ML19208A840
| ML19208A840 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/14/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19208A839 | List: |
| References | |
| NUDOCS 7909170644 | |
| Download: ML19208A840 (9) | |
Text
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FACILITY OPERATING LICENSE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JA!.tES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 t D 0917 06cf d 9.,ona 4 ('an '~ '_N u
Introduction On March 13, 1979, the Commission issued an Order to Show Cause to the Power Authority of the State of New York (licensee) requiring that James A.
FitzPatrick Nuclear Power Plant (facility) be placed in cold shutdown and the I Hensee show cause:
(1) Why the licensee should not reanalyze the facility pipine systems for seismic loads on all potentially 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 (3) Why facility operation should r.ot be suspendea pend-ing such reanalysis and completion of any required modi fications.
The licensee's response to the Order, dated March 30,1979 (date of receipt) stated that it is reanalyzing all potentially affected safety systems for seismic loads using an appropriate piping analysis methoo. The licensee also requested that the Order be modified or rescinded such that the FitzPatrick Plant would be allowed to immediately resume full pcwer operation pending resolution of items set forth in said response.
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 SHOCK 2.
It has also identified the other methods of seismic andysis used for other 5...imic Category I systems.
Furthermore, the licensee has reported the results of the reanalyses of SHOCK 2 safety systems and has provided support for the acceptability of the analysis me tht,
- used on the remaining Seismic Category I systems.
We have evaluated the results of all the methods of pipe stress analysis pre-vicusly utilized and used in the reanalyses for the facility.
Evaluation 1.
Systems Portions of the following systems were ioentifiec y tre licensee as naving Deen analyzed witn SHOCK 2:
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. 1.
Standby Gas Treatnent 2.
Control Rod Drive 3.
Reactor Wat.
- leanup 6.
Reactor Core Isolation Cooling 7.
Core Spray 8.
Reactor Building Cooling Water 9.
- 11. Drywell Inerting, CAD and Purge
- 12. Main Steam
- 13. Feeawater 14.
- 15. Chillec Water 16.
Fire Protection 17.
Combustion Air and Exhaust Emergency Diesel Generator The licensee has reanalyzed 96 pipe stress problems originally analyzed by SH0CK 2.
All supports in areas inaccessable during normal plant operation, including areas inside containment, were reanalyzec and mooifications will ca completed prior to startup. All of the analyses completed have included both the Operating Basis Earthquake (0BE) anc Design Basis Earthquake (DBE) loadings.
A portion of the supports outside containment have been analyzed and the remainder will be reanalyzed within sixty (60) days of the date of plant startup.
Ninety one of the stress problems were determined to have pipe stress values after reanalysis, considering "as-built" conditions, within acceptable allowable values.
The remaining 5 problems were resolved as follows:
(1) Problem 733 ( A) (Drywell Vent and Purge) - The "as-built" inspection disclosed that a reinforcing pad on a 30" x 20" Tee had been omitted during construction.
By modification of the support H27-4, the stresses were reduced to within allowable for the unreinforced Tee.
(2) Problem 650 (Residual Heat Removal) - Initial reanalysis of the problem showed the pipe stress to be acceptable. When this problem was remodeled and reanalyzed inclucing all appropriate branches, one TEE was over-stressed when the stress intensification factor was consicered. Two snuboers H10-50N and H10-51CN were added to recuce the stress to within allowable.
(3)
Problem 657 (RHR-Head Spray) - Fielo verification ciscloseo tnat a valve was located aoout three feet anc one 90 otgree cenc frcm its original analyzed position. This resulted in exceecing pipe stress allowaoles by 25% for tne OBE case. Pipe stress was satisfactory for tne DBE case.
Snubbers n10-383N ano H10-387 were acces to resolve :nis proolem.
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. (4) Problem 947 (A)(Fuel Pool Cooling) - Reanalysis indicated an 80%
over stress existed for the OBE case. Pipe stress was satisfactory for the DBE case. Supports iiTG-215N and FPSK 1000N were added to resolve this problem.
(5) Problem 909 (Control Rod Drive Cooling) - The interf ace between the Category I and non-safety related piping was independently defined.
To simplify the analysis, Suport Q8-160 (terminal anchor) was acced.
This was not a pipe stress problem but one of ensuring complete documentation of safety related stress analyses.
2.
Verification of Analysis Methods We have reviewed the acceptability of the analytical methods which are currently a Dasis for the facility piping aesign..The licensee has identified the following computer codes / analysis methocs as applicable:
PSTRESS/ SHOCK 3 Static Analysis Methods PSTRESS/ SHOCK 3 S&W has stated that PSTRESS/ SHOCK 3 calculates the intramodal responses by adding the absolute value of the response cue to the vertical earth-quake excitation to the (SRSS) combination of the response due to the two horizontal earthquake components. The intermoaal components are calculated by the SRSS method. A review of the code listing has confirmed these statements.
S&W has also solved th'ree benchmark piping problems provided by the NRC with this code, and its solutions show acceptable agreement with the benchmark solutions.
In accition, a comparison of the S&W and BNL solutions of the confirmatory problem also demonstrate good agreements (within 10%).
Static Analysis Much of the 6 inch and smaller Category I piping at FitzPatrick was analyzed using simplified static methods. The metnods were intended to keep the funcamental piping frequencies out of the range of the funcamental structural frequency by establishing span lengths oetween supports. Cal-culations were based on simple oeam formulations. Taoulations relating various spans, nominal pipe sizes, and acceleration levels to actual pipe stress levels were provided for use by the analyst.
The acceleraticn appliea to the piping was cependent upon where the piping fundamental frequency was relative to tne structural frequency. Calculated seismic stress was cased on an assumed three component eartnquaKe.
Support loac-ings were cased on stancarcizea loadings enveloping the various loacing concitions. Nozzle loacs were calculatec casec on similar, simplified methods.
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. Piping two inches and below was shown on the piping drawings " diagram-matically" (i.e., without detailed dimension,). The stress engineers located supports during the installation process working at the site with erection isometric sketches.
3.
Reanalysis Methods and Results The stfety related piping systems at the FitzPatrick Nuclear Plant have been re;iewed to determine the method of analyses. Ninety six (96) computer stress problems of safety related piping have Deen identified where the analysis used the computer code SH0CK 2 which used an algeDraiC intramodal summation of responses to earthquake loadings. The proolems where an algrbraic intramodal response combination technique was used in the design have Deen reevaluated using acceptaole methods. The reevaluation includea a dynamic computer analysis using SH0CK 3, which incorporated a lumped mass response spectra modal analysis technique.
The floor response spectra used in the reanalysis was the original ampli-fiec response spectra specified in the FSAR. The peaks in the amplified floor response spectra were broadened by +15*, in acordance with Regulatory 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 consiceration of eccentric masses at valves ano appropriate flexibility and stress intensification factors.
The dynamic analysis procedures meet the criteria specified in the plant FSAR and are accept-able. The resultant stresses and loads from the reanalysis were usec to evaluate piping, supports, norzles, anc penetrations.
All of the 96 SHOCK 2 pipe stress probleins have oeen reanalyzed ana verified by Stone anc webster Engineering Assurance and the licensee's Quality Assurance Program. This reanalysis completed the entire scope of piping stress reanalysis. Based cri cur review of tne computer codes being used for reanalysis, indepenaent check analysis performed by the staff and a review of medeling methods used by the licensee, we find acceptable the proceaures and methods used in reanalyzing these problems.
I&E Bulletin 79-04, "Velan Valve Weights", has been accressed and resolved for all 96 piping reanalyses.
At the request of the NRC, its consultant, EG6G INEL performed audit pipe stress calculations of five FitzPatricK problems using the NUPIPE computer coce.
The results of the EG&G audit compare favoraoly witn the results of the licensee's results.
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The pipeline support designs for affectet system piping was inspected by the licensee to verify the location, orientation, support clearances, and support type. Any deviations that were identifiea are incorporated into piping reanalyses. These piping systems were also verified by the t4RC Office of Inspection and Enforcement.
The pipe supports were reevaluated in cases where the original suppport design loading was exceeded as a result of piping reanlaysis.
In cases where the original support capacity was exceeded, the support reevaluation has included the consideration of Dase plate flexibility and a verifica-tion of actual field construction of the support. Where concrete expansion anchor bolts were used, their capacities, without compromising the origi-nally committed safety margin, were also incluced in the reevaluation.
There are 989 supperts on lines originally analyzea by SHOCK 2; of these all of the suports 335 in inaccessible areas including insice containment, and 273 of the supports in accessible areas have been evaluated and all necessary mocifications will be completed prior to operation of the facility.
tiine new supports were added to the piping systems, and 29 of the existing 608 supports analyzed to date were identified to require modifications.
There are approximately 381 supports remaining to be evaluatea. During the reanalysis it was determined that the majority of the support modifications arose as a result of the "as-built" supports which deviated from the original aesign. Only one of them can be qualified as cue to inacequate, original seismic analysis incorporating algebraic summation technique.
One support in inaccessible areas would not meet the intent of a factor of safety of 2.
This support was a "U" bolt originally installed for a lateral constraint. A lateral overstress of 339% was determined.
The configuration of this support f 5 such that deformation would occur but it would remain functional. The anchor bolts are also 104% overstressed but still have a safety factor of 2.5.
Based on the results to date, we expect other supports may be found that will be above allowable limits.
In the event the loads on a pipe support exceed allowable loads, the support will be considered operable if its loads do not exceeo the limits of ASME B&PV Coce,Section III, Subsection 14F, or a factor of safety of 2 to ultimate.
If support reanalysis incicates that a sunport is inoperable, we have required the licensee to inform the f4RC of the results of reanlaysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and tnat the affectea system De considered inoperaDie as specified in the facility Tecnnical Specifications until the necessary modifications are implementea or a reanalysis assuming support failure is completea.
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y,mm ad i f.6 L. f Five supports in accessable areas exceed conservative local stress limits, at attachment welds. Modification of these supports is being mace to satisfy these limits. It should be recognized that such local stresses are calculated based on theoretical elastic response. Further, these limits are purposely set quite low to enable the weld to withstand the fatigue cycles which will be enuured throughout the 40 year lifetime of the plant. JAFNP nas oeen in operation for less than four years; therefore, considerable reserve margin exi sts. This is particularly true since these fatigue allowables normally contain a factor of safety of 20. Should a DBE occur with its low numDer of stress cycles, the attachment welds would continue to function satisfactorily.
The only concern would De a shortening of the weld's fatigue life which in no way would affect its ability to withstand an earthquake.
The performance of the overall support structure should remain elastic.
It snould be noted that these supports are acceptable under the criteria to which they were designed..
Loaas of attached equipment nozzles and penetrations were cnecked and verified to be either Delow the initial allowaDie values or were evaluated and determined to De acceptable. Confirmation of the results of reanlaysis have been obtained from the equipment manufacturers where necessary.
The design and analysis of the supports and attached equipment are in accordance with the criteria specified in the plant FSAR.
The pipe Dreak criteria of the FSAR was reviewec in connection with the possible effect of changes of the high stress point resulting from the reanalyses. The piping systems and supports were designed to the allowable limits on ANSI B31.1 for tne gross properties and to the limits of AISC structure steel code edition six.
Results of the evaluation of the effect the reanalyses has on the FSAR pipe break criteria show that no new whip restraints are required. Therefore, we fina that the reanlaysis has not changed the pipe break protection.
The safety related piping systems supports and attachea equipment, where the original analysis used an algebraic intramoaal summation technique, have been, or are to be rEanlayzed with acceptable methods. The pro-ceaures used in the support reanalyses and their results have been reviewed against the criteria in the FSAR and found acceptable.
4.
Conclusion The licensee has cemonstrated that SHOCK 2 is the only method of analysis usea for the facility's safety related systems which comoines seismic loads algebraically. Safety related piping systems analyzea with SHOCK 2 have Deen reanalyzea with an acceptaDie cynamic coce. Results of the reanalysis inoicatea that the pipe stress ano equicment ioacs, af ter
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352210 necessary modifications, will be acceptable when compared with the FSAR allowables and the manufacturer's specified load criteria.
The reevaluation of pipe stress problems indicated that modifications in three problems were necessary as a result of the seismic reanalysis.
One problem was modified due to an "as-built" conoition which resulted in piping overstress. These modifications are identified in Section 1.
The licensee will complete'all modifications inside containment prior to plant operation. Evaluation of the supports and scheoule for completion of necessary modifications outside of containment will be completed within sixty (60) days of the date of the Order. Further, in those cases where reanalysis exceeds coae allowable, the staff requires that the criteria used to determine whether a fa.sor of safety of 2 to ultimate coes exist be linear elastic analysis techniques or no more than twice the rated load for snuceers. Use of cetailed finite element analysis for evaluation of local stresses due to integral attachment is accept-able. Supports in accessable areas which exceed tne factor safety of two to ultimate or the limits of ASME B&PV Code Section III, Subsection NF will be consicered as inoperaole as defined in the Technical Specifications.
We reviewed the analysis techniques which are currently the bases for the facility's piping design. We have determined that the application of these tecnniques at FitzPatrick assures that safety relatea systems will withstano the cesign Dasis earthquake. Although the reanalysis of supports outside containment is not complete, there is reasonable assurance that the facility can operate during the interim period until the reanalysis and any required modifications are completed without endangering the health and safety of the puolic.
This assurance is based on the following factors:
(1) All safety system piping both inside and outside containment which was originally seismically analyzed with the SHOCK 2 program has been reanalyzed and, subject to modification, is, or was made, acceptable.
(2) All of the affected safety systems inside containment have been reanlayzed (piping, supports, nozzles, and penetrations) and were found either acceptable as presently designea or will be modified as identified in this SER prior to startup. Modifications whicn still remain are celineateo in attachement 2 of the licensee's suDmittal of August 2,1979.
(3)
The review of 608 supports icentified i support that woulo not meet the intent of a safety factor of 2.
It is tnerefore, reasonaDie to expect that few remaining supports woulo exceeo a safety f actor of 2.
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(4) Confirmation of input data through "as-built" verification provides assurance that analytical results are correct and significant "as-built" deficiencies repaired.
(5) The licensea has completed the analysis for the High Pressure Core Injection and Reactor Core Isolation Cooling Systems assur-ing that these ECCS systems and systems necessary for maintaining hot shotdown will be capable of withstanding a design basis earthquake.
(6) The licensee has committed to complete all the support reanalysis in accessable areas outside containment within sixty (60) days of the date of plant startup.
(7) The probability of an earthquake exceeding the design basis earthquake during the sixty (60) day period (or the thirty (30) day period for finalizing the reanalysis for one of the two redundant trains for each safety system) that the remaining support analysis is being completed is small and the licensee has comnitted to shut down the facility in the event of an earthquake which exceeds 0.01 g acceleration and inspect all piping, penetrations, supports and nozzles which have not been reanlayzed for both OBE and DBE.
(8) The NRC will require prcmpt notification of inoperable supports within twenty four (24) hours and either resolution by reanalysis of the piping system assuming a failed support or modification of the affected support, if reanalysis of a support exceeds the factor of safety of two to ultimate and the limits of ASME B&PV Code,Section III, Subsection NF.
Based on the above, we conclude that the licensee has shown cause why FitzPatrick can be operated for 60 days pending completion of reanalyses required by the Show Cause Order of March 13, 1979.
Dated: August 14, 1979 3 4,112