ML20085B875
| ML20085B875 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 07/05/1983 |
| From: | Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8307080505 | |
| Download: ML20085B875 (6) | |
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PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 PHILADELPHIA. PA.19101 EDW A R D G. B AU ER, J R.
ano sans a6counsst EUGENE J. BR ADLEY assoconte esmemas couessk
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DON ALD BLANKEN RUDOLPH A. CHILLEMI E. C. MI R K H A LL T. H. M AHe:R CORN ELL PAUL AUERB ACH assesYamT SENERak Counsel EDW A RD J. CULLEN. J R.
THOM AS H. MILLER. JR.
IR EN E A. McM ENN A assestaa r couwssh Mr.
A.
Schwencer, Chief Docket Nos. 50-352 Licensing Branch No. 2 50-353 Division of Licensing U.
S.
Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
Limerick Generating Station, Units I and 2 I
Request for Additional Information (RAI)
From NRC Mechanical Equipment Branch l
File:
GOVT l-1 (NRC) l
Dear Mr. Schwencer:
Enclosed is a draft revision to RAI No. 210.33.
The information contained in this draft FSAR change will be incorporated inr.o i
the FSAR, exactly as it appears in the enclosure, in the revision scheduled for August, 1983.
Sincerely, h
Eug ne Bradley HDH/gra/75 Enclosure l
Copy to:
See Attached Service List
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8307080505 830705 PDR ADOCK 05000352 A
s
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cc: Judge Lawrence Brenner (w/o enclosure)
Judge Richard F. Cole (w/o enclosure)
Judge Peter A. Morris (w/o enclosure)
Troy B. Conner, Jr., Esq.
(w/o enclosure)
Ann P. Hodgdon (w/o enclosure) i Mr. Frank R. Romano (w/o enclosure)
Mr. Robert L. Anthony (w/o enclosure)
Mr. Marvin I. Lewis (w/o enclosure)
Judith A. Dorsey, Esq.
(w/o enclosure)
Charles W. Elliott, Esq.
(w/o enclosure)
Jacqueline I. Ruttenberg (w/o enclosure)
Thomas Y. Au, Esq.
(w/o enclosure)
Mr. Thomas Gerusky (w/o enclosure)
Director, Pennsylvania Emergency Management Agency (w/o enclosure) 1 Mr. Steven P. Hershey (w/o enclosure)
Donald S. Bronstein, Esq.
(w/o enclosure)
Mr. Joseph H. White, III (w/o enclosure)
David Wersan, Esq.
(w/o enclosure)
Robert J. Sugarman, Esq.
(w/o enclosure)
Martha W. Bush, Esq.
(w/o enclosure) i Spence W. Perry, Esq.
(w/o enclosure)
Atomic Safety and Licensing Appeal. Board (w/o enclosure) l Atomic Safety and Licensing Board Panel (w/o enclosure)
Docket and Service Section (w/o enclosure) i l.
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OUESTION 210.33 (Section 3.6.2.2.2, Page 3.6-40)
Provide the basis for assuring that the feedwater isolation check valves can perform their function following a postulated pipe break of the feedwater line outside containment.
RESPONSE
The basis for assuming that the feedwater isolation check valves can perform their function following a postulated pipe break of the feedwater line outside containment is described below.
he normal operating pressure of the valves is 1155 psig.
Each lve is designed, however, to withstand a differential pressur of 132 psi across the seat.
Design pressure, temperature, a ASM Code class are shown below:
1 Design Design ASM Valve P
ssure (psio)
Temperature (OF)
Code lass 1F010A,B 32 459 1
1F074A,B 2'
459 1
1F032A,B 213 459 2
l The valves are also set ically and dynam' ally qualified.
l If a break were to occur b ween valve 1F074 and 1F032 (Figure 5.1-3), redundant check valv 1F010 nd 1F074 (Figure 5.1-3) would both have to fail to ca a
OCA outside containment.
If a break were to occur upstream 1F032, redundant check valves IF010, 1F074, and 1F032 would v
to fail.
The probability of catastrophic failure of two thre of these valves accompanying the subject break is consi ed to be xtremely small.
A leakage detection sy em is provided i the reactor enclosure area containing the o outboard check val s 1F074 and 1F032 to alert the operator a leak so that correct' e action can be initiated.
Secti 5.2.5 contains a descript1 of leak detection provi ons.
A postulated pipe break Id be expected to provide wa ing indications and not an instant eou's double-ended failu that could theoretically generate unu ally large dynamic 1 ds.
Motor perated gate valve 1F011, located inside containme on the eedwater line, can be closed by an operator to isolate l
("
b en line if any of the check valves described above shoul l
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ail.
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Rev. [# 0)h83 1
9 ggg 210.33-1
T ZIO. 33 INSET @,
GuesTion 2io. 33 A study was performed which considered a feedwater pipe rupture outside of containment as part of the original project safety evaluations perf ormed ten years ago.
Analysis of a full circumferential break showed that the check valve inside containment (one of two check valves) would close rapidly because of flow reversal in the pipe between the reactor and break location just outside of con ta i nme nt.
This results in a pressure surge of 2780 psia between the valve and reactor.
It was determined that the piping and valves could adequately sustain this pressure.
To determine the capability of the check valve seat to withstand the initial impact caused by rapid closure and to sustain the pressure surge that follows, an estimate of the check valve seat er.ergy absorption capability was made for a postulated feedwater pipe break accident occurring upstream of the containment isolation valves.
The basis for the seat stress calculation was for a valve disk closing velocity of 100 rad /see and a pressure 4444A in the pipe I
T 5204-of 2780 psia.
The valve seat was assumed to consist of an assembly of six l
discrete, bilinear, elastic plastic elements.
The analysis assumed that the disc kinetic energy at impact equals work done in terms of seat under load, or area under seat load-displacement curve.
Valve seat yield strength was based on its being stressed to 50% of yield at a design pressure of 2132 psi.
The load displacement curve was cons tructed using Roark's stif fness equations for an annular plate loaded at the inner' radius and fixed at the outer radius.
Failure was assumed to occur at a ductility ratio of 30.
.T-50/1(6/83) 1
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The analysis indicated that all of the seat elements reach yield, but none reach ultimate strain, or none fail.
Effects not included which are believed to make the analysis conservative are:
o No credit was taken for dise deformation and energy absorption o No credit was taken for hinge de(ormation and energy a bso rpt ion o No credit was taken for valve body deformation and energy absorption o Hinge friction was omitted o No strain hardening or rate of strain ef fects in the seat was included.
The water hammer ef fects on the closed seat were determined.
It was shown that the natural frequency of the combined valve seat stiffness and disc mass was much larger than the frequency of the pressure pulse,.so that the effective pressure the seat must withstand is just the peak pressure in the waterhamrer surge.
The seat is able to withstand this maximum pressure.
Results of this analysis show that although the conditions of a hypothetical pipe rupture on the feedwater check valves the valves should remain together at impact, are severe, and are capable of withstanding 2800 psi.
The valve seat should yield at disc impact but not fail, and the water-hammer pressure pulse following closure will not cause failure.
In order to provide better documentation of the basis and confirm the assumptions used in the study performed ten years l
ago, a study was recently completec which usad a simplified model of single-phase, liquid flow from the reactor vessel to the pipe break, with the check valve disk being closed by drag forces.
Results of this simplified analysis indicate that the check valve disk closes in approximately 70 msee with a closing velocity at impact of approximately 65 rad /sec.
The peak the closed disk is estimated to be 2157 psi pressure at for a finite valve closure time of 70 msec.
The precise 2 **
T-50/l(6/83)
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f 2to.33 g-pressure at valve closure cannot be predicted rigorously by the simplified method used in this study.
The results of the study described above confirm the validity of the study performed as part of the original design and are consistent with the results of the analysis done for the Susquehanna plant and in part.icular for the Atwood Morrill check valves.
We conclude that these studies provide a sufficient basis for assuring that the feedwater check valves will perform their function following a postulated feedwater line break l
outside containment.
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T-50/1(6/83)
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