ML19275A417
| ML19275A417 | |
| Person / Time | |
|---|---|
| Site: | Surry, FitzPatrick  |
| Issue date: | 08/10/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19275A407 | List: |
| References | |
| NUDOCS 7910040367 | |
| Download: ML19275A417 (8) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION FACILITY OPERATING LICEf!SE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JAliES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333 1095 i':
7 910g4y y
Intreduction On l' arch 13, 1979, the Commission issued an Order to.show Cause to the Pcwer Authority of the State of flew York (licensee) requiring that James A.
FitzPatrick Nuclear Power Plant (facility) be placed in cold shutdom and the licensee 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 leads 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-s ing such reanalysis and completion of any required modifications.
The licensee's response to the Order, dated March 30,1979 (date of eceipt) stated that it is reanalyzing all potentially affected safety systers fer seismic loads using an appropriate piping analysis method.
The licensee also recuested that the Order be nodified or rescinded such that the Fit 9atrick Plant would be allowed to immediately resume full pcwsr cperation pending resolution of items set forth in said response.
Discussion The Stone and Webster (S&W) PSTRESS/ SHOCK 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 (Seisnic Category I) systems at the facility including those analyzed with SHOCK 2.
It has also identified the other methods of seismic analysis used for other Seismic Category I systems.
Furthermore, the licensee has reported the results of the reanalyses of SHOCK 2 safety systems ard has provided support for the acceptability of the analysis methods 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.
- valuation Systems crtions of the folicwing systems were identified by the 'icensee as W.ing been analyzed with SHOCK 2
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2-1.
Standby Gas Treatment 2.
Control Rod Drive 3.
Reactor Core Isolation Cooling 7.
Core Spray 8.
Reactor Building Cooling Water 9.
Fuel Pool Cooling and Cleanup 10.
High Pressure Coolant Injection 11.
Drywell Inerting, CAD and Purge 12.
Main Steam 13.
Feedwater 14.
Service Water 15.
Chilled Water 16.
Fire Protection 17.
Combustion Air and Exhaust Energency Diesel Generator The licensee has reanalyzed 96 pipe stress problems originally analyzed by SHOCK 2.
All supports in areas inaccessable during normal plant operation, including areas inside containment, were reanalyzed and nodifications will be completed prior to startup.
All of the analyses' completed have included both the Operating Basis Earthqua'Ke (OBE) and Design Basis Earthcuake (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.
val ues.
The remaining 5 problems were resolved as follows:
J' f.
(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 u
during construction.
By modification of the support H27-4, the 3 *-
stresses were reduced to within allowable for the unreinforced Tee.
.,d (2) Problem 650 (Residual Heat Removal) - Initial reanalysis of the proble -
o showed the pipe stress to be acceptable.
When this problem was remode;ed and reanalyzed including all appropriate branches, one TEE was over stressed when the stress intensification factor was considered.
Tv.
i' snubbers H10-50N and H10-51CN were added to reduce the stress to witnin al l owabl e.
(3)
Preblem 657 (RHR-Head Spray) - Field verification disclosed thct a valve was located about three feet and one 90 degree bend from its original analyzed position.
This resulted in exceeding pipe stress allowable:
by 25% for the OEE case.
Pipe stress was satisfactory for the DBE case.
Snubbers H10-383N and H10-387 were adced to resolve this cro. 2 7
1095 1 5 (4) Problem 947 (A)(Fuel Pool Cooling) - Reanalysis indicated an 80%
over stress existed for the OBE case.
Pipe stress was satisfactory for the D5h case.
Supports TITT-215N and TPSK 1000N were added to resolve this problem.
(5) Problem 909 (Control Rod Drive Cooling) - The interface between the Category I and non-safety related piping was independently defined.
To simplify the analysis, Suport 08-160 (terminal anchor) was added.
This was not a pipe stress problem but one of ensuring complete docunentation of safety related stress analyses.
2.
Verification of Analysis Methods We have reviewed the acceptability of the analytical methods which are currently a basis for the facility piping design. The licensee has identified the following computer codes / analysis methods as applicable:
PSTRESS/5 HOCK 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 due to the vertical earth-quake excitation to the (SRSS) combination of the response due to the two horizontal earthquake components. The intermodal components are cal cul ated by the SRSS method.
A review of the code listing has confirmed these statements.
S&W has also solved three benchmark piping problems provided by the NRC with this code, and its solutions show acceptable agreement with the benchmark solutions.
In addition, 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 methods were intended to keep the fundamental piping frequencies out of the range of the fundamental structural frequency by establishing span lengths between supports.
Cal-culations were based on simple bean formulations.
Tabulations relating various spans, nominal pipe sizes, and acceleration levels to actual pipe stress levels were provided for use by the analyst.
The acceleration applied to the piping was dependent upon where the pi;ing fundamental frecuency was relative to the structural frequency.
Calculated seismic s:ress was bas d on an assuned three conponent earthcuake.
Support load-ings were based on standardized loadings ence~ c:ing :te varicus loading ccnditicns.
Nczzle loads were calculated based on sinilar, simplified ethods.
1095 iM Piping two inches and below was show, on the piping drawings " diagram-matically"- (i.e., without detailed dimensions). The stress engineers located supports during the installation process working at the site with erection isometric sketches.
3.
Reanalysis Methods and Results The safety related piping systems at the FitzPatrick Nuclear Plant have been reviewed to determine the method of analyses.
Ninety six (96) computer stress problems of safety related piping have been identified where the analysis used the computer code SHOCK 2 which used an algebraic intramodal summation of responses to earthquake leadings. The problems where an algrbraic intramodal response combination technique was used in the design have been reevaluated using acceptable nethods.
The reevaluation included a dynamic computer analysis using SHOCK 3, which incorporated a lumped mass response spectra modal analy.,is technique.
The floor response spectra used in the reanalysis was the original ampli-fied response spectra specified in the FSAR.
The peaks in the amplified floor response spectra were broadened by +15% in acordance witn 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 consideration of eccentric masses at valves and 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 used to eval uate piping, supports, nozzles, and penetrations.
All of the 96 SHOCK 2 pipe stress problems have been reanalyzed and verified by Stone and Webster Engineering Assurance and the licensee's Quality Assurance Program.
This reanalysis completed the entire scope of piping stress reanalysis.
Based on our review of the computer codes being used for reanalysis, independent check analysis perfomed by the staff and a review of modeling methods used by the licensee, we find acceptable the procedures and methods used in reanalyzing these problems.
I&E Bulletin 79-04, "Velan Valve Weights", has been addressed and resolved for all 96 piping reanalyses.
At the request of the NRC, its consultant, EG&G INEL performed audit pipe stress calculations of five FitzPatrick problems using the NUPIPE computer code.
The results of the EGSG audit compare faverably with the results of the licensee's results.
1095 17 The pipeline support designs for affected system piping was inspected by the licensee to verify the location, orientation, support clearances, and support type.
Any deviations that were identified are incorporated into piping reanalyses.
These piping systems were also verified by the NRC Office of Inspection and Enforcement.
The pipe supports were reevaluated in cases where the original suppport design loading was exceeded es a result of piping reanlaysis.
In cases where the original support capacity was exceeded, the support reevaluation has included the consideration of base plate flexibility and a verifica-tion of actual field construction of the support.
<here concrete expansion anchor bolts were used, their capacities, without compromising the origi-nally committed safety margin, were also included in the reevaluation.
There are 989 supports on lines originally analyzed by SHOCK 2; of these all of the suports 335 in inaccessible areas including inside containment, and 273 of the supports in accessible areas have been evaluated and all necessary modifications will be completed prior to operaticn of the facility.
Nine new supports were added to the piping systens, and 29 of the existing 608 supoorts analyzed to date were identified to require modifications.
There c e approximately 381 supports remaining to be evaluated.
During the reanalysis'it was determined that the majority of the support modifications arcse as a result of the "as-built" supports which deviated from the original design.
Only one of them can be qualified as due to inadequate, 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 detemined.
The configuration of chis support is such that deformation would occur but it would remain functional. The anchor bolts are al so 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 exceed the limits of ASME B&PV Code,Section III, Subsection NF, or a factor of safety of 2 to ultimate.
If support reanalysis indicates that a support is inoperable, we have required the licensee to infom the NRC 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 /> at:d that the affected system be considered inoperable as specified in the facility Technical Specifications until the necessary modifications are impienented or a reanalysis assuming support failure is completed.
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. Five suppor :s in accessable areas exceed conservative local stress limits, at attachment welds. Modification of these supports is being made 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 endured throughout the 40 year lifetime of the plant.
JAFliP has been in operation for less than four years; therefore, considerable reserve margin exists.
This is particularly true since these fatigue allcwables normally contain a factor of safety of 20.
Should a DBE occur with its low number of stress cycles, the attachment welds would continue to function satisfactorily.
The only concern would be 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 should be noted that these supports are acceptable under the criteria to which they were designed.
Loads of attached equipment nozzles and penetrations were checked and verified to be either below the initial allowable values or were evaluated and determined to be acceptable.
Confirmation of the results of reanlaysis have been obtained from the equipment manufacturers where necessary.
The desig.n and analysis of the supports and attached equipment are in accordan:
with the criteria specified in the plant FSAR.
The pipe break criteria of the FSAR was reviewed in connection with the possib' effect of changes of the high stress point resulting from the reanalyses.
The piping systems and supports were designed to the allowable limits on At4SI B31.-
for the gross properties and to the limits of AISC structure steel code editior 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 find that the reanlaysis has not changed the pipe break protection.
The safety related oiping systems supports and attached equipment, where the original analysis used an algebraic intramodal summation technique, have been, or are to be reanlayzed with acceptable methods.
The pro-cedures 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 demonstrated that SHOCK 2 is the only method of analysis used for the facility's safety related systems which ccmbines seismic loads algebraically.
Safety related piping systens 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 1095 :D
. necessary modifications, will be acceptable when compared with the FSAR allowables and the manufacturer's specified loed 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" condition which resulted in piping overstress.
These modifications are identified in Section 1.
The licensee will complete all modifications inside ccr.tairment prior to plant operation.
tvaluation of the supports and schedule 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 code ailowable, the staff requires that the criteria used to determine whether a factor of safety of 2 to ultimate does exist be linear elastic analysis techniques or no nore than twice the rated load for snubbers. Use of detailed finite element analysis for evaluation of local stresses due to integral attachent is accept-able.
Supports in accessable areas which exceed the fac:cr safety of two to ultimate or the limits of ASME BSPV Code Secticn III, Subsection NF will be considered as inoperable as defined in the Technical Specifications.
kle 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 FitzPatrick assures that safety related systems will withstand the design basis earthquake.
Althcugh the reanalysis of supports outside containment is not ccmplete, 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 public.
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 SH0CK 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 designed or will be nodified es identified in this SER prior to startup.
Modifications which still remain are delineated in attachement 2 of the licensee's submittal of August 2,1979.
(3 ) The review of 608 supports identified 1 sup;cr
- .at would not meet the intent of a safety factor of 2.
It is therefere, reasonable to expect that few remaining supports would exceed a safety factor of 2.
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. (4)
Confirmation of input data through "as-built" verification provices assurance that analytical results are correct and significant "as-built" ceficiencies repaired.
(5) The licensee 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 carthquake.
(6)
The licensee has committed to complete all the support reanalysis in accessable areas outside containment within sixty (60) aays of the date of plant startup.
(7)
The probability of an earthquake exceeding the design basis earthquake curing the sixty (60) day period (or the thirty (30) day period for finalizing the reanalysis for one of the twL redundant trains for each safety system) that the remaining support analysis is being completed is small and the licensee has committed 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 prompt 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 Shod Cause Order of March 13, 1979.
Dated:
1095
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