ML18130A598
| ML18130A598 | |
| Person / Time | |
|---|---|
| Site: | Surry, FitzPatrick  |
| Issue date: | 08/09/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML18130A589 | List: |
| References | |
| NUDOCS 7909190066 | |
| Download: ML18130A598 (9) | |
Text
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION e
FACILITY OPERATING LICENSE NO. DPR-59 POWER AUTHORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333
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e Introduction On March 13, 1979, the Commission issued an Order to Show Cause to the Power Authority of the State of New York (licensee) requ1r1ng that James A.
FitzPatrick Nuclear Power Plant (facility) be placed in cold shutdown and the licensee show cause:
(1)
Whj 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 reana~ysis to be necessary; and (3)
Why facility operation should not be suspended pend-ing such reanalysis and completion of any required modifications.
The licensee's response to the Order, dated March 30, 1979 (date of receipt) stated that it is reanalyzing al 1 potentially affected safety systems for seismic loads using an appropriate piping analysis method.
The licensee also requested that the Order be modified or rescinded such that the FitzPatrick Plant would be allowed to immediately resume ful~ power operation pending resolution of items set forth in said response.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 (Seismic 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 and 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-viously utilized and used in the reanalyses for the facility.
Evaluation
- 1. Systems Portions of the following systems were identified by the licensee as having been analyzed with SHOCK 2:
).
e e l. Standby Gas Treatment
- 2.
Control Rod Drive
- 3.
Resid~al Heat Removal
- 4.
- 5.
- 6.
Reactor Core Isolation Cooling
- 7.
- 8.
Reactor Building Cooling Water
- 9.
- 10.
High Pressure Coolant Injection
- 11.
Drywell Inerting, CAD and Purge
- 12.
- 13.
- 14.
- 15.
Chilled Water
- 16.
Fire Protection
- 17.
Combustion Air and Exhaust Emergency 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 modifications will be completed prior to startup. All of the analyses completed have included both the Operating Basis Earthquake (DBE) and Design Basis Earthquake (DBE) loadings.
A portion of the supports outside <::ontainment have been analyzed and the remainder will be reanalyzed within sixty (60) *days of the date of pl ant startup.
Prior to startup the 1 icensee will provide a schedule for completion of remaining support analyses. A goal of this schedule is completion of at least one train of all redundant safety systems within 30 days of plant startup.
Ninety one of the stress problems were determined to have pipe stress values after reanalysis, considering 11 as-built 11 conditions, within acceptable allowable v*a 1 ues.
The remaining 5 problems were resolved as follows:
(l) Problem 733 (A) (Drywell Vent and Purge) - The 11 as-built 11 inspection disclosed that a reinforcing pad on a 30 11 x 20 11 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 including all appropriate branches, one TEE was over-stressed when the stress intensification factor was considered.
Two snubbers Hl0-50N and Hl0-51CN were added to reduce the stress to within allowable.
(3)
Problem 657 (RHR-Head Spray) - Field verification disclosed that a valve was located about three feet and one 90 degree bend from its original analyzed position. This resulted in exceeding pipe stress allowables by 25% for the OBE case. Pipe stress was satisfactory for the DBE case. Snubbers-in-0-383N and Hl0-387 were added to resolve this problem.
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)
(4)
Problem 947 (A)(Fuel Pool Cooling) - Reanalysis indicated an 80%
over stress existed for the 0BE case.
Pipe stress was satisfactory for the DBE case. Supports Hl9-215N and FPSK 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 QS-160 (terminal anchor) was added.
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 basis for the facility piping design.
The licensee has identified the following computer codes/analysis methods 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 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 calculated 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 beam 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 piping fundamental frequency was relative to the structural frequency.
Calculated seismic stress was based on an assumed three component earthquake.
Support load-ings were based on standardized loadings enveloping the various loading conditions. Nozzle loads were calculated based on similar, simplified methods.
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Piping two. inches and below was shown on the piping drawings 11di agram-matically11 (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 loadings.
The problems where an algrbraic intramodal response combination technique was used in the design have been reevaluated using acceptable methods.
The reevaluation included a dynamic computer analysis using SHOCK 3, which incorporated a lumped mass response spectra modal analysis technique.
The floor response spectra used in the reanalysis was the original ampl i-fied 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 pi~ing systems were modeled as three dimensional lumped mass systems which included consideration of eccentric masses at valves and appropriate fiexibil ity and stress intensjfication 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 evaluate 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 perfonned by the staff and a review of modeling methods used by the licensee, we find acceptable the procedures and methods ~sed in reanalyzing these problems.
I&E Bulletin 79-04, 11Vel an Valve Weights 11, has been addressed and resolved for all 96 piping reanalyses.
At the request of the NRC, its consultant, EG&G INEL perfonned audit pipe stress calculations of five FitzPatrick problems using the NUPIPE computer code.
The results of the EG&G audit compare favorably with the results of the licensee's results.
,.. The pipeline support designs for affected system piping was inspected by the licensee to verify the location, orientation, support clearances, and support typ~.
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 as 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.
Where 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 operation of the facility.
Nine 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 evaluated.
During the reanalysis it was determined that the majority of the support modifications arose as a result of the 11 as-built 11 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 11U 11 bolt originally installed for a 1 ateral constraint. A 1 ateral overstress of 339% was determined.
The configuration of this support is 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 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 inform 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 /> and that the affected system be considered inoperable as specified in the facility Technical Specifications until the necessary modifications are implemented or a reanalysis assuming support failure is completed.
e Five supports 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.
JAFNP has been in operation for less than four years; therefore, considerable reserve margin exists. This is particularly true since these fatigue allowables 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 perfonnance 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 design and analysis of the supports and attached equipment are in accordance with the criteria specified in the plant FSAR.
The pipe break criteria of the FSAR was reviewed 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.l for the 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. Therefor~,
we find that the reanlaysis has not changed the pipe break protection.
The safety related piping 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.
Canel i.rs, on The licensee has demonstrated that SHOCK 2 is the only method of analysis used for the facility's safety related systems which combines 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
e necessary modifications,'will be acceptable when compared with the FSAR allowables and the ~anufacturer'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 11 as-built 11 condition which resulted in piping overstress. 1 These modifications are identified in Section 1.
The licensee will complete all modifications inside containment prior to plant operation. Evaluation 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 allowable, 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 more than twice the rated load for snubbers.
Use of detailed finite element analysis for evaluation of local stresses due to integral attachment is accept~
able.
Supports in accessable areas which exceed the factor safety of two to ultimate or the limits of ASME B&PV Code Section III, Subsection NF will be considered as inoperable 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 techniques at FitzPatrick assures that safety related systems will withstand the design basis 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 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 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 designed or will be modified as 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 support that would not meet the intent of a safety factor of 2.
It is therefore, reasonable to expect that few remaining supports would exceed a safety factor of 2.
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,..... (4)
ConfirmatJon of input data through 11 as-built 11 verification provides assurance that analytical results are correct and significant 11 as-built 11 deficiencies repaired.
(5)
The licensee has committed to complete analysis and any necessary modifications for the 28 remaining supports associated with High Pressure Core Injection and Reactor Core Isolation Cooling Systems prior to startup assuring that these ECCS systems and systems necessary for maintaining hot shutdown 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. Prior to startup the licensee will provide a schedule for completion of remaining support analyses. A goal of this schedule is completion of at least one train of all redundant safety systems within 30 days.
(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 committed to shut down the facility in the event of an earthquake which exceeds 0.01 g ac~eleration and inspect all piping, penetrations, supports and nozzles which have not been reanlayzed for both DBE 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 Show Cause Order of March 13, 1979.
Dated: