ML19309B681
| ML19309B681 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 03/21/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19309B680 | List: |
| References | |
| NUDOCS 8004070066 | |
| Download: ML19309B681 (7) | |
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SAFETY EVALUATION BY THE OFFICE OF
!;UCLEAR REACTOR REGULATION FACILITY OpEPATING LICENSE NO. OPR-37 VIRGINIA ELECTRIC AND p0WER COM?ANY SURRY POWER STATION, UNIT NO. 2 DOCKET NO. 50-281 March 21, 1980 Introduction On P. arch 13, 1979; the Commission issued an Order to Show Cause to Virginia Electric and Power Company (the licensee) requiring that Surry Power Station,
'ini: 2 (facility) be placed in cold shutdown and the licensee show cause:
(1) Why the licensee should not reanalyze the facility piping systems for seismic loads on all potentially affected safety systems using an appropriate piping analysis computer code which does not ccmbine loads algebraically; (2) Why the licensee should not make any modifications to the facility piping systems indicated by such reanalysis to be necessary; and (3) Why facility operation should not be suspended pend-ing such reanalysis and completion of any required modifica tions.
The li:ensee's ~ response to the Order, dated April 2,1979, stated that it will reanalyze all potentially affected safety systems for seismic loads using an appropriate piping analysis cethod.
The licensee now requests that the Order be codified or rescinded such that the facility could be restarted based on the results of having analyzed all of th.e piping systems including nozzles and penetrations which previously used SH0CK 2, and all the corresponding piping supports.
In support of this request the licensee provided information by the March 21, 1930 letter,and the letter and the attached report dated February 22, 1910 v:hich documents the final results of all aspects of the analysis assc:iated with the Show Cause Order. It also identifies modifications re-latin; to the stress analysis of piping systems and pipe support evaluations.
A list of correspondence which provides supplemental information is also con:Sined in Appendix 0 of the report.
Dis:ussion The 5::ne and Webster (S&W) ? STRESS /SH0CK 2 computer code for pipe stress analyses sums earthquake 1cadings algebraically and is unacceptable for reasons is: f:r:h in the March 13, 1979 Order to Shcw Cause. This code was used in the D**D "D
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F seismic. analyses of certain safety and nonsafety related systems at the facility.
Tne li:snsee has identified the Seismic Category I systems at the fac.ility. analyzed with SH::K 2 and has rep rted the results of such reanalyses.
The basis of the licensse's' start-up request is the confidence of system operability during the seismic. events associated with the Design Basis. Earthquake (DBE) and the
. Operating Basis Earthquake (OBE).
We have evaluated the results of the seismic reanalyses and all the methods of pipe stress analysis previously utilized and used in the reanalyses for 'the facili ty.
Evaluation i
i.
Systems p:rtiens of the folicwing systems were identified by the licensee as;having
- ssn analyzed with 5H3CK 2.
Pressurizer Safety ar.d Relief Frassurizer Spray
':a Head Safety :nf ection
- -:f;h Head Safety Injection C:ntainment and ?.ecirculation Spray F.ssidual Heat Re. oval
- mponent Coolin; Water Service Water-Main Steam Hi;h Pressure Stsa-
?secwater k.xiliary Feedwa:er C:ntainment Vacu;m The licensee has rsanalyzed all 62 pipe stress problems originally analyzed by SH::K 2. _ The_ licensee's request for start-up is based on completion.of all pipe stress reanalysis and all resulting modifications installed prior to start-
- for all stress problems originally run on the SHOCK 2 computer program, and is also based on c
- mpletion of detail support analyses and resulting modifications ir.s;alled for all 5H
- CK 2 problems.
Of the.62 SH0CK 2 :roblems reanalyzed,17 required hardware modifications to bring the pipe stresses within allowables. These modifications consisted of 22 added, modified, Or deleted supports. The modifications include those r.5:sssary to the flexibility analysis of.the branch lines.
Also, modifica-ti:n, addition, or deletion of 57 supports on 17 problems were necessary to red;:e nozzle and ;snetration loads to acceptable levels.
Most of these m:difications are des :o differences between as-built and criginal design, while the remainin; c. as attributed, in par.t,-to the incorrect use of intra-
- isi cc binations in :he original seismic analysis. Support modifications
- - these problens ars listed in the report attached to the licensee's
?s:r;ary 22 le::er.
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.2. : Soil Structure Interaction Piping is analyzed in :st cases utilizin; amplified response spectra (ARS)
- nat are developed using soil structure interaction techniques (SSI-ARS).
The resultant stresses and leads are used to evaluate piping, supports, nozzles, and penetrations. Methods of soil structure interaction analysis which were acceptable to Surry Unit I are also applicable to Surry Unit 2.
In acccrdan:e with the NRC letters of May 25, 1979 and Ncvember 15, 1979 to Virginia Electric and ?cwer Company (VEPCO), the seismic inertial stresses and loads cc:?uted using the SSI-ARS have been increased by a facter of 1.5 for the 03E and 1.25 for.05E c:nditions.
3.
Verificatien of Analysis Methods We have reviewed the a::eptability of the analytical methods which are currently a basis fcr the facility piping design. The licensee has identifist the folicwing cc puter codes as a;plicable:
SU?!?E/S:cne & Webster NU?!?E/CDC NUPIFE/5 :ne & Webster In accordance with the letter of April 2, 1979 from V. Stello to the licensee, the licensee's Architect-Engineer, Stone and Webster (S&W) has submitted d: ucentation en the computer code NUPIPE which is being used in the reanalysis of the Surry Unit 2.
SLW has stated that this c:de calculates intramodal and intermodal resp nses according te the provision in Regulatory Guide 1.92. A review of the cod'e listing by the staff has confirmed this statement. The option used by che licensee specifies an intrat:dal cc bination consisting of the addition of the absolute value of the responses due to the vertical earth-quake ccaponent and the r:ot-mean-square combination of the response due to the two horizontal earthquake components. Additional documentation has also been submitted by the originators of this code (Quadrex) providing detailed information on the methods of modal combination.
The licenses has solved three NRC benchmark piping problems and its solutions show acceptable acreement with the benchmark solutions.
In addition, it provided a ccnfirmatory prcblem (No. 323A of Surry Unit 1 Safety Systems) to the Brockhaven National Lab for confirmatory solution.
A comparison of the solutions demonstrates good agreement (within about I C*i).
Eased en these considerations we find the use of-this code acceptable for seis.ic analysis by respcase spectrum techniques.
D"D D W Tf /3 NU?!?E/CJC In'acc:rdance with the le:ter of April 2, 19 9 frca V. Stello to VEFCO,""M "
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IJJ Ebas:: Services, Inc. has submitted docu en:atica en the computer code L7??E/ C C which is being used in the reanalysis of the Surry Unit 2 plant.
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_4 This code has previously been reviewed and has been found to satisfy the requirements of Regulatory Guide 1.92. Ebasco Services Inc. has solved three NRC benchmark piping problems and its solutions were found to agree closely with the benchmark solutions. They have also provided a comfirmatory problem (2503A) which v.as solved by the Brookhaven National Laboratory. Comparison of the solutions show good agreement.
Based on these results we find the use of NUPIPE/CDC by Ebasco Services, Inc. acceptable for seismic analysis by response spectra techniques.
4.
Reanalysis Methods and Results The safety related piping systems at the Surry 2 nuclear plant have been reviewed to determine the method of analyses.
Sixty two (62) computer stress problems of safety related piping have been identified where the analysis used the c:mputer code SH0CK 2 which used an algebraic intramodal summation of respenses to earthquake loadings. These problems have been reevaluated using a::sptable methods. The reevaluation included a dynamic computer analysis using UUPIPE programs, which incorporated a lumped mass response spectra modal analysis technique.
The floor response spectra used in the reanalysis include the original amplified respense spectra specified in the FSAR.
In some cases, piping was reanalyzed utilizing ARS that were developed using SSI techniques.
The peaks in the amolified floor response spectra were broadened by ! 15%
in'~accordance with ?egulatory Guide 1 122 to account for variation in material properties and approximations in modeling.
The piping systems were modeled as three dimensional lumped mass systems which included considerations of eccentric masses at valves and appropriate flexibility and stress intensification factors. The dynamic analysis pro-j cedures meet the criteria specified in the plant FSAR and are acceptable.
The resultant stresses and loads from the reanalysis were used to evaluate piping, supports, r.szzles, and penetrations.
All of the 62 SHOCK 2 pipe stress problems have been reanalyzed and will be verified by Ebas:o quality assurance, Stone and Webster Engineering Assurance and the licensee's Quality Assurance Program prior to start-up.
Eased on our review of the computer codes being used for reanalysis, independent check analysis performed by the. staff and a review of modeling methods usEd by the licensee, we find acceptable the precedures and methods used in reanalyzing these problems.
In the reanalysis, the new total stress, at the point of maximum total stress in the pipe, and new seismic stress, at the same point, were taken from the NUpIpE cer; uter runs with the seismic inertial stress magnified by a factor _ of 1.5 for the DBE condition for runs using the SSI-ARS, as
-required by "RC let er of May 25, 1979 to the licensee. For the OSE con-dition, a factor O' l.25 was used, in accordance with the NRC letter of November 15, 1979. Of the 62 problems 51 used the SSI-ARS and 1 used the
. Original ARS.
The s resses after the 1.5 and 1.25 magnification for the runs using SSI-ARS are below the allowable stresses.
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'To ensure that the pipe stress and pipe support reanalysis is performed.
as accurately.as possible, field verification of as-built conditions has been' performed.. The field verification produced detai, led piping isometric drawings and pipe-support sketches for each su,. cort upon which reanalysis is based.
All field-verified piping isometrics and pipe support sketches-are independently verified by Surry Power Station quality control-personnel.
The pipe Lsupports were reevaluated in cases where the original support design-loading was exceeded as a result of piping reanalysis.
In. cases where the original support capacity was exceeded, the support reevaluation ihas included:the censideration of base plate flexibility and a verification cf actual field ccnstruction of the support.. Where concrete expansion anchor bolts-were used, their capacities,'without. compromising the originally comitted safety margin, were also included in the reevaluation.
There are 702 suppcrts (252 inside the contair.mant, 220 outside the contain-ment) on lines originally analyzed by SHOCK 2, and all have been evaluated, at least as far as identification of necessary modifications is concerned.
Of the 482 supports inside the containment 166 supports were identified to recuire modifications.
Eighty-one supports outside containment are identified to require modification. During the reanalysis it was determined that 143 support modifications arose as a result of the "as-built" supports having deviated from the original design, whereas 104 support modifications can be qualifiec as due to inadequate, original seismic analysis incorporating algebraic summation technique.
L: ads on attached equipment nozzles and penetrations were checked and
.erified to be either below the allowable values or were made to be below
- ne allowable values by modification of supports.
For all the p oblems in
..hich the SSI-ARS are used, the seismic inertial nozzle loads have been increased by a factor of 1.5 for 08E per the NRC letter of May 2'5, 1979, and by a factor of 1.25 for OBE per the NRC letter of November 15, 1979.
Of the 62 problems reanalyzed, hardware modifications were made to 17 problems due to nozzle overload.
These modifications consisted of 57 added, modified, or deleted supports.
The pipe break criteria of the FSAR were reviewed in connection with the possible effect of changes of the high stress point resulting from the reanalyses..Only the main. steam lines were included in the stress re-analysis for pipe break.
Ea:h of the main steam lines has two terminal break locations, one at the containment ;enstration and the other at the main steam manifold. Each of the risers to the main steam relief valve headers has two terminal break location', one at the main steam lines, the other at the tee into the main s
steam header.
These terminal breakpoir.ts are' predetermined and are not changed as-a result of the stress reanalysis.
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Two intermediate break locations were originally determined based upon
' maximum primary plus secondary stresses.
Upon reanalysis, two additional breakpoints on each of the steam lincs were located. One-of these points is located immediately upstream of the check valve and the other point is at the elbow just downstream 6f the check valve.
All of these points will be included in the augmented inservice inspection program.
The piping syste s and supports were designed to the allowable limits of ANSI B31.1 for the gross properties and to the limits of At:SI B31.7 Appendix F for local stress considerations per the FSAR criteria.
The safety related piping system supports and. attached equipment, where the original analysis used an algebraic intramodal summation technique, have been reanalyzed with acceptable methods.
The procedures used in the support reanalyses and.their results have been reviewed against the
. criteria in the FSAR and found acceptable.
5.
Cenclusion
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The' licensee has demonstrated that SHOCK 2 is the only method of analysis used for the f acility's safety relav.:d systems which combined seismic loads algebraically.
Safety related piping systems analyzed with SHOCK 2 have been reanalyzed with an acceptable dynamic code.
Results of the reanalysis indicated that the pipe stress and equipment loads, after necessary support modifications, will be acceptable when compared with the FSAR allowables and the manufacturer's specified load criteria.
We reviewed the analysis techniques which are currently the bases for the facility's piping design.
We have determined that the application of these techniques, at Surry 2, assures that safety related systems will withstand
.both the-OBE and the OBE loading conditions. We therefore conclude that there is reasonable assurance that the facility can operate w1tnout encanger-ing the health and safety of the public.
This assurance is based on the following factors:
(1) All of the affected safety systems have been reanalyzed (piping, supports, nozzles, and penetrations) and were fcund either acceptable as presently designed or will be modified prior to startup.
(2) Confirmation of input data through "as-built" verification provides assurance that analytical results are correct and significant "as-built" deficiencies repaired.
Based on the above, we conclude that the conditions of the Show Cause Order of
- %rch 13, 1979, have been met.
Cate: March 21, 1980.
T-Tw: intermediate break locations were originally determined based upon aximum primary plus secondary stresses.
Upon reanalysis, two additional breakpoints en each of the steam lincs were _ located. One of these points is located imediately upstream of the check valve and the other point is at the elboyt just d:wnstream of the check valve. All of these points will
- e included in the. augmented inservice inspection program.
The piping systsms and ' supports were designed to the allowable limits cf ANSI B31.1 for the gross properties and to the limits of AtiSI B31.7 Appendix F for local stress considerations per the FSAR criteria.
The safety related ;iping system supports and attached equipment, where the original analysis used an algebraic intramodal summation technique, nave been reanalyzed with acceotable methods.
The_ procedures used in
- he support reanalyses and their results have been reviewed against the
- riteria in the FSAE and found acceptable.
5.
C relusion The licensee has de,onstrated that SHOCK 2 is the only method of analysis used for the f acility's safety related systems which combined seismic loads algebraically. Safety related piping systems analyzed with SHOCK 2 have been reanalyzed wi n an acceptable dynamic code.
Results of the reanalysis indicated that the oipe stress and equipment loads, after necessary support
.cdifications, will be acceptable when compared with the FSAR allowables and the manuf acturer's specified load criteria.
We reviewed the analysis techniques which are currently the bases for the f acility's piping casign. We have determined that the application of these
- e:hniques, at Surry 2, assures that safety related systems will withstand teth tne OSE and the OSE loading conditions. We therefore conclude that there is reasonable assurance that the facility can operate witnout endanger-ing the health and safety of the public.
This assurance is based on the.
fcilowing facters:
(i) All of the affected safety systems have been reanalyzed (piping, supports, nozzles,- and ;enetrations) and were found either acceptable as presently designed cr will be modified prior to startup.
(2) Confirmation cf input data through "as-built" verification provides assurance that analytical results are correct and significant "as-built" deficien:ies repaired.
Eised on the above, we conclude that the conditions of the Show Cause Order of Mar:5-13, 1979, hace 'een met.
Da:e:
March 21, 1332.