (Problem 5). The thermal transient analysis was verified by a separate analysis, "Tran-sient Thermal Gradient Stresses", E. B. Branch, Heating, Piping and Air Artiiur I)1.ittleinc
Conditioning, Volume 43, 1971, pages 132-136, "ADLPIPE Thermal Transient Analysis" (Reference 15). The computation of intra and inter modal moment component summation has been verified by a separate computer program for that purpose. A report "ADLPIPE Modal Response Combination for Closely Spaced Modes", is available as ADLPIPE reference 24. Various calcu]ntion procedures required by ASME Section III were verified in ADLPIPE references 10, 11, and 18 entitled "ADLPIPE Computation of Bending Stress in Tees and Branch Connections, ASME Section III, Class 1 Piping", "ADLPIPE Ccmputation of Resultant Moments for Section III class 2 and 3 Stresses", and "ADLPIPE Stress Computation of Piping Compo-nents: A Comparison with Hand Calculations for ANSI B31 and ASME Section III." In 1978 an independent third party review of ADLPIPE (Section III, Class
- 1) was performed " Verification of ADLPIPE, ASH 2 Section III, Class 1 Piping Stress Program", Teledyne Engineering Services, Report No. TR-2884-1, August 11, 1978.
ADLPIPE Development Policy
- The following policies have been in effect during the development of ADLPIPE: ,
- 1. The details of calculation processes are available to the public by free distribution of operating manuals and. references.
These are tabulated in Appendix I. Each major new feature of ADLPIPE is documented for user review.
- 2. Program listings are made available to licensees. Licenseen are not restricted from making program changes.
- 3. ADLPIPE is periodically improved and updated and licensees are notified of the modifications at the time of the release of the modified version.
- 4. ADLPIPE is hand checked wherever possible. Ilhen this is not possible, ADLPIPE is checked by experimental results or the results of other calculation procedures. Every modification, large or small, is checked.
- 5. Special versions of ADLPIPE will be written to a licensee's specification. However, the version of ADLPIPE released to computer service bureaus generally does not have such special additions.
- 6. Old versions of ADLPIPE are not retained by Arthur D. Little, Inc. Instead, beginning in 1971. all new versions of ADLPIPE were backward integrated. The present version of ADLPIPE
_ r, _ Ar thur 1)1.ittieInc
e maintains all past features which have been made available to the users during the period 1971 to 1979. y . I. W. Dingwell
- Arthur D. Little, Inc.
Cambridge, MA 02140
' April 19, 1979 \ - .
a Arthurl)litik,lix:
APPENDIX I
" ADLPIPE VERSIONS AND DOCUMENTATION Version Documentation and Features April 1968 ADLPIPE Thermal, Static, Dyne.mic Pipe Stress Analysis Operating Manual, undated.
April 1968 ADLPIPE Modification One: Thermal, Static, Dynamic Pipe Stress Analysis: Operating Manual, first version dated March 26, 1969. Features: Thermal, deadweight , external, acceleration and shock loads; singic load st ress analysis; code - B31.1 (1955). August 1970 ADLPIPE.... Static-Thermal-Dynamic Pipe Stress Analysis dated August 15,,1970 New Features: " Code - B31.1 (1967); equations 9-13, B31.7
~
January'1971 ADLPIPE.... Static-Thermal-Dynamic Pipe Stress Analysis dated January 15, 1971 New Features: Four modal summation techniques: maximum, maximum and square root sum of squares of remaining modes, square root sum of squares, absolute; square root sum of squares for stress calculations , July 1971 ADLPIPE.... Static-Thermal-Dynamic Pipe Stress Analysis
~~
Lt September 1971 dated April 1, 1971 November 1971 New Features: Stress summary report, B31.7 for multiple December 1971 loads June 1972 ADLPIPE..... Static, Thermal, Dynamic Pipe Stress Analysis July 1972 Input Preparation dated April 1, 1972 December 1972 New Features: ASME Section III, Class 1 (1971), summary stress report of multiple loads; closely spaced modal summation .. ,
References:
- 1. ADLPIPE Mathematical Analysis and Logical Procedure
- 2. Section III Sample Problem
- 3. Experimental Verfication of ADLPIPE
- 4. ADLPIPE Results of Modal Given by D. Young
- 5. ADLPIPE Modification 1, Response to Ground Spectra
- 6. Development of Modal Participation Matrix for General Three Dimension Shock Input to Lumped Dynamic System September 1973 ADLPIPE..... Static, Thermal. Dynamic Pipe Stress Analysis Input Preparation dated April 1973 New Features: English and Metric units; summary stress report, Sect'on III Class 1 (1971); fatigue analysis (Class 1); graphical output:
isometrics for input checking, dimensioned isometrics, stereo plots of deformed piping Ar thurI)l.ittleinc
^
References:
- 1. ADLPIPE Mathematical Analysis and Logical Procedure
- 2. Section III Sampic Problem
- 3. Experimental Verification of ADLPIPE MOD 1
- 4. ADLPIPE Results of Model Given by D.
Young
- 5. Generalized Piping System Response to Ground Shock Spectra
- 6. A Method of Computing Stress Range and Fatigue Damage in a Nuclear Piping System by W. B. Wright and E. C. Rodabaugh.
May 1974 ADLPIPE..... Static, Thermal Dynamic Pipe Stress Analysis Input Preparation dated May 1974 New Features: Codes - B31.1 (1973): Section III, Class 1,2,3 - New
References:
- 7. Section III Sample Problem Class 2, 3
- 8. ANSI B31.1 (1973) Sample Problem r
April 1975 ADLPIPE... Static and Dynamic Pipe Design and Stress July 1975. Analysis: Input Preparation Manual dated January 1975 New Features: Revised input organization (geometry and
. execution decks)
New
Reference:
- 9. ADLPIPE April 1975 Release April 1976 ADLPIPE... Static and Dynamic Pipe Design and Stress **
t Analysis: Input Preparation Manual dated January 1976 New Features Section III Class 1, 2, 3 (1974); force time history dynamic analysis New
References:
- 3. Documentation of ADLPIPE for Static and Dynamic Loads and Stress Evaluation, September 1973.
- 6. A Hethod of Computing Stress Range and Fatigue Damage in a Nuclear Piping System, W. B. Wright and E. C. Rodabaugh, Nuclear Engineering and Design, 22,(1972).
- 7. Sample Stress Analysis of ASME Section III Nuclear Class 1 and Class 2, 3 Combined Piping System and ANSI B31.1 (1973) Piping System Ccmputed by (DLPIPE.
- 8. ADLPIPE Skew Card Test Run, July 1975.
- 9. ADI, PIPE April 1976 Release.
- 10. Ani,PIPC Computation of Hending, StreNH In TeeH and nranch Connection 9, ASME Sect ion III, Class 1 Piping, July 1975.
- 11. ADLPIPE Computation of Resultant !!oments for Section !!I Class 2 and 1 Stresses July l975.
- 12. ADLPIPE Detection and Reduction of Numer-ical Round-off Error with Springs and Stiff H"mbers, July 1975.
1-2 Ar thurl)httleinc
Royalty Agreements C0!IPANY ADDENDUM EFFECTIVE DATE EXPIRATION DATE o Black & . 10/04/74 perpetual W Veatch h Blaw-Knox 10/04/67 perpetual g Brown & Root 11/07/75 perpetual g Burns & Roe 7/22/77 , perpetual P.O. automatic extension 8/19/77 l' PENDING Comision 7/ /74 perpetual Federal de - 1 2/23/76 Electricidad , Framatone 11/29/72 perpetual 1 11/16/75 2 7/20/76 Gibbs & Hill 2/20/70 ' perpetual (,\ -
~ 7/06/72 perpetual 1 11/01/78 automatic extension.
M. 'J . Ecliegg 5/12/70 perpetual
@ Company Charles T. 11/19/75 11/19/76 g Main, Inc. P.O. 10/26/76 perpetual Montreal 4/18/72 perpetual Engineering 1 6/25/75 Company &
Monenco Comput-ing Service Ltd. Northeast 7 automatic extension h Utilities Company Service = 1/01/77 Power Piping Co. 5/12/70 perpetual Il-3 l Arthurl)littleloc
Royalty Agreements (cont) , C0!!PANY ADDENDUM EFFECTIVE DATE EXPIRATION DATE Sener Ingeniera y 2/01/72 perpetual Sistemas, S.A. I 10/01/73 2 6/16/75 3 5/01/77 4 11/01/78 United 5/20/70 perpetual 1 . Engineers
& Constructors Westinghouse 11/27/67 perpetual Electric
() Corporation fl-4 At tlnir1)l it tie.ltK
APPENDIX LIl V. MODIFICATION 1: RESP 0:a;K '10 GlutlNDJh:E SeLCTRA The basic approach to be used in compnt inc, the response of pipinn systems to ground shock inputs in terms of displacement (or velocity or acceleration) spectra consists of nenerating the dynamic properties of the system and applying a modal superposition method (or nornal mode method) to define the structural renponse to the shock inpnts. The for-mulation in terms of normal modes follows nencrally the forr discussed by Young.U) As fomulated in this reference, the contributions from the individual normal modes are defined in terms of a nodal part icipa- - tion factor which depends upon ,the nomal shape (eigenvector) and the distribution of the load over the structure. 't h in fomulation is appil-cable, however, to sys tems excited by one-dimensional shock only, i.e. , with the inertial elements restricted to motions in a plane. For the general three-dimensional shock input and response case, the contribu-tions of the nor;nal modes can be shown to be defined in a modal partici-pation matrix. .. A description of the steps leading to the detemination of the re-sponso due to ground shock is given in the following paragraphs. A. Calculation of " Reduced" Stif fness Matrix in order to define the nomal modes of the piping systems, a ficxi-bility or stiffness matrix relating forces and deflections at the mass points in the system must be generated. Following the procedure in ADLPIPE, a network stiffness matrix is first f orned as an N by N array for a system of N network points. (Each of these N x N " terms" are 6 x 6 subsets. ) The numbering of the network points is carried out in the following priority: first, the m.nis polnis; 'a c cond . the interior branch points; and Iinally, the anrhoe poine.. lho utilln"e neat lx
- 1. Young, ILina " iter.ponno of St ruct us al Svut emu t o i;roinni Sloirk", Sinni and St ruct ural I(ennonse, American Societ y of Hei h.nilcal Engi nee rs .
N. Y., 1960. rit-i Arthur D UtticInc
thus formed will be ordered an Indicate.1 helow: A 15 C D E F G 11 1 A mass points sub-m.it rix B;.D, E interior points suh-nat rices C , F. G , 11, I anchor points sub-matrices As shown, the matrix is partitioned into the three categories of network points. The formation of the complete matrix is carried out by ADLPIPE. The rows and columns corresponding to anchor points are now deleted .' from the stiffness matrix, Icaving a matrix characterizing mass points and interior branch pqints only. s,, , j A B a e O IO . D E ay F 3 4 0 represents deficctions at the mass points, and T g up usents defice-tions at the interior branch points. Similarly, r greprem nt s to.nts at the mass points, and Fy loads at the interior branch pointe.. In the case . of free vibration, the loads F are in rtial loads due to the mass points 0 and the loads Fg are zero since interior network points are not loaded. The equations then becomo u * ' v 8 A* ?. g + li * , if D+ i g+E. t, u From the second equation, A t * " \0 .nlistituting into the first equation Ll1-2 Arthur Q Uttieinc
F i: . l
' I (A lI 13 ) \ O . I' O
This resnLLs in a " reduced" st i f fness mat rix , relating the forces and
~ deflections at mass points. This matrix la an n x n array where n is the number of mass points. l'or Inert ia loa 19, this m:itrix e<piation may be written as g.n o - ,2 33 9 where g = (A - B I ' D)
B. Calculation of Normal Hodes "Ihe eigenvectors, 40' "" "" ' "'"""'""*' 'n, or earli of the n normal modes are computed by solving the mat rix equation y.40"* "0
, 0 for each of its n characteristic sointions. This equation may he ' -.e \
solved by iterative procedure when put into the form AV =w V n n n This transformation is performed by defining . n . n /2 g /2 l 1 and V =M 1/2 A n 0 n thus defining the matrix A as A.M -l/2 gi //
-It assures that the Iteration will converg" anel have real anel prisi t ive eigenvaines' .
- 2. Wada, 8. St i f fncus Mat rix St ruct u ra l Ana lys is, Jet Propulsion 1.nhorntory, Technical Report No. 12-- 74, De t obe r 11, 19 M .
111-3 l Arthur D Litticinc, L u
With the mat rix equation in thir. lorm, the irrestive prnrenu will converge most readily on the eigenvalue Imvinn the larnest mannitude. For ground shock response appIicalions. It is nore deni rable for t he process to converge most readily to the smallest einenvalue. Conse-quently, the matrix equations are put in the inverted form CV = i V n n'n 1 where C=A and \ n= 1 /... il and application of an iterative method, such as the Stodolm-method, will produce the successive modal frequencies (eigenvalues )' and modal J columns (eigenvectors) of a system in ascending order. An alternative solution technique, which has been utIIIzed in ADt. PIPE MOD 1 is the Jacobi method . In this procedure, all of the eigenvalues and eigenvectors are produced simultaneously with equal accuracy. This m"cchod may, therefore, employ thc matrix equation in 2 In MOD 1, the second cither form (i.e. , with eigenvalues 1/wn or > n J. form, in terms of A , has been used. The modal f requencies are stored in : " frequency vector", and the modal columns are stored in modified The modal columns are modified by form as columns in a " modal matrix". first converting the V back to modal deflections:\ 0 and then by nor-malizing the column. Each of the set within a mndal column, ' in' " " represents a normalized deflection of mass I in mode n. C. Calculation of Equivalent Static Deflections As indicated in the appendix, the modal adplitude q n is shown to be given by the expression y = ): 7 n , n7 C#F)n
- 3. Greenstadt, J. "The Determination of t he Charactert 4 ic Roots of a Matrix by the Jacob i Method", Chapter 7 of : tat h"inal I. a t Mot henfs for Diettal Computers, John Wiley , New York , i W) .
Ill-4 Arthur D Litticinc
p: -
.O
- l i t, tin t, hock in-02,. )
where 79 , is the modal participat fini matrix and This general put displacement for cach coordinate and for each mode. threc-dimensional form reduces to the simpler formulation y = 't n Dn n
,ame in in the rase that the input shock motinn at the b.cse in th.
every coordinate. It'is this latter forp which is eleveloped by 1, Young (1) . For this one-dimensional cane as discusseil in R. ference
'the modal participation factor is def ined for each mode, while for the general three-dimensional case, -the modal participation is defined for each mass for each~ mode, and thus is In a square array form rather than in a linear array form.
The amplitudes (Dg ) arc obtained f rom the given input shock spec-In these spectra, tra (e.g., Housner spectra for carthquake fondings).
- the amplitudes are defined by the modal f requency and by the coordinate '
in axis. For each value of e , therefore, and for each conrdinate axis which there is a prescribed Input spectra, we have a value of (D;)9 Ane moaal amp utud,e q is then evaluated as the scalar summation of the products of the nth vector of the modal participation matrix and the spectra amplitude (D g) , or, as given previously,
'q n =E7,
- n. (D )n g
The modal amplitudes are now converted to amplitudes In the original co-t ordinate system by the relation l ng ~ +; n *I n t o G ronnel Short ", Shnrk
- 1. Young, Dana " Response of St ruct ural Svt.f.w:
and Structural Response, American Society of Mechanical Enr.ineers, N. Y., 1960. Irr-s Arthur D UttieInc.
"this now provides a set of direpliarement.. n;. lor carh of the n modes.
These individual sets of displacements can now he applied to the
'Ih" enrrerspondinn network system as equivalent static deficctions.
It should be forces are obtained by the usual procedures ol ADLP l PI.. noted that the stiffness matrix to be used f or thi:. procedure most he that resulting when the rows and columns corresponding to anchor points are dele ted, i . e. , A 11 U K
~
The reduced stiffness matrix (A-BE D) cannot be used, since interior points- (branch points) must be considered in the process of transferring
~
interior loads and deflections from point to point. ADLPIPE utilizes the network force sets. to generate stresses for each mode. The upper bound for the stress levels at any point in the erated for mystem is given by the absolute summation of the stresses gen each mode. Such a summation assumes that the contributions from each the same mode reach their maximum value at the point in question at time. Other methods of summation m,ay be used, of course, depending on A sunnosted alter-the degree of conservatism desired in the analynis. native, for example, might be the sum of the contribution of the f unda-mental mode and the rms summation of the higher mode. Ii1-6 ArthurD Utticinc
/ APPENDIX IV Olt GENERAL. 1)hVELOPMENT OF MODAL pAftTICipATION MNIR IX DYNAM IC SYS'llM* TilltEE-DIMENSION SHOCK INPUT ~10 LUMpl.1)
'lhe development of the modal participat ion f actor in the analysis of the response of a lumped dynamic system to a one-dimensional shock by applicalion of I.anrange's input is carried out by Young in Reference i equation with the system kinette and potential enernies expressed in In this appendix. thlu <levelopment is ex-terms of normal coordinates.
tended to include the general loading case in which differcut shork in-The terminology puts are allowed in each of the system coordinate axes. the extent possihlo. utilized by Young has been follo*wed to For the lumped system defined by the sym:netrie inertia matrix m be the clastic displacement in and by the stiffness matrix kg , let u g The th coordinate, and let u g +s g he the absolute displaecment. the i in the general case, six elenent vectors for each elements ug and sg are, mass point. . The kinetic energy T of the system is given by 1 T=7 I m (6 + s)) (6, + A() ij 1 i i*"I.i"Ikl+"lj*i j
*Y "Ij i j + *ij The potential energy V is given by 1
ug u V=7 ijI k We introduce the normal coord inat en y (t ) .inil p (t ) by the l inea r transturmaLionn ta g (t) = T +I" q" (t) n sg (t) = T. 41n Pn I" n
- September 30, 1974 Iv-t Arthur D Uttieinc
c - _ . _
- o. , .
n.r r. r.pein--
- (where + g are Lin' insultal ruininn : ni i lo nol.il in.it s i :< ) :ui I s h.
and i , zuni gj and ding transformations between E and [i , ii, and fl 9. [* g g We further define the generalized incrtin by p,. . M g =E m +
+7 1
a eilanonal Ilecause of orthogonality, Lin r,eneralized in r i.e mat rix it, matrix, and hencc may,be written as M kt" k 0 kt where'6 gg = Kronecker delta. Because of the symmetry of the inertin matrix, m g)*, the bilinear form E m ij EE becomes j i gj . E ijkt m)fik k +jt t " kt "k ^kt I k g I."k f k Ik Ik I and the bilinear term I mg ) E k k k ij g 5 also becomes k Em Similarly, the quadratic forms I mg Eg E) and I mg dg $) become L . (9k)2 and I g _(p k I"* EMk k k The kinetic energy may, therefore, be written as l l 2ft y p I ff p l
.l' T = ) Ek (M
*O From the definition of normal modes, we have ) (k g) - e.-{ mgy)i g ~
)
la the corresponding where + is an eigenvector (modal coltimn) aiul .- tv-2 ArthurD pttieInc
1e . 0 . . .
'/~ - cir,envalue.
this becones For the nth modal column and the nth cir,envalue, U
); [k -
(w, ,) m}l ]n J or T. k i = (n.ie )2 3: m il in . Il in J
=-(wn) 4 =u~n !! 4 T. k ij jnt
&is T.
mi. ) ;in is n ns g). 1 ug u Now the potential energy is given by V = 7 I k g IJ 1
"2 'ij in9 n is 9s l; ijns 1
"2 i
(' lj ln js Un 9s , ns 1] 1 T m 6 ns Un 9s
=7.w ns i
1 2 2
- n en 9n
,w -
= 7 ).0 The appropriate form of Lagrange's equation is bdt(3q ) + &q.' =0
.which gives the equations of amtion, il + ../ q v - p, The solution of this equation for the meulal ampiituite q i :.
L 1 qk (E) ~ ~ Uk (T) sin w ( t-T)ilT k O. IV-1
~
Artt.ur QLitticinc.
-- ----.-u__._ _ _ _ _ _ _ _ _ _ _ _ _ _ _
From the transformation equation, the vector F, p in related to the vector This 5 which describes the accelerations of the base of the system. n where ): +~ + , g, the iilent i ty mat rix. relation in pk"} + t 1 -l ~}
- llence, qk (t) " ~kll kt t(>"'" k 1
If [R N L (t) sin wk (t-T)ilT and the shock input (t)]k " ~k g displacement (Dg )k = j[R (t)] l , , the input spectra are defined as the maximum modulus of the response value gR IL). The modal amp!itudes
-1 where t is an element now become qk # +kt (Dg ) , or qk #I#Mr, (D ) ke t
b of the modal participation matrix for mode k, mass point E.
.In the special case in which all of the elements o fU are the same and all of the (Dg)k are the same, this expression reduces to l9 k max k kt t
This is an alternative expression for Young's result l9k max " k 'k is equal since it can be shown that the modal participation f actor, y, , to E Y g. The modal participation factor is appropriate, however, only t for the special case when n11 of the elements of the base acceleratinn This in not Ihe case, of courtee , in theev-vectors Ug are the same. dimensional shock motions with different shock inputs (spectra) in the various coordinate axes. TV-4 Arthur DIJttieInc
Al*l'F.Nie l X V DYNAMIC STRESS ANAL.YSIS IlY .\lll.Pil'I' by I. W. Dingwell Arthur D. Little, Inc. On page 5 of the reference an expression for a set of displacements is developed for each mass. degree of freedom and for each mode: i "#nn i 9 These displacements are developed from the normalized set of displace-ments q, as transformed by the modal matrix t in f r mass direction i and mode n. The displacements, Xg , represent the zero to peak displacement of each mass degree of freedom when subjected to a shock loading which is described by a (displacement / velocity / acceleration vs. frequency) response spectra. The displacements have a consistent set of algebraic signs Reversing
- which define the mode shape of the deflected piping system.
the signs of the' displacements gives the opposite peak modal deflec-tions of the piping system. From this set of modal displacements, Xf , the displacements of the non-mass points are calculated. There are two types of non-mass points: a) non-mass network points, and b) interior points within a pipe section. Since ADLPIPE uses a transfer matris: technique for combining several pipe elements in series to formulate the stif fness of the section (a section is a series of connected elements), the non-mass network points are calculated first, then those deflections are utilized to calculate Finally, internal forces, moments, reactions at the network points. and deflections are calculated by transferring the initial boundary conditions across each member in a section. Thus, for each mode, a set of moments is calculated:
't
'kjn
- to Ground Shock Spectra by Generalized Piping System Response Inc. , Cambridge , Massachusetts .
Irving W. Dingwell, Arthur D. 1.ittic. V-l Artinir D1 ittielit u
k = orthogonal axis (X, Y, Z coordinate) where J = earthquake direction (X, Y, Z axis response spectra) n = mode With a normal mode analysis, all coupling and phase relationships { Ilowever, since these moments have algebraid between modes are unknown. the signs and refer to a consistent position on the piping surface, question of how to sum the modal mnmonr= arises. d The present version of ADI. PIPE assumes that earth motion is orientA along a single vector and is composed by a spectra with components in_ Therefore. In a sinzie endo. the nipinc the three orthogonal axes. responds "in phase" IN W 6OW , pt(y&6 HUG 3 ( 9"" " } W## M nk " j=1 "kj n from the
- and the algebraic sum is taken of the motion which results The response is independent of axis orientation.
single earthquake. a mean summa-Cince there is no phase relationship between modes, The present version uses the square root sum of tion must be taken. squares. 9N g(L55
" max 3 12 (Equation 2) g=(E (I )2) f n=1 j=1 es that closely spaced There is an alternative technique which impid Therefere, when that
- modes are coupled and are taken to be in phase.
occurs, the square root sum of s ,uares is taken of the absolute sum of the closely spaced modal m ments. For instance, modes 1 and 2 are closely spaced 3 3 1/. M 3
'kj l
- j
+
j=1 bjM k " ((j=1 j=1 v-2 Arthur t)littleinc p
s . 'r The test for the closely spaced modes is: (f 2 ~f if 1 < k, then the bandwidth factor (k) for these modes f 1 (This type summation will cause the program to form an absolute sum. must be requested of the program by t. h e analyst. At present, the factor - - l i If 22 = 0., K in percent is entered in the Z2 field on the S!!OCK card. then equation 2 is utilized.) ALTERNATIVE SOLUTIONS A conservative assumption is that the ' vibratory energy in an ' earthquake is random and the component moments along each axis are ' independent of one another. Realistically, the carthquake acts 4s
}
three different earthquakes, with the axis orientation a variabig. e I Therefore, since phase relationships are unknown, a mean solution is In mode n, taken independently for each shock direction. 3 . (Equation 4) _t M =(E (M )2)1/2
- nk j.1 kjn Following the square root summation for the modes a to n,x
" max 3 (Equation 5) l Mk ~(t rlHkjml nul j=1 Since the absolute sum is overly conservative, an alternative is to ta*4e the maximum modal response plus the square root sum of the squar_e IW- .
of the remaining moments. 4 ) g @Q
" max-1 3 i 3 (Equation 6)
Mk ~(Z j=1
)m +Cr n=1 (I
j=1 h))f2 Each of these alternative solution summation schemes or variations there-on can be inserted, upon request, into the ADLPlPE program. The resulting stress analysis is dependent on the summation of the modal moments. The example given here is not a statistical mean but certainly indicates that the present version of ADLPIPE is unconservative As a consensus is reached, other summation techniques but realistic. will be introduced. V-3 Arth,ur Dlittielir
. . . _ . _ _ . _ . m--
.e r ica l twaP le ' ') .
At point zero in ASME Section III Sample Problem (Class 1. Class 2) , Mz E x N - e Shock mode n = 1 67221 Dir. x -47467 -1297 y -153624 -4199 217557 z -27185 -743 38498_ (Equation 1) -228276 -6239 323276 algebraic sum (Equation 4) 163071 4457 230936 SRSS Shack mode n =2 -5128 2882 Dir. x -27343 y 159117 29843 -16774 2 -851446 -159690 89758 (Equation 1) -719672 -134975 75866 algebraic sum (Equation 4) 866617 162535 91357 . SRSS 9
# Shock mode n = 3 3195 1426756 Dir. x 101890 y -29914 -938 -418883 z -8862 -278 -124098 (Equation 1) 63114 1979 883725 Algebraic sum (Equation 4) 106559 3341 1492144 SRSS TOTAL moment computed by (fix2+my2+m 2)1/2 RATIO 7 (Equation 2) 757641 135133 944098 1218031 1.0 SRSS of alg. sum (Equation 4) 828241 162630 1512670 1761701 1.44 SRSS of SRSS 143193 1282867 1639664 1.34 (Equation 1) 1011062 i kC Absolute sum of alg. sum 1814437 2147615 1.26 (Equation 3) 1136247 170333 C Absolute sum of SRSS
- 141520 1215783 1553406 1.27 I__ Max. + SRSS of alg. sum (Equation 1) 956512 ,
2045531 1.67 b (Equation 6) 1061416 168105 1740499 Max. + SRSS of SRSS
')
)
O CONTACTS FOR SEISMIC FIVE HOME PHONE NO. 0FFICE PHONE N0_. NAME_ FACILITY BEAVER VALLEY 412-343-1728 Gil Moore 412-456-6523 Vice President 412-266-7631 Jack Carey 412-643-8800 (338) Licensing Contact 41'2-456-6910 412-456-6593 (Verification) Telecopy (Office) MAINE YANKEE _ 617-263-5411 Wendall Johnson 617-266-5805 x2803/2804 Vice President 617-366-0135 Bob Groce 617-266-5805 x2868 Licensing Contact ~ Night - 617-366-4533 (Day) 617-266-5805 Auto - No Verify Telecopy x2900 Manual 24 til-frill re- 627-Df 5 PASNY - FitzPatrick f.3.6MLi 314-764-5418 Geo T. 8erry 212-397-6211 Executive Director v s. 3 W 4311 m Heud NYC- 212-397-7950 609-667-7528 Licensing Contact Jay Iyer ) Phila- 215-422-3335 212-397-7618 (Auto)
# 212-397-6242 (Auto) ! Telecopy ,, \
l ..'
)
3;'l - cu.'Est'd. G .- jr Fo'l-171-381*f yoi_y Q -7 M SWRY g,w 3 ,u,v (yg 8(g _gy ,,3g o 804-771-3264 804-81232388 Joe Stallings Vice President 804-798-3639 Dave Speidell 804-771-3916 Licensing Contact 804-771-3348 Telecopy i STONE & WEBSTER Bill Kennedy 617-973-5276
.~ fe ,:v-a j
1 Don King 617-973-2034 l' ?t1* A. b w tosso 98(o Illi i
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LM & k pn VIRGINIA ELECTRIC AND l'OWER COMPANY ' Rxcumoun,VtmorazA 20261 cu a oon June 27, 1979
/ .h Fr. Joseph M. Hendrie, Chairman United States Nuclear Regulatory Commission Washington, D. C. 20555 )T/
Dear Chairman Hendrie:
Since the Order to Show Cause was issued on March 13, 1979, Vepco has endeavored to reanalyze the piping systems for Surry Power Station and to keep the NRC staff informed of our progress. To this end, there have been numerous meetings, approximately a dozen letters, and almost daily, phone contact between the NRC staff and Vepco. - The NRC, Stone & Webster and Vepco now recognize the overall analysis effort is of a greater magnitude than was thought in March. The massive magnitude and complexity of the reanalysis was not recognized early, and at times trends seemed apparent which have led us to incorrect assumptions about certain aspects of the overall program. For instance, several thousands of engineering manhours were expended and about 40 percent of the pipe stress analyses were completed before the need for modifications were finally identified. In May and early June, it seemed that this experience could be extrapolated to future analyses and it could be concluded that very few hardware modifications would be required for the remaining systems to be analynd. it turnmi out that this conclusion was not valid since, in the last few days, we have identified pipes which require modifications due to overstress. Early observations were associated with completed pipe stress analyses that may not have been representative of the remaining cases. Recent experience indicates that remaining cases may be susceptible to additional hardware modifications. Several problems that require hardware modifications have been identified, although for the most part they are for reasons other than that outlined in the March 13 Order. In each case, the Staff was notified prior to the modifications being completely evaluated and through the entire analysis procedure. We now believe, as does the NRC Staff, that a large percentage of the analyses must be completed before start-up can be allowed so that sufficient modifications can be completed to assure the safe operation of Surry. The continued shutdown of Surry Power Station remains a frustrating condition for Vepco and our customers. We now recognize the need for many thousands of engineering manhours yet to be expended which probably will not be completed in time to allow the operation of Surry Unit 1 during the summer peak load period. -The Company has made arrangements for 800mw of capacity from an adjacent utility in July and August to replace the Surry capacity. However, the Company's resource level is still below our normal level. We will not, however, in any way compromise the thoroughness of the reanalysis. The safe operation of Surry and effective and open comunication with the staff are equally as important as the generation of power. Therefore, as modifications have been identified, the Staff has been notified promptly, as as the case in our letters of June 8, June 12, June 19, and June 25, 1979. T( f g{ ? 0 {0 f
vi=mu nECTDC AND Ports CCMPANY TO Mr. Joseph M. Hendrie, Chairman 2 The Staff's availability and responsiveness have been extremely beneficial to our reanalysis effort. We appreciate their continued accessibility, and we will continue to provide information in a timely and professional manner so that both Vepco and the Staff may be completely satisfied that Surry can be returned to service and operated safely. Input from the Staff has been and continues to be valuable. We are desirous of providing whatever information the Commission may require to complete the review of the Surry pipe stress reanalysis effort. Very ruly you ,
, ~t ,
Stan ey R one Pr nt
~
cc: Mr. Harold R. Denton Governor John Dalton' Attorney General J. Marshall Coleman Virginia State Corporation Commission North Carolina Public Utilities Commission North Carolina Public Utilities Staff m h v - 4
. _ _ - . - ._ ~
. . LICENSEE EVENT REPOWY j CONTROL BLOCK: l 1
l l l l l lh 6 (PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION) i ]A.lP l A l " l V l S l 1 l@l 0 l 0 l- l 0 l 0 l 0 l 0 l 0 l- l0 l 0 l@l4 25 26 LICENSEl1 l1TYPEl1 JO l1 l@l l S1 CAT 68 [g 9 LICENSEE CCpE 14 15 . LICENSE NUMSER
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_S 60 61 DOCKET NUM8ER 68 EVENT DESCRIPTION AND PROB ABLE CONSEQUENCES h l ]l During a review of SI piping stress calculations an error in the original piping l The line supports were modified. In l l g ; stress analysis for one SI line was discovered. g ; March,1979, further investigation of the A-E piping stress computer program revealed l g g ; th2 program provided pipe s. tress calculations unacceptable to the NRC for certain The station was shut down-on March 9, 1979 for resolution of the g p[ caismic events. g , g ; pipe stress reanalysis. l ( l ]l69 7 80 ) SYsTE M CAusE CAUsE COMP. VALVE , Co0E CODE $U8COOF COMPONENT CoOE SubCoDE suuCODE 1 @ l s l F lh (B_jh l A lh .l S l U l P l0 IR lT lh l D lh l z l h 18 19 20 t 9 10 11 12 IJ OCCUR R E NCE stEPORT R E VISIC*4 SEQUENTIAL R EPonT No. CCOE TYPE N o. _ EVENT YE A R h ,"LER/Ro I 7 I 81 [---j l0l5!3i I/l 10 11 l lTl l-l l1l 5 ,_21 22 23 28 26 27 24 23 30 31 32 HoJR$ $8 iT SCf 8. SU PL E MANU.*A T FR
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33 34 Izl@ 35 Iz 36 l@ l0101010140 33 lY l@ 41 lY 42 l@ l Alh 43 ld4X l 9 l 9 l 947 l@ - CAUSE CESCRIPTION AND CORRECTIVE ACTIONS I Z I Tho incident resulted from the use of an unacceptable comouter crocram for i i I All vicine systems which urt be ! ca'-"'--da- *a w a"a= of =eismic loadine. ] 1 Anv i ]l saismically an'alyzed will be coecleted using an acceptable cocouter orocram. I ]l modifications to supports which are deter =ined necessary as a result of these I gl nnalyses will be installed during the fall refueling outage. so
- 8 YT'SY seowEn oTwEn starus b o$sEoORv' oiscovtav etsCairtioN @
l D lhl Architect-Erigineer Review l 2 lGl h l 0l 0l Ol @ l" N/A l 8 - AEnvlTv CCSTENT mrt. A:Ec cc acuAsE LOCATION OP PELEAsE @ wouNT oP ACTiviry h ll N/A l l .ij* W '
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_ ': !Cio!@l a o a WA so l L:~s C,P on DAMACE To f acitiTY T**E CE!CpiaTiON i ! T , l:'31 ,o N/A .: r';0 Lici?v NRC USE CNLY ,
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Att chment to LER 78-53/01T-1 l . Beavsr Vclicy Pow 2r Stationa l Duqu2cn2 Light Comp:ny l Dockst No. 50-334 During a review of,SI piping stress calculations an error in the original ' piping stress analysis for n-e SI line was discovered. One support on the SI line was modified ano one snubber was added. The Duquesne Light 1 Engineering Department continued a review of the architect-engineer findings. In March 1979, this continued review of the A-E pipe stress computer program revealed that the program provided pipe stress calculations unacceptable to the NRC for certain seismic events. The station was shut down on March 9 for resolution of the pipe stress reanalysis. All piping systems, which the present regulation require to be seismically analyzed, will be reanalyzed dsing Nu Pipe-SSI calculational techniques as described in Section 7 of the June 15, 1979 Report. s These NU Pipe-SSI computer analysis will than become the calculations of record for all piping systems which require computer analysis. All supports for the computer analyzed problems will be evaluated or analyzed to 1'nclude the requirements of IE Bulletin 79-02 relating to baseplace flexibility and factors of safety for the concrete anchors. Any modifications to the supports which are determined to be necessary as a result of these evaluations and analysis will be installed during the fall refueling outage. Any modifications determined to be necessary to limit equipment nozzle and containment penetration loads to within Code allowable values will be installed during this same fall refueling shutdown. . All hydrculic inubbers for which the seismic loading exceeds the allowable load but is included within the one event capability of the snubber will be identified. These snubbers will either be replaced or included in the appropriate technical specification as snubbers which must be tested for operability, subsequent to a seismic event of a magnitude which results in forces greater than the allowable snubber load prior to continuing operation or returning the unit to operation. All of the above mentioned activities will be completed prior to returning the unit to service after the fall refueling outage. e YWzo -
-} / #"% umTso cTArms NUCLEAR REGULATORY COMMISSION h Ma r-#
i
-{t..hIL(j a5> - .
WASmNGTON, D. C. 20555 A.:R 171979
%l.....% /e
- MEMORANDUM FOR: John Davis, Acting Director Office of Inspection and Enforcement .
William T. Russell FROM: Technical Assistant for Systems and Projects Division of Operating Reactors
SUBJECT:
10CFR21 NOTIFICATION BY ARTHUR D. LITTLE, INC. CONCERNING SEISMIC STRESS ANALYSIS OF SAFETY RELATED PIPING Representatives of A. D. Little, Inc met with members of the NRC Staff on April 16, 1979 in Bethesda, Maryland. The meeting was held at A. D. Little's request. A list of attendees is attached. ADLPIPE is a computer code developed by A. D. Little, Inc. to analyze earthquake loading of piping systems. At the time the code was developed, a single horizontal direction earthquake was analyzed.
' This single direction earthquake was divided into North-South and East-West components for ease of solution. The N-S and E-W earthquake components are mathematically dependant and the internal mathematics of the code results in algebraic summation of the co-linear responses.
This procedure is correct for a single direction earthquake or for simultaneous multiple direction earthquake inputs which are dependant. Howev'er, if multiple independant earthquakes are input simultaneously into the code (i.e., one independant earthquake input on the North-South Axiz and another on the East-West Axis), the algebraic summation of the co-linear responses results in non-conservative load predictions. _ This problem is identical to that described in IE Information Notice No. 79-06 and IE Bulletin No. 79-07. Current versions of ADLPIPE offer several options to the user. One current option is a subroutine which complies with Reg. Guide 1.92 for combining model responses. Other options result in algebraic summation, absolute value summation, square root of the sum of squares and others. A. D. Little, Inc. is not aware of actual methods or options used in application of ADLPIPE for design of piping systems for nuclear power plants. They did identify that the following firms have used g(jk
.h +^.% e , .y
. . . . . .e.ge.*
g' g A 8 9 Q--
*O 8% (D ' 48 * ' *i
.}-
2-ADLPIPE on Nuclear facilities: Westinghouse United Engineers and Constructors Teledyne r. Gibbs and Hill Burns and Roe Brown and Root -
) Black and Veatch A. D. Little, Inc. will provide written 10CFR21 Notification on Thursday j April 19,1979. They have also agreed to notify their customers of g the potential for incorrect apolication of ADLPIPE, such that their r customers can evaluate the actual application of AOLPIPE on their plants.
[ This.information will be provided on an expedited basis to ADLPIPE
! users to allow licensees to respond to IE Bulletin No. 79-07.
t A single copy of the current ADLPIPE " Users Manual" was provided for staff review. I - h' T.A:-"M William T. Russell Technical Assistant for - Systems and Projects Division of Operating Reactors
' Enclesure; As Stated List of cc's Attached l .
t 6
~
l
....; _ _ . _ ,l l
e NAME ORGANIZATION . Wally Wright , Arthur D. Little, Inc. l I. W. Dingwell Peter D. Lederer Baker & McKenzie,t!YC R. Lowenstein Lowenstein Newman, etc. Vince Noonan EB/ DOR Marv. Hartzman DSS /MEB
^
Keith Wichman EB/ DOR [ Steve Hosford EB/ DOR R. G. LaGrange EB/ DOR Arnold Lee EB/ DOR J. R. Fair EB/ DOR K. Heiring EB/ DOR L. Brenner ELD W. Russell S&P/ DOR
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B. D. Liaw EB/ DOR D. G. Eisenhut D0R t ** a e e
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cc's: N. Mosley, IE H. Thornburg, IE D. Thompson, IE W. Reinmouth, IE J. Glenn, IE U. Potapovs. IE, Region IV J. Scinto J. Murry ELD (3) D. Crutchfield Docket Files NRC PDR Local PDR C; Kammerer J. Fouchard R. Fraley, ACRS (16) H. R. Denton - " E. G. Case V. Stello , r R. Mattson R. Boyd R. DeYoung . D. Eisenhut
" R. Vollmer
. B. Grimes
~
R. Denise J. P. Knight A. Schwencer D. Ziemann ~ T. Ippolito R. Raid P. Check G. Lainas D. Davis V. I:conan F. Schauer R. Bosnak . L. Heller K. Wichman D. Brin < man Project Managers: (C. Nelsen. J. Neighbors, P. Polk. D. Wigginton) M. W. Paranich, IE 050.(3) S. Showe, IE E. Jordan, ID
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i 22 - r - l ) i s - l were effected by this, members of the NRC staff, on i ) March 10, lith and 12th accompanied the licensee's . 1 i , 1
- % representative to the offices of Stone and Webster to, l
-- 2 - e their architect / engineer to review the detailed designs. l 3 and computations. Concurrently, on March 9th the licensee s'uspended power operation.at the facility. COMMISSIONER AHEARNE: Why? 6 MR. DENTON: Let me ask Vic what his impressions. ,
-7 are? -
8 y ,
. MR. STELLO: He stated to us that there were a 9
variety of reasons he was doing it, one of them related to
- a convenient time he didn' t need t.;e power, it was over 11 the weekend and he was going to shut down and do some .
12 , i maintenance. 3:I 13 It is our belief that he, as we did, after i 14 their committees met, decided that that was the prudent . 15 - thing to do until he had been able to resolve this issue, 16 which remember, I said, the first time he was presented 17 with it was March 8th. So after we told him they needed .to . 18 ; go back and meet with their ccmmittees and make a decision, i 19 and be able to speak for the licensee the difficulty of 20 how much of this problem was there, but it is my judgment 21 that it was because of the problem he was faced with on 22 , March 8th . _,,,
~~
MR. CASE: Although that is not what he has told us, 24 j or told Stone and Webster. ~~ 25 j[ ti N 1 9 4
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/
UNITED STATES 8 ,% NUCLEAR REGULATORY COMMISSION O E WASHINGTON, D. C. 20555 G, h.... + 8 AUG 181980 DocketNo.5b-334 MEMORANDUM FOR: Ebe C. McCabe, Chief, RPS 2, RO&NS Branch, Region I THRU: Samuel E. Bryan,. Assistant. Director for Field Coordination, Division of Reactor Operations Inspection FROM: John I. Riesland, SROIS, FC, DROI
SUBJECT:
BVPS UNIT 1 - PIPESTRESS SEISMIC ANALYSIS (AITS F01004822) Yourmemorandum,datedJune19,198b,providedinformationrelating.toan audit of Stone and Webster Engineering Corporation conducted by Duquesne Light Company on. December 18 and 19, 1978. Because seismic analysis issues resulted in Show.Cause Orders of March 27, 1979, the above audit report may not have been made.available on a timely basis to the NRC. YourequestedthattheinformationbeeYaluatedastotheappropriatenessof additional NRC action in this matter. Wehahethismatterunderreviewandevaluation,butfindthatadditional.infor-mation is needed before a determination for additional action.can be made. A list of items and questions that need to be resolved was prepared by R. Hoefling, ELD, and is enclosed. - We request.the participation of.those on the distribution list at.a meeting on mTuesday. Auaust 2L 19An at 10:00 A A in P-440 of the Phillips Building, to c)scuss the concerns listed in the enclosure. Please call if you have questions regarding the forthcoming meeting.
& h L e$
John I. Riesland SROIS, FC, DROI
Enclosure:
BVPS UNIT 1 - ADDITIONAL INFORMATION cc:w/ enclosure See Page 2 CONTACT: J. I. Riesland 49-28019 l 1 l
-Q;f,s ,A- cS'fI,,YW l
E. L. McCabe - 2- AUG 181980 cc:w/ enclosure J. Lieberman, ELD R. Hoefling, ELD K. Cyr,. ELD 5: llr:ni' a t e - A n D. Wigginton, NRR . U. Potapovs, IE:IV D. Beckman, IE: Resident Inspector G. Gower, IE O e 8
ENCLOSURE , BVi3 UNIT 1 - ADDITIONAL INFORMATION REGARDING PIPESTRESS SEISMIC ANALYSIS REVIEW
- 1. The. names of the NRR staff members who met on May 20, 1980 with members of the legal staffs of the PPUC and Consumer's Advocate. i
- 2. S&W'sresponsetotheDPLletterdatedJ$nuary 22, 1979. '
Response
was requested by March 1, 1979. ,
- 3. The names of the NRC staff to whom this information was.made .
available during the weekend of March 9, 1979 and the circumstances; by which it was made available. t
.i 4. The basis for the claim that the report " appears to contain.infor-
' # mation which was not.available..to..the1RC staff during a review of this matter."
- 5. A copy of LER 78-53/01P.
- 6. AcopyofthesupplementElreporttothatLERd$tedDecember6,1979.-
- 7. Given that the memo concedes that the discrepancies were subsequent 1y' ,
brought to light during the NRC. staff review of the matter, what is the concern? ,
- 8. Giv'en that the report was transmitted to Stone.and Webster on -
January 22, 1979, what is the basis for the claim that the infor-mation in the report was available "long before" .it was made. available to the NRC staff? When did the NRC staff first have access to the report?
- 9. What is the basis for the claim.that the information included in the report does not appear to have been made readily available to the NRC staff at the time of its review?
10.'WasthisreportwithheldfromtheNRC$tanytime? .
- 11. Copies of the " completed in estigations" referred to in the memorandum.
In addition to the above, responses.to.several additional questions would aid in determining the. presence of a Part 21 violation. Any Part 21. violation would appear to involve Section.21.21(a) which requires an entity subject to Part 21 to adopt procedures to (1) either evaluate deviations or inform licensees.or i purchasers of deviations.in order that they may cause a deviation to be evaluated l and (2) assure that a Director or responsible officer is informed.if the construc-i tion or operatilon ofTfacility contains a defect. Thus the relevant questions ! are: - i i l
~
- 12. Wasa"dehiation"present?
- 13. Werethereproceduresineffect13thforthelicenseeandforStone and Webster for evaluating deviatici. and were these procedures acceptable?
- 14. Wasan"ehalu$ tion"conduc'ted?
- 15. Was a'" defect" present? ,
e
- 16. WEsthereaproceduretoassurethataDirectororresponsibleofficer
. was informed of the presence of a defect?
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/ygf/ JLCrooks:jaf ch y 9/10/80 Conrnent 3: Not incorporated. Though the staff expended considerable ~
(more than normal) efforts to correct deficiencies in the piping designs at the five plants shutdown by the Show Cause Order, staff did not independently verify all modifications and thus cannot provide a date for completion of verification of all modifications. Staff licensing and inspection actions have been based on (1) the licensees' commitments and statements of completion plus (2) independent verifications, which though more than normal, were still done on a sampling basis. In addition, the verifications usually did not distinguish between the modifications made for the Show Cause Orders and those made for the related IE Bulletins (79-02 and 79-14). The essence of the
&utn,Le wf verifications was to assure the necessary work was properly done rather than for a full accounting. Some commitments are still open as the scheduled completion time has not yet arrived. The licensing and inspection staffs have been, and are, working closely together on these activities.
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- UNITED STATES
[ *,$ NUCLEAR REGULATORY COMMISSION i WASHINGTON, D. C. 20555 l U{
% n O T ,# l
.....* ADG 2 81980 Docket No. 50-334 MEMORANDUM FOR: Ebe C. McCabe, Chief, RPS 2, RO&NS Branch, Region I A .
THRU: Samuel E. Bryan,. Assistant Director for Field Coordination. Division of Reactor Operations Inspection FROM: John I. Riesland, SROIS, FC, DROI
SUBJECT:
BVPSUNIT1-SEISMICPIPESTRESSANALYSIS(AITSF0lb04822) As a result of.a meeting held on August 26, 1980 between participants identified in the distribution lists, it was determined that additional. enforcement. action on the above subject against Duquesne Light Company.(DLC) would not be justified. This is in response to your memorandum to S. E. Bryan, dated June 19, 1980. The bases for this decision are:
- 1. DLC reported by LER the possibility of a pipestress problem at BVPS Unit 1 in accordance with Technical Specification requirements.
- 2. The 2-day audit report was in. a preliminary stage of v'erification in March, 1979 in that quest. ions regarding the report and the report were submitted to Stone and Webster Engineering Corporation (S&W) by DLC but a response had not been received by DLC.
- 3. Subsequent pipestress evaluation,by S&W, proved that no significant hazard existed.
If there are other concerns regarding significant information reporfing by DLC, we suggested that these be discussed at a future meeting of Region I with DLC management.
, u$ &
~ John I. Riesland SROIS, FC, DROI cc: F,. E. Bryan, IE G. C. Gower, IE
Participants:
See Page 2 l () 90 x .
ucfore che , Nuclear lit y,ulatory Cn= inion of the United Scaces of A= erica
. ~
In t'ne m eter of : Docket No. j Duquesne Light Corpany 1 50-334 1 obio Edison Company - Penns.ylvanis Power toepany , f . I (9.sw.r VaI{cy I' aver Ststion, Unit No.1} 2 . acquasr von nAwuc i?rn 5 PEIITION FOR LTr,T Id Ih7ERVENE CH EEBALF Of EmmxANIA PUBLIC - UTILITI Cot"MLSSION 7-3, . I . -_ t r
'DW hnssylvania Publio UCLILcy Casa:Lu Lua ("Veciticaer"),
I. t throut,h its adorneys, hereby petitionshor.~.1 cave to intervene and requeses ( , and in support thereof svers a follows:. l !~ iserting in ths above.-captioned mat:ct, .
- 1. W Petitioner is an . independent administrative coE5iS350D l ,
whh the exc2aelve and plenary authority, under the lav= of'the consonwealt.h
\
of Piransylvania, to regulate the raras, **rvice and fiscilitias of public. J= ,: l
% ucI23 ries providinr. public ucl21ey rcevice in Pennsylvania.
- 4 :
do j 66 r,z. c.5..ii101, et scq. .. . . . . . . . . . . . . - - . a, . .- j( l 2. The Pect,tioner is charged with the enforcement uf the b' Code of Pennsylvanis, 66 Fa. C.S. If 101( et seq., which L' ruhtic Utdidt? . wndstes esfe, 4dequate and relhble gublic uc111cy service at .)ttse and . Qr :r l lil ca.sronable rates. 1 1 Duquesne Light Comp ny i.. a public utility which provides I;
?
d..-cele utility us:11M in hansylvania subject tu the jurisdicciou of 5 Duqudme Li Ehe Cor.pany (s1.uce, with Ohic Edison f.c.epany
- he t'stitioner. -
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s=3 yennsylvanft Power Cc:psny) is a licencec of the Nuclear P.e;',dcar'; [in
. Cc.-%eion h4bf f ag F*cility Operacifoi.1.iunu No. DPR-66 which' euthortt.es R
;m the. oprsthw of Beaver Valley Powue 'ite.'ve% Unit No,1.
?
- 4. The cost = cf conutructing, r.=intaining, and operating 9,IE Bf tvvr Valley Fenrer Station are recovarad chrough rates fixed by tha
~
Fe.n:::,ylvs=1: Public Utility Co*=nission.
- 5. The Nuclear Reguistory Cot =nizzion has, by ita Orde.r to Shau Cause is' sued March 13, 1979, required that the 2=sver Valley Power gr H
sceu.f.on be placeI5n ci d shutdown tondition until further order of the
? >
y Cr/arrission. .
*L 6. The ordered shubfoEnTdifrTroit in substantial increases
! i;.
.y,. cc.tcicity 'to Duquesno Light Company's customert, including
. .. Jn the ense e .
~
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hundnds of chnusands of reridentist cumeomers and thousands of doe =:seccial La .
' and, Industrial customers, and may result in un.just'and unreasonaMe rates,
}( .
o , . Thu ordered shutdown may jeopardise the coatinued provision of safe, d j Y" adegenea an'd relishle electric service in renas'y1'vania; 7 p g. l L Tfte Pennsylvania Public Utility Commission finds that it
+3 h l '. .. Ir. in c.he public interest to inquire fato the followivy, aspeces of the abovevcaptioned peaceeding:
nL ,= "-" ..
~ _ . . (a). .k' bat is the present risk to the public besith p - .
r :'
,; # and zafety,11 any, in terms of probebilities, due to alleged deficiencies
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; E:
i fl f a che piping syste=s of the Besver 7611ey Fower Station? -
;h ^
E (b) What is, the probable le=gth of L1== tequired to analyse sad cemedy these alleged deficiencins? h o; (c) *kt n addi:1. cur.1 costs vill rMult from thn fp. cr,!ered. shutdevn ni the Besve.r Valley Ter.cce ScsLien? - n.. 5n D
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*spects (a) - 4). 12 it necette.ry to coni,th..t. the ordcred chutdevn of ..
..o t
- g. The Scaver 'ialley Potver SEdt10d id OYde.t to &SSurc that thor 6 *..S no e
Q ucuke risk to the pubUc health and safety? t.
- 5. ,
The bases for Petitioner's avertents are its expericace s u ac,1 upertin virh tr.spcet td pubiir; utf14ty rates, etvice and facilities,
%n -
ek its assessment of the hotential.;icpact of the ordered shutdevn on the bl Et
. putdic he.21th aifety and wlfdre and reqcercic vitality of Wstern t
p_f Pennsylvanta,, its statutory cMigstion to enforce the Phb11c Uc111ty Ea - PD Coac al rennr.ylvr.ula, its nandace to ensure safe, adequate and rclimbie (% , p . h put:1.ic utility service at just and ressca.zble rates, its concern that I]l substdact.aL continuing and irreparabic =.:y.be caused by_t!e P . 1 continued
?. >f 3 F shutdown c1 texver Valley Power Star _ loa, ash its concern that celther .
.a .
k! un/fue riv.k t.o t.he,public hesith and s.sfety nor unduc ecoprain damage - 1
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@ . occur in Western Pennsylvania as a ruult. of tk ordered shutdown of the kl ;I-g . .
w' ' Beaver vs21er Power station. . t1 - eN : VRE7.EERE, the Peunsylvania. Puh11c Utility Cocalusion respect-
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v ; [' fully petitions the Nuclear Regulstory Comalssion en: h ;- g l.1) grant Fetitioner's request for hearing and for leave to
- D :
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+b i (2) schedule inrnadiate hearings heIure the Atumic Ssfety arrd Liccn. sing Appeals Suard. -
- 0) consider the aspects st.sted la Parar,raphs 7(a) M) above s.:.
E$ la du accelerated and crpediefness ranner, f.:>
- G*d (4) set pro."ptly and appropristely to end the ordered shutdown
$.e c
Di D of saaver Vene.y Power Statlun if no uudue risk to the public heakh an'd E b a '
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- s. . .., . . . . . - . . . . .
~ a- , . . . . . ..... . . _'. . cafc y would rendt, er citernativel'y, specify the pradec nat=c cf.
..- c j re;: sin, r4pIgec==nts or ecdifications that ciust ba made hefore Beaver Yt.Lis7 Pov&T St.ntion can he operated, vithcut undus tLak to ths pu'uLLc sgl beal:h sad safaty, ..
p (.5) gesnt euch other ind further relief as may he necessarf, V
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I[ just and Appropriate. ,
- t i e
t; y r P-{ , f.cspect!ully suhr.itted, ti e [ bGt I .
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( G.- eI John A. Levitt wi - Assistant Counsel f.d l i 5teven A. McClaren
)>y Deputy Chief Counsel jj 8 '
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d - P.D. Box 3265 . ~ ( natzi=*ourg. PA 17120 - Gl?) 747 W
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On March 13',1979/ 'the Directbr o.f' tha d_ifice of Nuclear Reactor negulatihn-
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! ' issued an 0rder to Show Csu' e.to Duque'sne' L..i.ght CcEpany, . _ Ohio . ._. Edise.n. ismic:.
Company and Pennsylvania Power Company (Lice.nses)' relating to the.. se design of certain piping systems, at. Ec'aveE Valley P'c w'er"St.itiod[ Unit Hof.' 1. ~~
. "<1. . The order. requir.' es Li.censsa to sho. w'%d.tset.,I'. . . . ' . S; ' '..'. D.. . W ". :.','.' . , , .
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i (1) why I.icens'e a should not' 'resnalyza ,thess' fhcility pipin5 .systens , "*
~ . ' - whichi were analyzed by a computer coda usi=g in algebraid. sur.ation # *
. of the lohds predicted' separately 'for th.a horizontsi anli bert'icar * '
cc=ponents of saissic. events .. such'.'r'eanalysis ,to be performed using f' .. ~ an appropriata piping analjsis.* co=pdt'ar coda' which' dois. not co=bina4' 7 h..
~
leada. algebraically,-' "r. ^~ ;W il '" ~ .. P. " h >>. 'i ' . - t 0., , . " . . . y , . P.t ... a ..< .S y :. y
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i why i. ice.ssee. s.hould not' caka'any E...difiE. stl.i,_ns__ be.to.' "v"., We "n 9 f__ici 4 necessary4. and'. . .;'
. piping sy.. stems- indicated.' by such rsanal.. W..:.'ysis . to_
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. I.. Why the Licensee should not readilyze- the fact 11ty. pleinir systees fer;~ '-"
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- seismic loads on all notentially affected safety systens usint atr -"
"7 . apurooriata niping analysis com:: uter. coda ubich does. not ecebine_ .
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The Licenser; wilr reanalyra the seismic load'.for affected' fac..ility; - V I~~. piping i r *l _ :...__' .rwhich..t .s.y..s -
.'e.s.srus' n$ an.afp- @f..i.rt3 w
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2.E Why the-ticonsee should' not eake ady .codifi'estions to the facilitve - pipint systess" indicated by such resnalysis to be necessary ..- .a -
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. . m,7.. v-L .. N- - Licenses vill naka ^any. m:odifications..to the"iffacted' facility . '- T' :*
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- piping systema.vbich it detirmines' totbe nacessary based:'on'the .
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4 :rr - . - should' not- be suspended gending such. .re.2nalysis .
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. and. complotdon .'of 'ans recuired: eedifichtions.c-" .. * . . ~ f.
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1 . a . Licenses..has.undertskan tha program discussed belov' for',reanarysis ~ "of' -
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J. , . 'any nucoscary modificatiens to tho aff acted piping cystc s res::::d ts.
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" lcssura, ssfe shutdown,copability and. th.a, capability of effected' piping.
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' systems assoilatid'with tha Engineered' Safety ~ Features and' the Emergency - P j ...I
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Cora Cao11ng Systcri,~0a rEhuest that ths facility be persitted'to resu=a" "
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, operation pending completion of resnalysis of the balancs' of"the affectat
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- " fpipinz systems and any necassary .ed,difies.ticosoff - .: ths rc5ainini affscc.ad'. -'I'. -
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- f ., p-Rennalysis of' piping: stress Grich "a code v_hich-dcas not ,use: an algebraic -
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shi:istiosT tschnique) ' aid %alEstioW of' suppdits in: those_ pip.irig. rush.. " I. ? - ~ 7-
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- which were esihina.lly. indlyzed' Ud..i'ng' an alsc.brile: shtdon s !5sI6s.ii-. v .74..
.:.7...ans. which. a..re par.t of systems n.eces.sa.ry tdownito- ._ and.b.the assure .4 sa operability of engiEeered safeEy' featuresfor.'esirgency'coraWoling;i.' arzW
, .i ..
Esing' conducte's one'a 2" shift per day- (10 hrs /shif t). Basis.'
' " ~ * '
f ; '- ..v;: a.c . . . .. . , ... . . . . v.O. . Me. . . ':. c...' . W;. .. .c :' ..,* ,l t d71,. s - ;. . . g As.of March 30,. 1979 _66% bfths pipe sties i ii . f j The remaining runsushe' ld be complated by X.s comput e'r ru s "we e . j runs,. 90% are acceptable .and' 10% ara ~presa'ntlyjundurgoidg,'reanalysi,s '
, As- oimarch. 30;. 26% of the pip,e support dasigs ra riswa wara,de=placedi ," .
l Of: the completed' pipe support: design reviewsi.,63% are acespeablag 2M; y , t d are. presently undayoing,'furthir evaluatica ai. ? 95-ara- undarssii:ia data 112d'- jl 2c analysia ,as applicable Ccmpletion is"schedul'edlof April. 20,. 1979' K
. ~ Related design reviews.of equipment: sasociated' with. thaai pipini sista=s. ' %
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e , Assuming the.same :nta'.of.progress as recently...,~ experien. j .- . _ _ _ 4nd..$-5 , '
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t w;""..Csubmit dc'cumentiation u T.:. .r:. w:.r::+,.-.;.:.;.+~.~. .- -.n. fi 1979f.T.'-4--
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r of sufficient infor=.ation to assess the seismic d'asign espabilities of:- - the affected piping' systems. ife~JilisFa"It is~iHEu=he'nt"dpon. the Nac;;_ ~ ~"'-
- however . to consider and* evaluate:ths engineering a536M ~iid'~da't a~"
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bains developed by Stone alnd Webst'erC. NRC'should* determine on'tha basis.," 5,'
* . . ef .~th. at inf.ormation,wh. ether,. ther.e. is. a ' sufficient threat 35 tha public ~
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<._ health and' safety to warrant the dra.stik step of a further ordar-unti=.u tha~ raanarfils 3dd' possi,ble,1:ddif.i.es.t].cr , .
- ' ~ . - _the shutdown.of.' U61t"1 pendings k6ardia.ss,-oft thai .35forVinc4 ts. thie
~
-of al1~affected piping syit@msI * ~ '
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. . .'a -". . . -that- : o.the. Licensee' . . .Mr"~'_w~: :...- ._. .
..:.the
_ . s to :~. dmitted extent 'g ~ Paragraph II, first sentence is a
. s .- '
! - - filsd a Licensee ~ Event Report, LER 78-53, indiesting'that theresera two " ~
^
piping systens fhr which' stress computist. icds'E h' ce=plete'd' using- I + j . a piping analysis coipater code..* . .. 3, .; ,9.. v,.
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.' - computaF code used,' c'o analyza, earthquake _ load'ing~~for. eiirtainfacility-
-piping systems 'and,a curf antly. acieptabIs' coda 4P' Stress and t he re 'l ated'"'
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.' Paragraph II, fifth 'santenes,'.'the Licensee is~vihhost3dhierant-
- knowledgi'Eo'admitler' deny. ".. "g' -
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.- .. Paragraph IIi sixth santancs; is.,... admitted' to 'tha ftxtain. t ' that '. -
algebraic s'uz=ation withou't time' history analysis c'an, incerthir"" '"' f' - casasp yield. lov'ar esiculate'd'stresass thsi.tho's a esiculatied ' d. ll using.' tachniques: suc. h. as the squara.. rook of tia. e'ud.of. th,.e n
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,,.= Paragraph II,. seventh' and eight senteiices are admitted' i:o thE extsac 1....,_
i . . . . . that current:. industry practicEaeco[snts"for'ths:effacts 'off'eirtfuiua'Ai="2~ f
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