ML19224C397

Submits Supplemental Info Re Piping Stress Reevaluation. Request for Plant Startup Was Contingent Upon Completion of Rept & Necessary Mods to Constraints.Expected Date Now 790715

ML19224C397
Person / Time
Site:	FitzPatrick
Issue date:	06/29/1979
From:	Early P POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	Ippolito T Office of Nuclear Reactor Regulation
References
JPN-79-38, NUDOCS 7907020270
Download: ML19224C397 (49)

Text

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POWER AUTHORITY OF THE STATE OF NEW YORK 10 COLUMBUS CIRCLE NEW YORK N. Y.10019 (212) 397 6200 TRUSTEES GEORGET. BERRY FREDERICK R. CLARK cwainuaw THOMAS R. FREY

1. ocuraatcouwen GEORGE L. ING ALLS 1 JOSEPH R. SCHMIEDER I,

RICH ARD M. FLYN N Culf F ENGINEER ROBERT 1. MILLONZI ge o wituAM F. tuooY June 29, 1979 JPN-79-38 " ^co,* 0."L t .

Director of Nuclear Reactor Regulation United States Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 3 Division of Operating Reactors

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Supplemental Information Supporting Request for Plant Start-Up

Dear Sir:

A request for plant start-up about July 1, 1979 was submitted on June 8, 1979. This request was contingent upon completion of the piping stress reevaluation effort and any necessary modifi-cations to constraints in inaccessible areas. A program plan was provided.

This supplemental information is submitted both as a result of discussions with the NRC staff (Messrs. Polk, Noonan, and Fair) on June 19, 1979 and to report progress made since June 8, 1979.

The Authority reaffirms its intent to complete the reevaluation effort of pipe stresses and pipe constraints in inaccessible areas and repeats its request for authorization to start up JAFNPP. The expected date is now July 15, 1979.

The status of the effort is shown below. Tabulations of the results are contained in Attachments 1 through 4.

1. PIPE STRESSES b All pipe stresses for the 96 piping lines are within allow-able limits. Since the Authority's June 8 letter was sub-mitted, branch piping having a moment of inertia greater n10 7 907020 A -

U. S. Nuclear Regulatory Commission JPN-79-38 than 10 percent of the run pipe moment of inertia were remodelled and incorporated into the reevaluation of the run pipe. As a result, 17 problems were recombined into 6 analysis packages and reevaluated for stresses. These packages are indicated in Attachment 1. This reevaluation again confirms the determination that the safety-related piping is within allowable limits. The tabulated pipe stress results are shown in Attachment 2.

2. PENETRATIONS All loads on the 56 penetrations involved in the piping lines are within allowable limits (see Attachment 3) .

3. NOZZLES All loads on 50 of the 83 equipment nozzles involved in the piping lines are within allowable limits. The loads on the 33 remaining nozzles, subject to vendor confirmation, are also within allowable limits. Confirmation by the vendors is expected by July 2, 1979 (see Attachment 3).

4. PIPING SUPPORTS All loads on 239 of the 342 piping supports in inaccessible areas are within allowable limits. Twenty-six supports tentatively are scheduled for modification and 77 supports remain to be evaluated.

Of 656 piping supports in the accessible arcas, the loads on 243 are within allowable limits, and 12 are designated to be modified. Evaluation of the remaining 401 supports is continuing. The Authority's priorities for reevaluation in the accessible areas are noted in Attachment 1.

Of the pipe supports being modified, 9 pipe supports in the inaccessible areas and 5 in the accessible areas were designated for modification, not as a result of the pipe stress reevaluation, but as required to correct as-built deviations and for plant conformance with its as-designed basis.

In summary, of the total 998 piping supports involved, 482 have been evaluated acceptable and 24 are to be modified due to pipe stress reevaluation.

2171 077

U. S. Nuclear Regulatory Commission JPN-79-38 In order to estimate the number of structural modifications that might be required for pipe supports in the accessible areas, and to provide an assessment of the expected integrity of the accessible systems, a closer look was taken at the pipe supports which the Authority decided to modify due to pipe stress reevaluation in the inaccessible areas.

This examination tentatively indicates that:

1. As indicated in the June 8 letter, the stress analysis of piping incorporates numerous conservatisms for which no cumulative credit has been taken. (See Attachment 5)

2. Six of the supports are trunnions, which were originally designed utilizing basic analytical techniques and for which modifications are now being planned to satisfy local stress conditions. These trunnions are still acceptable based on the macro-stress criteria to which they were designed and will withstand the reevaluated loads.

3. One of the supports is within allowable loads for the combination of pressure + thermal + deadload + DBE conditions, although they have been designated for modification because they do not pass the OBE criteria.

For this support, the OBE cr 4teria was exceeded by 40 percent of the allowable.

4. Three supports are within 2.4 Sh (present ASME Code cri-teria for DBE). Five supports are not subjected to loads in excess of their ultimate strength and will retain their structural integrity under all projected load conditions.

Of the remaining 2 supports being modified in the in-accessible areas, preliminary indications are that even if none of these supports provide any seismic restraint, the associated piping integrity will not be compromised.

Nothing in this analysis has identified any condition, as designed or built, which would endanger plant. integrity or provide a basis for believing that JAFNPP would not withstand a seismic event. Final confirmation of the above will be transmitted prior to start-up.

Thus, extrapolation of the completed constraint evaluations to the accessible areas does not provide a basis for anticipating that any of the unresolved constraints in the accessible areas will render a critical system inoperable. On this basis, as well as that presented in the June 8 letter, the Authority believes 271 078

U. S. Nuclear Regulatory Commission JPN-79-38 that JAFNPP should be permitted to start up prior to the completion of the seismic reevaluation flowing from the Show Cause Order.

After restart, if a constraint design evaluation indicates that such constraint may not be able to perform its intended function, and the Authority's analysis confirms that this inadequacy renders a safety-related system inoperable, the Authority will notify the NRC thereof within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the determination. These safety-related systems are those which have plant shutdown requirements in the Technical Specifications in the event the system is declared inoperable. Repair of the deficiency shall be completed or justi-fication for continued operation will be provided to the NRC, within 7 days or in accord with the appropriate plant technical specifica-tions, whichever is less. If the above requirement cannot be met, the reactor will be placed in a cold shutdown condition within an additional 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The completed reevaluation analyses and modification engineer-ing documentation can be made available for NRC staff review. As the remaining analyses and modificaticns are completed, these will also be made available.

Finally, as requested by the NRC at the June 19 meeting, the effect of the reevaluation on the high energy pipe break analysis was investigated. As a result of this investigation, it was con-cluded that the original high energy pipe break analysis remains valid. A discussion of the high energy pipe break analy' sis is presented in Attachment 6.

Very truly yours,

/Y

/ 'Q I u. /

/

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Paul J. Early Assistant Chief Engineer-Projects PJE:rz 271 079

s Attuenment 1

!J j...s 34 W N, Shh ghNr 1-1 g dk IDENrIFICATION OF SYSTEMS AFFECTED AND PR10dITIES 1.12 T_he reevaluation erf ort included those piping systen.s originally 1.13 computer analyzed with the Sh0CK2 code and which pertom or 1.14 affect sarety lunctions. r The >6 original problems are icentitied 1.15 in Appendix A were reevaluated using LiiuCK3. The SHOCX3 problers 1.16 are listed by piping system below, with the corresponding FSIdt figure (flow diagram) noted. Seventeen or tie SHOCK 3 problems 1.17 were subsequently recond>ined into 6 abranch groulens " and 1.18 reevaluated . These problems are identitied in the t.ible below as 1.13 B1, B2, etc. P_ortions of non-sarety piping syste:x, were included 1.20 in this effort wnere it was determined that r,uen lines would 1.21 attect the analysis or . saf ety related system. _Zxun;ples include 1.22 nonsafety interconnecting lines gast the tirs t automatic trip 1.23 valve or the first normally closed nunual valve anc signiricant 1.z4 branch connections.

For the purpose of expediting startup at JAFNPP, t_he original 90 1.20 computer problems were divided into 5a priority classes ror 1.27 reevaluation. P_riority 1 problems were those located in areas 1.2d inacces sible during plant opera tion; utese are areas where 1.z9 inspection and maintenance cannot be performed due to a nigh .JJ radiation environn.ent. Priority 2 are taua e in 1.31 2 I2roolems1 080 li-12966.41-131 0o/ 28/7:3 13 '

Attachment 1 1-2 accessible areas; these systems were sundivided into priority subclasses A, B and C n_ased on safety considerations. System 1.33 accessibility and Priority 2 classirication are also noted nelow.

Located in 1.36 Problem lnaccessible FSAR 1.37 System No. Areas Fig. No. 1.38 Standby Gas 942 W No (C) 5.3-2 1.40 Treatment w

1.41 941 G.*4:1 No (C) 5.3-2 1.43 Control Rod 909 Yes 3.5-5 1.45 Crive 1.46 Residucl 637 m No (A) 4.6-1 1.60 Heat 1.61 hemoval 1. ~>

271 081 641 (B4) .No (A) 4.8-1 1.64 643 No (A) 4.8-1 1.bb 1i-12966.41-131 06/23/79 131

i Attachment 1 1-3 Located in Problera Inaccessible FSlut System No. Areas Fia. No.

64o (Bo) No (A) 4.3-1 1.68 647 No (A) 4.8-1 1.71 650 (33, U 6) Yes 4 . 8 -1 1.75 657 Yes 4.8-1 1.78 s

664 (B2) Yes 4.8 -1 1.80 682 (52) Yes 4.8-1 1.82

$ 737 (B3,B6) Yes 4.8-1 1.84 m .

738 (32) Yes 4.8-1 1.8o 739 Yes 4.8-1 1.88 757 (63,E6) Yes 4.s-1 1.90 271 082 li-12966.41-1j1 v6/28/79 131

fjf Attachlaent 1

'~"

,',$1ft, Located in N.

Problera 2.naccessiule ti;/Jt Sys tem No. Areas Fig. No.

948 No (A) 4.8-1 1.92 951 No (A) 4. 8 -1 1.94 888 No (A) 4.0-1 1.96 867 No (A) 4 . 8 -1 1.98 ti70 No (A) a . d-1 2.1 871 (B6) No (A) 4.8-1 2.3 877 No (A) 4.e-1 2.5 879 No (A) 4.d-1 2.7 880 (B4) No (A) 4.8-1 2.9 bo4 No (A) 4.d-1 2.11 271 083 li-12966.41-1 31 C6/23//9 131

s Attactuaent 1 1-5 Located in Proble.y Inaccessicle FSAR System No. 4reus Fi(T . No.

866 No (A) 4.8-1 2.13 863 No (. a 4 .8 -1 2.15 d69 No (A) 4.8-1 2.17 878 (B6) No (A) 4 .8 -1 2.19 W

S tandby 931 No (C) 3.9-1 _

2.22 w,.

Liquid ' mit3 2.23 awam Cbntro1

(

2.24 Reactor 006 Yes 4.9-1

~

2.27 s- -

Wa ter 2.28 Clean up L"O mM 2.23 Reactor Core 656 Yes 4.7-1 2.31 p^,^-

Isolation 2.32 Cooling 271 084 2.33 li- 12 90 6. 41 - 1 j l 06/28/79 131

s Attaenn.ent 1 1-6 Loca ted in Problem Inaccessible FSidt System No. Areas Fio . :10.

o67 Yes 4.7-1 2.35 742 No (b) 4.7-1 2.37 r- --- E 933 No (B) 4.7-1 2.39 b

Core Spray 651 Yes 7.4 o 2.41 m- m, 669 (32,33) No (n) 7.4-o 2.43 673 No (u) 7.4-o F"*e 7 2.45 674 No (E) 7.4-o C@ 2.47 m

934 No (6) 7.4 o 2.49 Reactor 873 No (b) 9.7-1 2.51 Euilding 2.52 coouns 271 085 m. ,

Water *. +

li-12966.41-131 06/28/79 131

Attacuaent 1 1-7 Located in Problem Inaccessible FSidt S ys tem No. Areas _Fb.No.

872 No (B) 9.7-1 2.50 Fuel Pool 949 No (C) 9.4-1 gy 2.58 As cooling & N 2.59 Cleanup 950 F

No (C) 9.4-1 -7 2.60 T

N4 952 No (C) s.4-1 2.62 953 No (C) 9. 4 -1 2.64 947 No (C) 9.4-1

@ 2.66 d5L, Ifigh Presure 655 Yes 7 . 4 -2 2.u9 Coolant 271 086 2.70 Injection 2.71 679 (B 1) Yes 7 . 4 -2 2.72 681 No (B) 7.4-2 2.75 6d4 No (5) 7.4-2 2.77 li-12966.41-1 31 0o/28/79 131

4 Attacament 1 1-0 Located in Problem inaccessible FSta S ystem _No . Areas Fla. No.

693 No (ii) 7.4-2 2.79 668 Yes 7 . 4 -2

%g 2.81 Drywell N

912 No (C) 5.2-9 ( 2.83 Inerting,  % 2.84 CA.D & Purge  % 2.85 893  ;?o (C) 5 . 2 -9 2.8o 894 No (C) 5. 2 -9 h 2.88 894X No (C) 2.90 733 No (C) 5.2-9 2.92 740 No (C) 5.2-9 2.94 271 087 Main Steam 2.97 574 Yes 4.11-1 2.98 li-12 96 6.41 - 1 j 1 06/23/19 131

Attacnn.ent 1 1-9 Located in Proulem Inaccessible FSAR System No. Areas Fio. No.

575 Yes 4.11-1 3.1 631 Yes 4.11-1 3.3 891 Yes 4.11-1 3.5 M

716

~'

Yes 4.11-1 3.7 9,3,P fm W

714 Yes 4.11-1 3.9 715 Yes 4.11-1 3.11 717 Yes 4.11-1 '"M 3.13 718 Yes 4.11-1 @ 3.15 m

719 Yes 4.11-1 3.17 271 088 720 Yes 4.11-1 3.19 li- 129 6 6. 41 - 1 j l 06/28/79 131

Attachment 1 1-10 Located in Problem Inaccessible FSAR System No. Areas Fiu. No.

721 Yes 4.11-1 3.21 722 Yes 4.11-1 3.23 723 Yes 4.11-1 3.25 F

ud 724 Yes 4.11-1 N

- 3.27 C=y zcz wu 725 Yes 4.11-1 3.29 uma Feedwater 578 (B 1) Yes 10.8-2 3.32 Service 663 No (A) 9.7-1

&m 3.36 Wdter 3.37 sa, 865 No (A) 9.7-1 3.40 874 (BS) No (A) >.7-1 3.42 271 089 902 (b5) No (A) >.7-1 3.44 li- 12 9 t2 6. 41 -131 Ob/2e//t> 131

Attachmen t 1 1-11 Located in Problem inaccessible FSAft S ystem No. areas Piu. No.

901 (BS) No (A) 9.7-1 3.46 900 No (A) 9.7-1 3.48 881 No (A) 9.7-1 3.50 8

case=

wp 875 No (A) 9.7-1 [h_

, a 3.53 876 No (A) 9.7-1 r '

3.57 m

Chilled 960 No (C) Na 3.o1 Water m 3.b2 (Admin- 3.63 istration 3.o4 Building) 3.65 958 Nc (C) NA 3.08 959 No (C) NA 3.70 271 090 li-12 'J 6 6. 41 - 1 j 1 06/t3/79 131

Attacnment 1 1-12 Locatea in Problem Inaccessible FSAR System No. Areas Fiq . Ilo .

961 No (C)  ;IA 3.7 Fire 3.74 Protection 3.75 Mb 916 No (C) 9.8-2 g 3.77 fDE mi 917 No (C) 9 . 8 -2 ( 3.79 rm 918

'N :::3 d

No (C) 9.8-2 um 3.81 919 No (C) 9.8-2 f.741:::3 3.83

'ATt::::3 920 No (C) 9.8-2 3.85 gSt.

Cuabustion 945 No (ii) 12.3-22 3 . d '/

Air S 3.a3 Exhaust 3.39 (Emergency 3.90 Diesel 3.91

}[l li- 12 96 6. 41 - 1 j 1 0o/28/79 131

Attacnrcent 1 1-13 Located in Proolem Inaccessiale FCAR System Iio . Areas Fiu. No.

Genera tor) 3.92

g 4, a 271 092 li-12966.41-1 31 06/28/79 131

JAVIS A. FITZPATRICK tiUCLEAR IOWER Pl.UT 1.5 IsTTACineI4T 2 1.7 SUMS 1 sky OF COMBINED LP4E STRESSES 1.9 All problems wre reevaluated using the Sh0CK3 Code. 1.12 Stresses shown are Oueratinu basis Eartnaua ke (Oac.) Streuses 1.14 Design basis Earthquake (L M) Stresses 1.15 STATUS AS OF 6-27-79 1.17 Reevaluation 1.20 System and Allowanle Maximum 1.21 Proclem No. Stress (PSI) Stress Concen t_s 1.22 Standby 1.25 Gas 1.2o Treatment 1.27 942 19000 11805 1.29 27000 11978 1.30 941 18000 2063 1.34 27000 2684 1.35 Control 1.39 Rod Drive 1.40 9 09 17220 8258 1.42 258J0 10483 1.43 Residual 1.o1 Heat 1.o2 Rmoval 1.63 6 37 18000 8039 1.65 27000 7448 1.6o 641 (34) 18000 12348 1.70 27000 1134o 1.71 643 18000 12762 97 no7 1.74 27000 18924 L/jI U7J 1.75 li-12 90 6. 41 - 1m1 06/ta/79 131

JAK5S A. FITZPATRICK HUCLEAR POWEh PIA:JT ATTACliMENT 2 SUV. MARY OF COMBINED LI:1E S'IR EliSE S heevaluation System und Allowable Maximum Problem No. Stress (PSI) S tr ess Comments 646 (B6) 18000 15442 1.77 27000 139o1 1.78 647 18000 15544 1.u0 27000 14717 1.81 650 (B3, bo) 18000 12309 1.u4 27000 7458 1.85 657 18000 15664 1.u8 27000 18192 1.89 664 (b2) 18000 10231 1.92 27000 11520 1.93 682 (B2) 18000 10734 1.90 27000 91o9 1.97 737 (Bo) 18000 12455 1.99 27000 15081 2.1 738 (B2) 18000 11347 2.4 27000 13384 2.5 739 18000 91o9 2.9 27000 939o 2.10 757 (B3, E6) 18000 7623 2.13 27U00 5971 2.14 948 18000 ~/ 9 3 2 2.10 27000 3137 271 094 2.17 l i- 12 90 0.41 - 1m1 Go/ a,/ / > 131

JAME,S A. FITZPATRICK tiUCLEAR PLWER PUJIT ATTACliME 4T 2

SUMMARY

OF COMEIIJED Li!IE STMSSES Reevaluation System and Allowanle Maxin:u;a Problem I;o. Stress (PSI) Stress Comn.ents 951 18000 9107 2.19 27000 10424 2.20 888 18000 17o81 2.23 27000 16022 2.24 8 67 18000 4153 2.27

_ 27000 4057 2.28 8 70 18000 4755 2.30 27000 Sob 3 2.31 871 1E000 4416 2.34 27000 4338 2.35 8 77 18000 11713 2.39 27000 11716 2.40 8 79 18000 9813 2.44 27000 8634 2.45 880 (34) 18000 12348 2.48 27000 1134o 2.49 Bo4 16500 12053 2.52 24750 11393 2.53 86o 18000 16123 2.50 27000 13599 2.57 8 bb and 669 15000 4o23 2.59 27000 4ab9 2.60 271 095 li- 12 96 6. 41 -1m1 0o/28/79 131

JAMES A. FITZPATRICK NUCLEAR PLWLK PiitNT ATTACHMENT 2 SUyvid<Y OF COMBI:IED LI:4E STRESSES heevaluation System and Allowable Maximum Preblem No. Stress (PSI) Str es s Comments 878 18000 5137 2.63 27000 4890 2.64 S tandby 2.68 Liquid 2.69 Control 2.70 9 31 21840 9273 2.72 32700 9207 2.73 Reactor 2.77 Water 2.78 Cleanuo 2.79 666 17390 14466__ 2.81 26000 17003 2.82 Reactor 2.86 Core 2.87 1sola tion 2.68 Coolina 2.89 6So 13000 8635 2.91 27000 8531 2.92 6 67 18000 3949 2.95 27000 1171o z.96 742 18000 03o2 2.99 27000 b133 3.1 933 18000 15953 3.4 27000 14885 3.5 g

11- 129 o 6. 41 -lal 0b/28/79 131

JA'ES A. FITZPATRICK uUC1Eldt PUWER P1/d/r ATTACliv1NT 2

SUMMARY

OF COMEINED LINE STR ESSES heevaluation Syste:a and Allowable Maxi::.uia Problem No. Stress (PSI) Stres.s Conments Core 3.8 Suray 3 ,3 651 18000 15121 3.11 27000 22294 3.12 669 (B2, B3) *8000

, 6703 3.16

_27000 14o37 3.17 673 18000 _ 3735 3.20 27000 3761 ~ 3.21 674 18000 16063 3.24 27000 1457o 3.25 934 18000 9030 3.29 27000 9087 3.30 Reactor 3.34 Buildin9 3.35 Cooling 3.36 Water 3.37 8 73 18000 768o 3.33 27000 10708 3.40 872 18000 8030 3.44 27000 11121 3.45 271 097 l i- 12 96 6. 41 - 1:a1 06/28/79 131

JAM]:.S A. FITZPATRICK NUCLLu PUdus PMNT ATTACHMENT 2 SU.v.M14Y OF CO.vblNED LINE STRESSES Reevaluation System and Allowable Maximum Roblem No. Stress (PSI) Stress Comn.ents Fu al Pool 3.49 Cooling 3.50 S Cleanuu 3.51 949 21960 10425 3.53 32940 104d8 3.54 9 50 21180 19548 3.57 31770 203o2 3.58 952 18000 14557 3.60 27000 13340 3.61 953 18000 4057 3.64 27000 6291 3.65 947 18000/20850 4100/18982 3.66 27000/31275 3870/17379 3.69 Ifigh 3.72 Pressure 3.73 Coolant 3.74 Iniection 3.75 655 18000 9006 3.77 2~000 96 13 3.78 679 (B1) 18000 1o955 3.82 27000 loo 91 3.83 681 18000 _ 8506 3.80 27000 8229 3.87 271 098 li-12966.41-1: 1 06/28/79 131

JAMES A. FIT' PATRICK 2 iiUCIEAR PUWER Pl L4T .

ATTACliMENT 2 SUMMid1Y OF COMBINED IJNE STRESSES Eeevaluation

. rstem and Allowable Maxiluum

)_ abiera No. Stress (PSI) S tr ess Coran.ents 684 18000/2065d 8481/8202 3.90 27000/3097J U179/8502 3.91 693 22500 22485 '.94 33750 15245 2,95 668 18000 3993 3.98 27000 7913 3.99 Drywell 4.4 Inerting 4.5 CAD S 4 .6 Purae 4.7 9 12 18000 8233 4.9 27000 10d80 4.10 8 93 18000 10328 4.14 27000 10002 4.15 894 16000 8616 4.19 27000 22787 4.20 894x 18000 15035 4.22 27000 26304 4.23 733 18000 3526 4.26 27000 10977 4.27 740 18000 5498 4.29 27000 6599 4.30 271 099 li- 12 96 6. 41 - 1:a1 0o/28/79 131

s JAMES A. FITZPATRICK DUCLEM POWER Pl.'dir ATTACllMENT 2

SUMMARY

OF COMBINED LI:.E STR ESSES Reevaluation System and Allowaule Maximum Problem No. Stress (PSI) Stress Comr. ent s Main 4.33 Steam 4.34 574 18000 10096 4.36 27000 11078 4.27 575 18000 13026 4.41 27000 15o24 4.42 631 18000 12941 4.40 27000 15704 4.47 891 18000 15754 4.50 27000 19260 4.51 7 16 18000 7773 4.54 27000 "1340 4.55 7 14 18000 7086 4.59 27000 7120 4.60 7 15 18000 12445 4.64 27000 10589 4.65 7 17 18000 6892 4.69 27000 8414 4.70 7 16 18000 8153 271 100 4.,o 27000 7191 4.75 7 19 18000 5980 4.79 27000 5829 4.80 li-12 9 e b . 41 - 1m1 06/x3/19 131

JAMES JL. FITZPATRICK NUCIEJdt PUdER PLVIT ATTACliMENT 2 SUMM1J<Y OF COMBINED 1.Ich. STRESSFC heevaluation System and Allowuole Maxilauta Problem No. Stress (PSI) Stress Com:wnts 720 _18000 3706 4.84 27000 4/94 4.85 721 18000 _ 10931 4.89 27000 1427b 4.90 722 18000 11207 4.94 27000 9019 4.95 723 18000 11754 4.93 27000 11852 5.1 724 18000 2622 5.5 27000 2785 5.6 725 18000 3308 5.9 27000 3300 5.10 Feedwater 5.12 576 (31) 18000 16374 5.15 27000 20378 5.lo Service 5.20 Water 5.21 863 1o500 6156 5.23 24750 5254 5.24 805 18000 9171 5.28 27000 6359 5.29 874 (35) 18000 3;;o 271 101 5 32 27000 3,4a 5.33 li-129f,6.41-1m1 0o/ 0//9 131 ,

JAMES A. FITZPATh1CK IiUCIZAR PulEh PIJJJT ATTACEMENT 2 SUMM7.EY OF COMBINE.D LINE STKF.SSES Reevaluation System and Allowable Maximum Problem No. Stress (PSI) St_ress Con.rcen t s 902 (BS) 18000 3682 5.30 27000 3197 5.37 901 (BS) 18000 4122 5.41 27000 362o 5.42 900 18000 31e8 5.40 27000 23' 93 5.47 881 18000 17466 5.51 27000 2313S 5.52 875 18000 10376 5.56 27000 10417 5.57 876 18000 5345 5.61 27000 4631 5.62 Chilled 5.6o Water 5.o7 (Administratica 5.08 B ld a. ) 5.69 960 18000 2017 5.71 27000 2264 5.72 9 58 18000 6913 5.75 27000 035o 5.70 959 18000 986 5.79 27000 1J50 5.80 901 16000 1207 5.84 27000 1411 5.85

.s 11- 12 96 6. 4 i- 1u 1 06/20/79 131

JAMES A. FITZPAThlCK tiUCLuut PCW1.A PLlen' ATTACH'ENT 2 SUMLEY OF COMBI;iED LINE S'ITiESSES hoevaluation System and Allowable Maximum Problem No. Stress ( PSI) Stress Comnents Fire 5.89 Protection 5.90 9 16 18000 7527 5.92 27000 7335 5.93 9 17 18000 3060 5.96 27000 3118 5.97 9 18 18000 3o61 6.1 27000 37 11 6.2 9 19 180C0 n659 6.6 27000 o414 6.7 920 18000 3n27 6.11 27000 3d23 6.12 Caabustion b.1o Air S Exhaust b.17 Energency 6.13 Diesel Gen. c.19 945 18000 5212 c.21 27000 4520 c.22 271 103 li-12966.41-1m1 06/28 7/9 131

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ATTACllMENT 5 JAMES A. FITZPATRICK NUCLEAR POWER PLANT STRESS RE-EVALUATION CONSERVATIONS; CLARIFICATIONS AND ADDITIONS The following provides additions to the conservatisms noted in the letter of June 8, 1979 (JPN-79-32) from Mr. P. J. Early of the Power Authority of the State of New York to Mr. T. A. Ippolito of the Nuclear Regulatory Commission's Division of Operating Reactors:

1. Combination of Stresses The stress values for each loading condition are combined in accordance with the intent of ANS1 B31.1. 0-1967 to determine the acceptability of piping stress. Since the procedure used stresses from individual loading cases, and does not combine moments as per ASME Code III, the calculated total stresses are more conservative than the stresses calculated with the present ASME Code III. (Clarifies and supercedes Conservatism No. 1, first paragraph, of JPN-79-32).

2. Constraint Load Combinations The evaluation of constraint design considered the coincident application of deadload, temperature and occasional loads, in-cluding earthquake, even though their occurrence may not be simultaneous. Consideration of realistic application of these loads would result in reduced loadings. (Clarifies and supercedes Conservatism No. 1, second paragraph, JPN-79-32).

3. Floor Response Spectra Seismic response for these conservatively postulated earthquakes is represented by families of ground response spectra which envelope the effects of ground motion upon a suitable range of damped single-degree-of freedom (SDF) oscillator systems. To determine structural response to seismic loadings a mathematical model is developed which closely approximates the real structural containment system in physical and response characteristics.

Amplified response spactra are "onerated from the structural response for specific floors (elevations) in the structure. This is accomplished by using a damped sinusoidal support motion (i.e. modal response of structure to ground-shock-spectral) to determine the response of a range of damped SDF oscillators.

The procedure is carried o.ut for all significant modes of response of the structural system. (Supercedes Conservatism No. 3, JPN-79-32). -

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4. The analyses and reanalyses of seismic piping systems are based upon the conservative stress limit of 1.8Sh under DDE loading conditions. The corresponding ASME Section III Code piping stress limit is 2.4Sh under the DBE conditions. In July 1978, the NUREG/CR-0261 report, using the limit moment theory to address the Code rules, established that gross plastic deformation may occur when primary stress exceeds 1.5 to 2.0 times the yield

e strength (Sy) of piping material which corresponds to 2.4 to 3.2 Sh or higher.

5. Computed pipe stresses are magnified by the application of At1SE B31.1.

0-1967 intensification factors, which we believe were intended to represent fatigue factors and thus are not strictly applicable to the seismic load conditions.

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ATTACllMENT 6 IIIGli ENERGY PIPE 13 RE AKS The analysis of high energy pipe breaks, inside and outside the primary containment, are discussed, respectively, in FSAR Apppendix E and the Special Report on the Effects of a Iligh Energy Piping System Break Outside the Primary Containment (Supplement 25 to the PSAR).

The original analysis inside the primary containment considered, at least, the ten highest stress points on each of the main, steam, medwa te r , core spray, ilPCI and RCIC lines. As these points encompass the high stress points determined during the reevr.luation cffort, the pipe break analysis p~esented in Appendix E remains valid.

As described in FSAR, all safety related, high energy lines outside the primary co) :ainment were originally considered and analyzed for pipe break as ASMC Code III systems. For each line, terminal points plus two intermediate points were analyzed for pipe breaks. The terminal points are fixed. The reevaluation stresses at the intermediate points were reviewed and none were found io exceed the 0.8 (Sh+Sa) criteria. Therefore, the criteria used originally for the selection of postulated break points, based on possible impacts, and the points selected, remain valid.

In addition, all high energy piping penetrations of the primary containment have been reevaluated and found to bo acceptable.

Therefore the original high energy pipe break analysis remains valid, and such validity has not been affected by the current pipe stress reevaluation.

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