ML19274F298

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To Rept on Reanalysis of Safety-Related Piping Sys, Prepared by S&W
ML19274F298
Person / Time
Site: Beaver Valley
Issue date: 07/17/1979
From:
DUQUESNE LIGHT CO.
To:
Shared Package
ML19274F297 List:
References
NUDOCS 7907200476
Download: ML19274F298 (13)
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"

3EAVER VALLEY POWER STATICN, UNIT 1 REPORT ON THE

'. ANALYSIS OF SATETY-RELATED PIPING SYSTEMS FOR 3EAVER VAlfEY UNIT 1 DUQUESNE LIGHT COMPANY ORIGINAL - JUNE 15, 1979 REVISION 2 - JULY 17,1979 7907200 76 Stone X Webster Ingineering Corporation 30ston, Massachusetts

3EAVER VALLEY FCWER STATION, UNIT 1 SECTIC!T 1 SUICfARY AND CONCLUSIONS In response to the Nuclear Regulatory Commission's Order to Show Cause, dated

!! arch 13,1979, a reanalysis is being conducted of safety-related piping systems which were originally dynamically analy:ed using the SHCCK2 computer program. This program, which used an earlier load combination methodology, is no longer considered acceptable by the NRC.

This report specifically addresses the return to operation for an interim period which is defined as the period during which the remaining 6 to 7 weeks of nuclear fuel in the first cycle is expended. The minor modifications required to ensure that all pipe stresses are within licensed code allowables for the D3E case vill be l

completed prior to startup for interic operation, rimilarly, all modifications required to ensure that the stresses in all pipe supports are within acceptable allowables for the DBE case vill be completed prior to startup (see Section 5 and Table 5-2).

Any modifications to pipe supports that are required to satisfy the OBE requirements vill be made during the refueling outage that will commence as soon as the present nuclear fuel is expended.

1-1 Revision 2

BEAVER VALLEY POWER STATION, UNIT 1 Problem 171, the supply to the CC heat exchangers has one support (H-65) out of 15 which has a local overstress of 4 percent.

Problem 270, the dischargy from the CC heat exchangers has one support (M-56) out of 13 which has a loc.a1 overstress in a lug of 78 percent.

It is considered that, if a DBE vere to occur, failure of these two supports would not cause a system rupture or a resultant loss of function.

The fuel pool cooling and puriiteation system is presently isolated since there is no spent fuel being stored.

The quench and recirculation spray systems have been completely analy:ed for DBE and water hammer loads using NUPIPE and are acceptable.

The OBE case vill be run in the long term.

3.

How have stress intensification factors been applied at branch connections during the reanalysis?

Reseense Appropriate stress intensifications from 331.1 have been applied to the run pipe at reduced cutlet branch connections.

3 ranches which are uncoupled have been evaluated for the effects of the movements of the run 3-25 Revision 2

3EAVER VALLEY POWER STATION, UNIT 1 SECTION 4 PIPE STRESS RESULTS A total of 121 pipe stress problems have been identified for reanalysis and l

are being analyzed by Stone a Webster Engineering Corporation in Boston,

& ssachusetts.

The pipe stress reanalysis consists of substituting the SHOCK 3 or NUPIPE code for tie SHOCK 2 code.

SHOCK 3 is a current seismic code that calculates both intramodal and intermodal seismic forces using a modified square root of the sum of the squares (SRSS) technique and an SRSS technique, respectively, rather than an algebraic summation.

The NUPIPE Program utilizes modal response combinations as follows:

Intermodal - SRSS for combination, grouping for modal combination (where closely spaced modes are combined by absolute sun).

Intramodal - SRSS for direction combination.

4-1 Revision 2

BEAVER VALLEY P0"ER STATION, UNIT 1 Field verified piping fabricator iscretric drawings provide the basis for program inputs for the pipe stress reanalysis.

Additionally, in some cases, piping is analy:ed utilizing amplified response spectra (ARS) that are developed using soil structure interaction t e cht.ique s (SSI-ARS).

The resultant stresses and leads are used to evaluate piping, supports, no::les, and penetrations.

These techniques are discussed in Section 8.7.

Of the 1 1 SHCCK2 problems, 93 have been reanaly:ed and are within allevable l

stress values. Table 4-1 lists the problems including the peak stress values for the SHOCK 3 and NUP:PE pipe stress runs.

Stresses were computed by the SHOCK 3 cr NU?IPE program using different mass models and in scre cases different ARS than the original calculations.

More importantly, the reanalyses were based on field-verified, as-built cenditions which in some cases differ significantly frem the original design conditions.

?cr these reasons, the originally calculated stresses are not ec= parable to the new stresses.

Table e-summarizes the no::les and penetrations evaluated under the reanalysis program.

Of a total of 87 no::les en problems within the secpe of the interir effort, 82 have been evaluated and fcund to be acceptable, and 5

-2 Revisien

3EA'lER VALLEY PC'JER STATION, UNIT 1 TABLE 4-1 (Cont)

System A110Vable Reanalysis and Stress

. Maxi: ;a Teanalysis Problem No.

fosi)

__ Stress Method 159'**

19.000 10.443 NUFIPE/SSI-ARS 27,000 17,277 129 19.000 10.760 NUFIPE/SSI-ARS

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127 19.000 13.394 NUPI?E/SSI-ARS 27,000 15,970 125 19.000 10.760 NUPIPE/SSI-ARS 27,000 12,56:

1:4 19.000 13.39a NUPIFE/SSI-ARS 27,000 15,970 123 19.000 10.861 NUPIFE/SSI-ARS 27,000 17,797 120 19.000 57' NUFIPE/SSI-ARS 27,000 16,471 106 19.000 13.?9a NUFI?E/SSI-ARS

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6 ef S Revision

3EAVER VALLEY PC'JER STATION, UNIT 1 TA3LE 4-1 (Cont)

System Allowable Reanalysis and Stress "aximum Raanalysis Prebler No.

(esi)

Stress "ethod Diesel Generator Exhaust 651 12.960 1.201 NUPIFE/SSI-ARS 19,440 1,717 Arplified respense spectra developed using soils Notes:

SSI-ARS :

stru:ture interaction techniques Stresses shown are Oeerational Pasis Earthcuake (03E) Stresses Design Basis Earthquake (DBE) Stresses

TP30a/IP316 allevables

'2' After =cdification

'** Froblens are no longer within scope of short-term reanalysis effort.

See Appendix 3.

+ S Preblens 213 and 2113 include s

+SLP + S and SDL + SLP DBEI DL OBEI c. _, y.

'5' 3eing terun with SSI-ARS.

'68 Froble: 159 in:ludes Problems 160 and 161.

Evaluated for the D3E case only (S

+S

+ 5)

'7'

)

DL LP 3 Of 5 7,evisien

BEAVER V AL' EY FO'n'ER STATION, UNIT 1 SECTION 5 FIFE SUFFORT RESULTS Table 3-1 summarizes the pipe supports evaluated in the reanalysis program.

!here are 696 pipe supports on lines within the interim reanalysis effort; of these.

SOS have been evaluated and found acceptable and 7 have been modified to be acceptable.

A support is considered acceptable if all the load components are lower in magnitude than those for which the support was originally designed.

If some load components are greater than the original design load components, the support is reanaly:ed using the new loads.

Of the total 181 supports requiring reanalysis, 145 have been found to be ace.eptable based on DBEI+DL, 36 have not been accepted at this time.

There is sufficient analytical information available for the reinaining 36 supports to exercise engineering judgment in determining whether the unacceptable condition will become acceptable.

1.

The use of ASME III Section NF faulted allowable stress values for structural members 5-1 Revision 2

BEAVER VALLEY FO'JER STATION, UNIT 1 TABLE B-2 SATETY SYSTEMS TO 3E ANALYZED IN THE LONG TIEM Problem FSAR systen Eine Nup ar No.

Fie. No.

Tuel ? col Cooling 6"-TC-4-152-Q3 104 4 Furification System 6"-TC-5-152-Q3 g

(FC) 6"-TC-6-152-Q3 6"-TC-8-152-Q3 105E 9.5-1 6"-TC-9-152-Q3 10"-FC-1-152-Q3 198B 9.5-1 6"-TC-2-152-Q3 6"-TC-31-152-Q3 4"-TC-10-152 107 9.5-1 4"-TC-11-152 6"-TC-14-152 6"-FC-17-152 6"-FC-32-152 Quench Spray 10"-QS-4-1533-Q3 614 6.4-1 (QS) 10"-QS-3-1533-Q3 617 6.4-1 4"-QS-6-1533-Q3 210 6.4-1 10"-QS-4-153B-Q3 4"-QS-5-1533-Q3 1 '. 8 6.4-1 10"-QS-3-1533-Q3 10"-QS-3-1533-03 134

6.4-1 10"-QS-3-153t-Q3 135

6.4-1 10"-QS-4-1533-Q3 31

6.4-1 10"-QS-3-1533-Q3 23

6.4-1 10"-QS-4-1533-Q3 8"-QS-22-1533-Q3 8"-QS-23-1533-Q3 1 of 2 Revision 2

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3EAVER VALLEY PO*JER STATION, UNIT 1 LIST OF TSAR FIGURES Ti2ure Title 6.4-1 Contain?.ent Depressurization System l

9.4-1 Component Cooling t7ater System (Sheet 1) 9.4-2 Component Cooling Water System (Sheets 1 8 2) 9.5-1 Fuel Fool Cooling and Purification System 9.9-1B Intake Structure l

D-1 Revision 2

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CONTAINMENT DEPRESSURIZATION SYSTEM BEAVER VILLEY POWER STATION FIN AL' SAFETY ANALYSIS REPORT

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.H REVISION 2

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