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Evaluation of Electrical Conduit Connections Between Auxiliary & Control Bldgs for Arizona Public Svc Palo Verde Nuclear Generating Station
	ML20090F502
Person / Time
Site:	Palo Verde
Issue date:	03/31/1983
From:	Bausch H, Pazargadi S, Scott Wilson; WYLE LABORATORIES
To:	;
Shared Package
ML20090F496	List: ML20090F493; ML20090F502;
References
	26406-1, NUDOCS 8307060025
	Download: ML20090F502 (154)
	v • d • e

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W i' N'PO,, NO-SCIENTIFIC SEtvlCES & SYSTEMS GROUP Y WESTERN OPERATIONS. NORCO FACILITY { Wyle Job No. AM-26406 1841 HILL 5IDE AVENUE, NORCO, CAUFORNIA 9176o AREA CODE 714727407: rwx eio432. iso 4 rEtECO*v(7:4 7274e7: cy,,,,,,,g,o, no. 10407-13-EM-600 154 Total Pages tf.is Report Date: March 31, 1983 r EVAI UATION OF ELECTRICAL CONDUIT CONNECTIONS BETWEEN THE AUXILIARY AND CONTROL BUILDINGS FOR THE APS PALO VERDE NUCLEAR GENERATING STATION a kfN p% **. Q "\\ l w r. [ 2,. y i PREPARED BYr IM 8 D7 son +t gg% VERIFIED BY- $, eg

5. Paz Gadi f

APPROVED BYr d- ~ N* g g H. P. Bausch ON[ENGINEfR QUALITY ASSURANCE. P PR 'L. Housteau 8307060025 0 -'- I PDR ADOCK 0 00 PDR s

..-m f f ., M REPot? NO. ?nnL mmans seem*c senaces a syswass onov, N MRAT1oNs,NotCo FAOuTY PAGE eso y ~

SUMMARY

l This final report presents the results and conclusions derived from the test evaluation of the flexible fittings (in line between the auxiliary and control buildings at Palo Verde) and the subsequent loads determination produced by the imposed deflections from building motion in a seismic event. Once the flexible fi tings were characterized from obtained test data, the critical design case was identified, evaluated, and determined to be safe and in possession of all necessary structural integrity. Since the evaluated case was deemed the critical or worst case, all electrical connections between the auxiliary and control buildings at Palo Verde Nuclear Generating Station at all elevations are not endangered by the relative motion between these buildings in an SSE event. i 1

L -. w.-- - -.- - i~-~2 -. _ ~ ~. l t ' M.= =s see=wesea m sas e m one, aEPORT NO. 7AhnA 1 uman l westenw openAnows,woeco pacun 3 l TABLE OF CONTENTS 3 Page Number i j TABLE OF CONTENTS 3

1.0 INTRODUCTION

5 i 2.0 SCOPE OF WORK 5 t 3.0 ' DISCUSSION 6 4.0 ANALYSIS 7 i 4.1 Analytical Assumptions 7 4.2 Stress Analysis 11-l 4.3 Natural Frequencies of In-Plane Bending Modes 30 4.4 Natural Frequencies of Out-of-Plane (Torsional) Modes 34 4.5 Dynamic Stress Induced by Seismic / Inertial Factors 38 )

5.0 CONCLUSION

S 42

6.0 REFERENCES

43 7.0 ATTACHMENTS l l ATTACHMENT No. I Desif,n Configuration No.1 (Elevation) 44 ATTACHMENT NO. 2 Design Configuration No.1 (Plan) 45 ATTACHMENT NO. 3 Design Configuration No. 2 (Elevation) 46 ATTACHMENT NO. 4 Design Configuration No. 2 (Plan) 47 i ATTACHMENT NO. 5 Design Configuration No. 3 (Elevation) 48 ATTACHMENT NO. 6 Design Configuration No. 3 (Plan) 49 l l 9 I

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'M neroaf NO. "M 00A.! usou==s see=c spass a svinw WESTERN oPERAT1oNS, NoRCo FACUTY g ~ TABLE OF CONTENTS (continued) Page Nurnber 7.0 ATTACHMENTS (continued) ATTACHMENT NO. 7 Vendor Design Specifications for Flex-Fitting 50 ATTACHMENT NO. 8 Y vs F* Plot of 2-inch Fitting Test Data 51 fr vs F* Plot of 2-inch Fitting Test Data 52 Y ATTACHMENT NO. 9 ATTACHMENT NO.10 7 vs F* Plot of 3-inch Fitting Test Data 53 ATTACHMENT NO.11 $Y vs F* Plot of 3-inch Fitting Test Data 54 ATTACHMENT NO.12 $

vs F* Plot of 4-inch Fitting Test Data 55 ATTACHMENT NO.13

$Y vs F* Plot of 4-inch Fitting Test Data 56 ATTACHMENT NO.14 Auxiliary Building - SSE Horizontal Acceleration 57 Response Spectra -Elevation 120 feet ATTACHMENT NO.15 Auxiliary Building - SSE Horizontal Acceleration 58 Response Spectra - Elevation 140 feet ATTACHMENT NO.16 Auxiliary Building - SSE Horizontal Acceleration 59 Response Spectra - Elevation 156 feet ATTACHMENT NO.17 Contact Report with Flex-Fitting Manufacturer 60 ATTACHMENT NO.18 Test Evaluation of the OZ-Gedney DX 61 Type Flexible Fittings l I )

^ ..... --.. :- u '...; ----- :- - ~. - - - ~ ~ w. a, . - = i I 'M REPORT NO. - w sc=.n.c s-a=== " 5 WESTERN oPf4AfloNL NoGCQ PACluTY ,,g,,g

1.0 INTRODUCTION

This report was prepared for Bechtel Power Corporation under Agreement No. 10407-13-EM-600, Equipment Qualification. Conduit Flex Fittings were evaluated under seismic loading conditions to assess whether structural failure would occur due to relative displacements between adjacent walls of the Auxiliary and Control Buildings. 2.0 SCOPE OF WORK The reope of work for this assessment is as follows: 1. Identify any structural inadequacy of the conduit / flex fitting assemblies under seismic excitation combined with a relative displacement between the building walls. 2. Evaluate the six configurations (Attachments 1 through 6) at three elevations in acordance with engineering sketches provided. 3. Calculate the resonance frequency for each configuration. 4. Determine induced stresses and identify any overstress condition encountered. 5. Identify problem areas and recommend follow-up action. 6. Preparation of a test plan for characterizing the flexible fittings. 7. Conduct tests on flexible fittings in accordance with aforementioned test plan. 8. Evaluate the test data. 9. Preparation of test data documentation. I 1

,M aeroa? NO. 26t406-1 ~ WESTERN opftAT1o0C., NoeCo PACluTY pa e 6 3.0 DISCUSSION The analysis covered th ee possible configurations (Attachments I through 6), two possible orientations (plan and elevation), three elevations (120,140, and 160 feet), three conduit sizes (2, 3, and 4-inch nominal), and four permutations for the deflection direction (eight if the vertical direction were included). The analysis first determined the loads (and consequently the conduit stress) for the static case, then determined natural frequencies, and finally the stresses induced from seismic / inertial considerations (consequently the total maximum conduit stress). To avoid the necessity for conducting a detailed stress evaluation of all 432 load cases described above, a single worst case configuration was identified and analyzed to determine the total structural integrity of this enveloping case (all other 431 load cases are less severe in terms of conduit stress and forces on the flex-fitting). The configuration chosen appears in Attachmen't No. 2, and is hithert' o referred to as Confipration No.1. This configuration was selected since it is the stiffest, consequently will produce the highest loads (and stresses) for an imposed deflection (relative wall displacement). Hence, if it satisfies all stress allowables (and it does), it will envelope all other cases. Since this is the case, the 2, 3, and 4-inch diameter conduit configurations shown in Attachments I through 6 and installed at elevations 120,140, and 160 feet are considered to provide safe and adequate protection for the enclosed cables. For these cases, under the imposed deflections and seismic excitations provided 6y Bechtel, none of the conduits were found to be over-stressed or failure predicted for the flexible fittings. i o I - - + +.- 9- -, ,e w, ..r,- -.,r._,,---,,,- w ,%-c,,we,,--y,r%-m,.m,v, ,-e-.,,,,_.--.,..---we,w-+-,..

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WYLE W=,"""" LABORATORIES GROU* OU hb6 fun l3Mnhalfref0Er1tf k he }hM h.'e. $ hvelopd by soubkinj #e fudwenial frepeney ofHe i valoakt )>w(f;) =na % lu+ "ws (&) by applyiq Aderley's equthin - sg ,g4 q 1 s '3 $L . C 9-I 'S l r art 1 % jI M *1, 4 k w e h '7e gk g i I 5 3 l Jr V~ 9.rrlhq>FV nt / ~ g h M L A Or) (f/2/d) {#/atAC) Ok) i b i s o.88,1 i.sge-, ,.7ce go.s o vo us e T I.I31 3.I3c-+ I. 40E-S. 91.9 c;31 wl 4" 2.eoy y.69g-3 /, yyy_S s.> o,vi g.u. 5 .. ~ __ .-. 1

e...-4.+we r.. -m. D.... su 6 .M 4' a', m. i W Y.1o..LE PA1 s""""

H

~ l yts dir'y h % -foffowhy hbulgak - d h, { Sirs ~ ( (#9 My) Ch) j 1 a. 70

zyy B.7f f

1 '5 / 3. 9 31.C /3, o 4

90. 'f SS.1 19, {

!) 0 d H -p f.{NatuM -freqximes of 004of-hane (Torsid)bichs h -4 q Suise ne hae, h M piht, no resistame h forswral ( (pnhed supporb) we. shoofd mit(y this stane bra-gnhe 1% Urlattti oduitsupport unald probbbly %e all&tWs (5ther Miior rotabbtf so the sydem am be. 7 - A dni-t r u f, c.',. wowed as 444 F 111ll t r [ g y& g7 L g

1 w ..s ..e e ..s... t WYLE 1735!'""" m._,- 37 e TAe g,ssa( load T, hoe lo (umpelsw N s* ( s ca., g [' vg 1h+j,$0-S9)W+ NOM) g -vdlMl..+Ll+P'([-l) + ha (&p K) a e I r I and it prodes ike rvtsfibn 9 abov1 y d \\ Y Q ON where k +omanti shdtkngth N appreimfed as k a>kers B, = shsar modutos = l. tee 7 +/i? e a $$b f0fW MOMG C l Whta yrden a defledin 2F N d [ f b' - (L.t e)fu % whia 4 addel % #e j i a 4. h b g I byt. 4.,+ gg)(' ' 4,.,tpi Mg, a v gy a' eex wx wr l ~ t r e

-......._.... ~. M '-/ WY_LE '#ts:s"""" m.o. ,c.ov. 3 sua thf the fflou>iny Mal Idleefong are prdoael d' - 4 Q,y,7e) u + ~&(L.*Y$ eex , M ( L. tk)' g* g Ser eer Pro 4lcitr] %e lamped Maw [regverlay

of f Ja'e2r${ E

" Er di /g.

r

.L k g f h stilari hvelopcd nherebf f [ * 'Mf [ 92l,+f* $*(/1Tf~I I C + g j 2nd 9, e fr0$00119b, = {l.,+f)W$, a ;,au s e a sk av,etop.vnw 9:' N /A A; Whereby

{l-., f f)0

'I b i eEr I

..u .....m_.. L _ ;; e

' ~'

WYLE i=il'""" LASOAATOAlES GROU* e donSefuer1t(f f' % b, + b, = (L.tL)%&, + 0"*$(J 8Er /)Pdowy a.rl,a7 f,, = arr/f 5 A = i m 6//edAj, calculdk, faMafi$ -, y ci w s W L, K' J' S IA @M/93 (*NS) (H (N (u) (%) h) (*3) /,Sre4 1.74s4 St.!r 14.o 0.VM /5 I /6.3 //./ 3 9.I5fA l Vo&?- 15;o 175 g.>98 GI.9. pa.1 it.f 4 Ifl454 l 9tEO. jo.otr 95.A 6'6tB Pb 51.6 Icj,7 where & rewitan4 fondwentst frequeruy is cahulated l by tis Dunkedey equafda shou >n earloir 3-p 3+1alear by compra'm kw valxs utk tie i-plane -fregosuf valves %4 +he seds sfrew stress prodocod g I

_ 1 _... _ __ _. = _ =.. E

'#i WY_LE, _ff"m"'1"""

g by Odiof-f ht18 h10Elon 56 Oh$f\\e 6301G OGf r kh1h}/ifD5 f (acoatly' 1hs In'-plane uitlI begrealer) u Jhe ih-plane jaddy so ha wyantetprewstel evher-{wglet1I7 the tierttia' entababbn tb $5tt0nhoh Pr nejkday it hve. ff.cDynamic Stress thduced by Setbsaf.Tartia% dorp !soking 2+ ffe E W motuor, 2s the destfn ease (as 4siussed above), & follouing generaineddynauk stress #ponht a (ae, Q ) on be expresse{ j D E-w E4 M' h = M' 11%k rh r/c/c = c g .((l-tipi cP+ M (L,+A) + v0 Yh ,Y h ~ 4 I k rja J a l U

. _ =. _ i i WYLE ff=l'""" j LASOMATORIES GROUP f Olene-Q

l2fjef lobd(edcr aMonpTG,g { Qq u)here-Se<SSE ((or!wda( Mcderafhn pMffe

[ E e Conkel blg, on a Griole LO,E 5 dee 7 i R c = SSE: ffonkostN( aledenihon hou-ff.e A Au$ltiry Bldg, on a Sri

3. o.E 6pA

-for the &ndamta( fegaenss sfready defermkl f l 2nd r<pe2 fed befoa> fr completenen, 9ef[7] help I._ l genera 1c -Ife (ellcahhbuhtin - y .c.w ELEV 6a Se G~ 6tsf CH C4) Gl+) CRA) h L i .2-fl. I !AO c55

0. W l.6 l

/Y o 0.58 6p-f.T h /6 o 0.63 a85 /. 6 i 5 iTY 19 0 060 0.7T l.D i j t't o 666 0.7s /.O l l-r l l /co BN der /./ ) Y //. A /A0 d48 0.75-B.7 /'/O 054 65 4,7 0 l i i (6o 060 6Or 0.b \\; Y Jere 5, t Se are dekrathe.d fw Refb] ash (=11.dampij A)pe. flerf c9. " sees'ony on u>allh ucelech4hn shee he {W yhhg s dynwklly unsopa p 7 p , - +. -

WYLE !?70s"""" LA80AATORIES GROUP Qg (i.e ' tk xla41dely very. soft neoprene stesve isoths the ko halves 6f +he 4%-[btiin so % sysieut onder coriderehou s [ree 9 ce ed). S-3 v 'bY lfif 0

l. 9.

G&l dfn1W56 6'fe']j f ~ e p.3 = ? %,[ $ w h % > L9-T 4 h Xfc [s.l (mL, t M) #- + IU O "h 'rpt j b rjs < ~ due R = smale4 lok S w o f k N { c see$4 is anay teen ivote :,%4 Acte praedue k mhkd k preke +he brjet (mt aonsenthie 0;;) I

e 8

W4 WYLE S=,""""

v..c.. ~. q, ~ RE1= $ otAkidyty fAe 6-d ( Mi d d3MId 6M'5 M BA186$ y U %ch Tor 4 Q ,6. %j& g-wpt s wis ae de 4 P4[ p %-Jodng result, are ohrW i w-3 r bl Y b 2 \\ses ha wr [ (kil) hD5 h$l j-i- JL (M 17.& l, o M.O t e I 140 t j,a- /.o J2), o M 3o.6 /, f g( f b 90 m.& o.9

16. 6 110 IU t.o 19 4 160 GM Iul

%4 ( 4 )+c e.6 0.e 77 l ) 110 f,b 07 /t.7 o a e

e.,

i.o a.s l l ? l G .'. all elevshbas 2nd siges has y I db ) y ter t S _____.i )

g g g.. - es nVH p -1... o., i l S'. o 6 n clu sco. ns Flex-1GNfities v The +11ree shes Ge a,3 aal il

os-6edny es-pamion-Tyye J)X dovpli&9) can all eHkdand 4he loads are>1ed by relative polld)pwidn be-b>een +he /&hvy ud todrol busIdtjs (af l9D+ 'Ev51&noNS) at k /PS fak \\lerde h) dest leenerat} Statiin subyal&d k 4he ksijn 6$E',

GondunY The- }bree nomIn2l ptsef (S,514') 6%n all S$h-stml He stresses credal by &relaise}&G ,eth Jefween Se /bSry soul CdrolJulIday (a+ 90+ ' SEMnokJS ) 9 k 86 lalo Yerle Mew geneaHg sws $4kNsk kaju SSE. . ~--~?E, ,,-m---- ~..

... -... ~... - ~ ~. -* ^ ~~"~~~" O---- 26406-1 M. sc==c==ss a :m== = REPOtf NO. uma=== 43 westfaN oPERADoNs, NoACO FAQUTY

6.0 REFERENCES

1. Popov, Egor P., " Introduction to Mechanics of Solids", Prentice-Hall, Inc.,1968. 2. Roark, Raymond 3., " Formulas for Stress and Strain", McGraw-Hill, 5th edition. 3. Thomson, William T., " Theory of Vibration with Applications", Prentice-Hall, Inc., 2nd edition. 4. Blevins, Robert D., " Formulas for Natural Frequency and Mode Shape", Van Nostrand Reinhold Co.,1979. 5. Shigley, Joseph E., " Mechanical Engineering Design", McGraw Hill, 2nd edition. l 4 = +-e--I e I

Cewou!T l?Lbow.5 FCf. 2~s 14 ' ggot, g 3 " s. )95" 4'. 19 2' I l il t, CLASS 15 JUNC. TION 3 SOX (TYPACAL) l ~ g CLA% i& CcNDUIT %IHS ~EA~o u s z s.t 4 4n .zy-1 m o 1 T( P icol v' DEFLECTLON ~~ FITTING -TTPE "Ox" 9..,wmir uew i pgg gg(,,, .Y. 4'~ O ~ E L E VATION VIEkl I l 6 TYhChi. INSTALLATic$ AboVE EL.'s toog I40, 4140 r e-. - - -, -,, - -,,, _,,., _,,,,., _,,

.5 s y t@ it = al k i 4 - ZJ - k bt a ^- 4DEFLEC1* ION q T FITTlNG* TYPE DX g S *N I s l} '9 T

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I N C). Ass IE JUNCTl6M 256X (TYPICAL) y I o u 0~* / CLA % i E CONDUIT y _g-SIZEC Z",3"$ 4" E_ } vi ii T - 2 A-PLAN 1/IEk/ TYPacAL #NSTALLATiod A6ouE ELEV.'S 900s [h's k 14N. H ~ m.

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= -I Q.17" a Dave J. 4 Q 35"sbae J. B. 'o Y j CLASS IB C4VDulT-%IEE$ (2", S'$ 4 N, 1 j I4 a, l Jord+t u 1;, ven.ecuou nnin t C3 is. Z n r ~ rvee.on p ____q q 2 2 ,1 > ool s ' <, " n,..,. - aks ",,a,,. M - Z A-5 r fe m I < 4 e"~+ ea.... c. nil.o - PL Ah - VIEh! TYPlcAL I95TALLATlos) be0UE ELE @'S .- o ,..r N ~j i

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Obh4 TEST EVALUATION OF THE OZ GEDNEY DX-TYPE FLEXIBLE FITTING NOTE: The test evaluation of samples 2, 3, and 4 inch OZ-Gedney DX-Type Flexible fittings was conducted at Palo Verde Nuclear Generating Station for APS by Mr. William Biehl (Bechtel, LAPD Office) and site personnel. The tests were performed in accordance with to the attached Test Procedure and conducted during January 24-26,1983. The testing devices was censtructed by Bechtel site personnel and was capable of applying tension, compression, shear offset, shear offset with simultaneous tension, and shear offset with simultaneous compression. The resultant deflections were measured by hand and all data was recorded (raw data sheets attached). The data was reduced and the trends plotted (also attached). r I \\ I

l as %-( (o 2 t ~ TABLE OF CONTENTS Page i Table of Contents 2 i Test Procedure 3 i S Test Sequence 30 2" Compression - Comments and Summary 32 i 3" Compression - Comme.1ts and Summary 33 4" Compression - Comments and Summary 34 j Data Sheets 35 1 4 I ii l h i i, t 1 1 1 r f = = _ -

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t "sG mhI'N-e4* heA*+$ M 'hw h-m*e eem e 6+ 4's-@9m s u o'- &Ob ~[ j b3 j = f a DEFLECTION / EXPANSION FITIING TEST PROCEDURE 4 e i l l I

... w L.. .:... w =.a.= ....w.._... -. c Mk[ j \\ 1.0 Purpose i Establish the physical characteristics of the 02/Gedney Deflection / Expansion Fitting Type "DX" used as seismic fitting for conduits between the Auxiliary Building and Control Building for ANPP'. 2.0 Test Equipment 2.1 Flex conduit tester-test apparatus to establish the physical characteristics of flex fitting. Apparatus exists at ANPP j ~ jobsite; however, minor modifications are required to the t apparatus - See Attachment A and drawings FGD-3337-E-1, 2, and 3. 2.2 OZ/Gedney Deflection / Expansion Fittings, Model Nos. DX-200, DX-300, DX-400 (9 fittings of each model shall be available for testing.) j 3.0 Special Instructions l 3.1 Bonding jumper lo:stion (see Attachment B for description) - orientation is identified with respect to its position on a i clock face when viewed from the flex fitting end. Position "A" i refers to the bonding jumper positioned at 3 o' clock or 9 o' clock and position "B" refers to the bonding jumper at the 6 o' clock or 12 o' clock position, when mounted in the Flexible Conduit Tester. I j 3.2 Displacement Measurements - shall be taken between a fixed point on the test apparatus and the displaced end of the test specimen. { Measurements shall be recorded to the nearest 0.01" (one-one-hundredth of an inch). l 3.3 Failures - any specimen that fails before completion of the i testing shall be replaced by another specimen, noting the failure mode. 4.0 Pretest Checks 4.1 Verify all necessary modifications are completed. 4.2 Verify test apparatus is complete and operable. 4.3 Trial test to be performed in accordance with Section 5. i 5.0 Test Procedure i j 5.1 Model DX-400 5.1.1 Remove all weights from both weight pans. 1 i j 5.1.2 Assemble the test apparatus such that the 4"$ Model DX-400 flex-fitting can be installed. 4 t t 4 !y I

, s. a. x ~ ...w x=.:< . &.: :.- -= e - ~..w.- .. :.... a ~ a.. + -. _.. . ~ n... ..n eMhN 5.1.3 Install the flex fitting in the apparatus with the bonding jumper in Position A. 5.1.4 Add weights to weight pa*.2 to " balance" the vertical . loading assembly to produce a zero load situation. 5.1.5 Apply loads to weight pans in accordance with Table 1. 5.1.6 After each load is applied, the test apparatus shall be lightly tapped such that any binding of sliding parts is overcome. 5.1.7 Measurements - Deflection measurements shall be recorded in accordance with the Data Sheets. 5.1.8 Test shall be repeated three (3) times using new specimens for each test. 5.2 Model DX-300 5.2.1 Remove all weights from both weight pans. 5.2.2 Assemble the test apparatus such that the 3"$ Model DX-300 flex-fitting can be installed. 5.2.3 Install the flex fitting in the apparatus with the bonding jumper in Position A. 5.2.4 Add weights to weight pans to " balance" the vertical loading J assembly to produce a zero load situation. 5.2.5 Apply loads to weight pans in accordance with Table 2. 5.2.6 After each load is applied, the test apparatus shall be lightly tapped such that any binding of sliding parts is overcome. 5.2.7 Measurements - Deflection measurements shall be recorded in accordance with the Data Sheets. 5.2.8 Test shall be repeated three (3) times using new specimens for each test. 5.3 Model DX-200 5.3.1 Remove all weights from both weight pans. 5.3.2 Assemble the test apparatus such that the 2"$ Model DX-200 flez-fitting can be installed. 5.3.3 Install the flex fitting in the apparatus with the bonding jumper in Position A. S~ I e

.c ._a -.,,. 1. 9M%-I p 5.3.4 Add weights to weight pans to " balance" the vertical loading assembly to produce a zero load situation._ 3.3.5 Apply loads to weight pans in accordance with Table 3. 5.3.6 After each load is applied, the test apparatus shall be lightly tapped such that any binding of sliding parts is overcome. 5.3.7 Measurements - Deflection measurements shall be recorded in accordance with the Data Sheets. 5.3.8 Test shall be repeated three (3) times using new specimens for each test. l b I

Nh$ 07 Attachment A ~ Test Apparatus Hodifications ~ 1. Disassemble entire apparatus, sandblast, paint and reassemble 2. Part No. S4, TS 3 x 3 - 1/4", Ground Smooth 3. All sliding parts shall be gress.ed with axle grease or bearing grease 4. Part No. S6, L4 x 4 x 3/8 shall be supported by MC 10 x 6.5, as shown in Figure 1 5. Part No. P3 shall be sodified as shown in Figure 2 6. Part No. T2, TS 4 x 4 x 3/8, shall be modified as shown in Figure 3 7. Part No. L10 shall be new rigid conduit, threaded both ends 8. Part No. Lil shall be reased both ends to provide chaefered inside edge J 4 l a l l i e 4 ] 3-

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P ,, 3 J 1 UNITED STATES OF AMERICA J2 c[. 2 NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD ' _j@; 3 4 In the Matter of ) 5 ARIZONA PUBLIC SERVICE COMPANY,) Docket Nos. STN 50-529 et al., ) STN 50-530 6 ) (Palo Verde Nuclear Generating ) 7 Station, Units 2 and 3) ) ) 8 9 JOINT APPLICANTS' SUPPLEMENTAL RESPONSE TO WEST VALLEY'S FIRST AND SECOND SETS OF INTERROGATORIES 10 11 Interrogatory No. 1(b) [First Set] 12 1. NUREG-0030, Vol. 6, No. 2, Nuclear Power Plants 13 Construction Status Report as of June 30, 1982, NRC. 14 Interrogatory No. 3 (b) [First Set) and 2 (b) [2'nd Set). 15 1. Letter from K. Martens to Wm. Boles, dated 16 November 3, 1982. Custodians: Marley and Bechtel. 17 Interrogatory No. 7(b) [First Set] and 4 (b) [2nd Set). 18 1. Internal correspondence (NUS) from T. Iaccarino 19 to M. Goldman, dated January 11, 1983. 20 Interrogatory No. 18 [First Set). 21 1. Letter from M. Septoff to E. E. Van Brunt, 22 dated December 14, 1982. 23 Interrogatory No. 20 (b) [First Set) and 7 (b) [2nd Set), 24 1. Work Scope / Work Plan for drift deposition assess-25 ment at PVNGS, Units 1, 2 and 3, No. 2976-09 (NUS). 26 830/060156 800701 PDR ADOCK 05000529 O PDR 33 mg ] c
J k Interrogatory No. 30 [First Set]. 1 1. Purchase Request and Purchase Order for Schmidt 2 M del 2-AXP-0 Sampler and related cartridges and filters dated 3 November 9, 1981 and March 3, 1982, respectively. 4 RESPECTFULLY SUBMITTED this I day of July, 1983. 5 SNELL & I R 6 7 8 By Arthur C. Gehr Warrer E. Platt 9 Char s A. Bischoff 10 Vaug n A. Crawford 3100 Valley Center Phoenix, Arizona 85073 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 -- ]
r-A UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of ) ) ARIZONA PUBLIC SERVICE ) Docket Nos. STN 50-529 COMPANY, et al. ) STN 50-530 ) (Palo Verde Nuclear ) Gr. aerating Station, ) Units 2 and 3) ) ) CERTIFICATE OF SERVICE I hereby certify that copies of " Joint Applicants' Supplemental Response to West Valley's First and Second Sets of Interrogatories" have been served upon the following listed persons by deposit in the United States mail, properly addressed and with postage prepaid, this 1st day of July, 1983. Docketing and Service Section U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Chairman, Maricopa County Board of Supervisors 111 South Third Avenue Phoenix, AZ 85004 Atomic Safety and Licensing Appeal Board Panel U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l Robert M. Lazo, Esq. Chairman, Atomic Safety and Licensing Board l U.S. Nuclear Regulatory Commission Washington, D.C. 20555
p-e o 6 1 .] Dr. Richard F. Cole Atomic Safety and Licensing Board U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dr. Dixon Callihan Union Carbide Corporation P.O. Box Y Oak Ridge, TN 37830 Lee Scott Dewey, Esq. Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Edwin J. Reis, Esq. Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Lynne Bernabei, Esq. Government Accountability Project Institute for Policy Studies 1901 Q Street, N.W. Washington, D.C. 20009 Kenneth Berlin, Esq. Suite 550 2550 M Street, N.W. - Was hington, fD. C. 20037 f I r i...
The Light Company ri,,,,si,,,, usiiw i,m im m im, ii.,,,si,,,. i_,mii gimusmii June 30, 1983 ST-HL-AE-968 File Number: G9.15 Mr. Thomas M. Novak Assistant Director of Licensing U. S. Nuclear Regulatory Comission Washington, D. C. 20555
~~Dear Mr. Novak:~~
South Texas Project Units 1 & 2 Docket Nos. STN 50-498, STN 50-499 Deletion of the Emergency Boration System HL&P has requested that NRC determine the acceptability of the deletion of the Emergency Boration System (EBS) at STP. NRC concurrence on EBS deletion hinges on acceptance of the main steam line break (MSLB) analysis which does not take credit for the EBS. This analysis was submitted as a part of FSAR Amendment 2 in October 1978. An NRC request for information concern-ing the MSLB analysis was answered via telephone, and resulted in agreement on what additional information should be included in the FSAR. The NRC staff then formalized their request for information in RSB Question 440.1. We have attached our response which we understand should be sufficient for the NRC to determine the acceptability of EBS deletion at STP. This response to RSB Question 440.1 will also be included in FSAR Amendment 33, scheduled for submittal in September 1983. The EBS is currently being deleted from the design documents. Should the EBS be required to be installed in the future, significant construction and schedule impacts will result. We, therefore, request an expeditious review of this material and confirmation of the acceptability of the STP position in a written response to this letter as soon as possible. If you should have any questions regarding this matter, please contact Mr. M. E. Powell at (713) 877-3281. Very truly yours, fi i J e Vice Pr sident Nuclear Engineering and Construction SSR/mg \\ 030/060170 830630 PDR ADOCK 05000498 l 1 PDR A
Ilouston Lighting & Power Company June 30, 1983 cc: G. W. Oprea, Jr. ST-HL-AE-968 J. H. Goldberg File Number: G9.15 J. G. Dewease Page 2 J. D. Parsons P. G. Barker M. R. Wisenburg R. A. Frazar J. W. Williams R. J. Maroni J. E. Geiger H. A. Walker S. M. Dew J. Te Collins (NRC) H. E. Schierling (NRC) W. M. Hill, Jr. (NRC) M. D. Schwarz (Baker & Botts) R. Gordon Gooch (Baker & Botts) J. R. Newman (Lowenstein,Newman,Reis,&Axelrad) STP RMS Director, Office of Inspection & Enforcement Nuclear Regulatory Commission Washington, D. C. 20555 G. W. Muench/R. L. Range Charles Bechh'efer, Esquire Central Power & Light Company Chairman, At,mic Safety & Licensing Board P. O. Sox 2121 U. S. Nuclear Regulatory Comission Corpus Christi, Texas 78403 Washington, D. C. 20555 H. L. Peterson/G. Pokorny Dr. James C. Lamb, III City of Austin 313 Woodhaven Road P. O. Box 1088 Chapel Hill, North Carolina 27514 Austin, Texas 78767 J. B. Poston/A. vonRosenberg Mr. Ernest E. Hill City Public Service Board Lawrence Livermore Laboratory P. O. Box 1771 University of California San Antonio, Texas 78296 P. O. Box 808, L-46 Livermore, California 94550 Brian E. Berwick, Esquire William S. Jordan, III Assistant Attorney General Harmon & Weiss for the State of Texas 1725 I Street, N. W. P. O. Box 12548 Suite 506 Capitol Station Washington, D. C. 20006 Austin, Texas 78711 Lanny Sinkin Citizens for Equitable Utilities, Inc. Citizens Concerned About Nuclear Power c/o Ms. Peggy Buchorn 5106 Casa Oro Route 1, Box 1684 San Antonio, Texas 78233 Brazoria, Texas 77422 Robert G. Perlis, Esquire. Hearing Attorney Office of the Executive Legal Director U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Revision Date 04-29-83
. _. ~. __ Li STP FSAR [ Question 440.01N . In response to our-previous question (211.85) regarding deletion of the emergency boration system (EBS) from the STP design, you have indicated that EBS deletion was justifiable, since, in the event of a main steam line break, t i the DNB design bases are met and the radiation releases are within the limits set forth in 10 CFR Part 100. We have reviewed the systeri aspects of the revised steam line break analysis in FSAR Section 15.1.5. Based on our F review we have determined that the following additional information is required. If this information has been included elsewhere in your FSAR, appropriate references in Section 15.1.5 will suffice. Likewise, if the i infonnation has been provided in the form of other documentation (e.g., Westinghouse topical reports), reference to such documentation (please be specific) is appropriate. i a. Clarification of the methodology for calculating reactivity feedback, including the effect of nonunifonn core inlet temperatures from the reactor coolant loops; justification of the conservatism in the methodology with regard to the peak power obtained. 4 l b. Clarification of the methodology used in calculating DNBR and i verification.that the power distributions used for DNBR calculations reflect the effect of nonunifonn core inlet temperatures from the i reactor coolant loops. t i c. With respect to ESF actuation functions for an SLB, describe and justify the differences between the-protection functions at the STP and the l actuation functions in NUREG-0452, " Standard Technical Specifications for Westinghouse PWR's". Describe the " excessive cooldown protection" function, which, in accordance with the FSAR, provides safety injection -in the event of an SLB. Identify the actuation set points. . Response a) See revised Section 15.1.5 b) See revised Section 15.1.5 c)- With respect to ESF actuation functions for. a SLB, the differences l between STP and NUREG-0452 are as follows: 1.~ The functions of (a) low compensated steam line pressure and (b) low-low compensated T-cold coincident with P-15 have been added. l -2.- The functions of (a) high steam flow coincident with low steam line j: . pressure-or low-low T,yg and (b) high steam line differential L v e ur %-f v g g-yy.y o-w e 7 g4 w-ev e rwy -t v yirr --- w--+ ai-'+w =d-W-mW- ,ast d *trsw a,- r'we-w wvte-*
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STP FSAR pressure have been deleted. This change was described to the NRC in December 1974 in Section 7.3 of RESAR-41. During CP review, the STP PSAR description of the ESF actuation functions referenced the RESAR-41 design and was subsequently approved by the NRC via the issuance of the STP SER, NUREG 75/075, August 1975. The " excessive cooldown protection" is comprised of the functions described in 1) above. A more detailed description follows. The excessive cooldown protection logic is shown on Figure 7.2-9. Actuation (of SI) can be caused by either two of three low-low compensated T-cold signals in any reactor coolant system loop coincident with P-15 or two of three low compensated steam line pressure signals in any one main steam line. P-15 is caused by two of four instruments indicating a neutron flux below 10% power or a reactor trip. The actuation set points will be established during Technical Specification review. i
SW NM on Figures 15.1-12 and 15.1-13 is more rapid than the case of steam release from all steam generators through one steam dump, relief, or safety valve. 2 The calculated transient is quite conservative with respect to cooldown, since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators. Since the transient occurs over a period of about five minutes, the neglected stored energy will have a significant effect in slowing the cooldown. 15.1.4.3 Radiological Consequences. The inadvertent opening of a single steam dump relief or safety valve can result in steam release from the secondary system. Normally, no activity release to plant personnel or the public is expected. However, if steam generator leakage exists coin-cident with the defected fuel conditions, some activity will be released. 2 The dose, being a function of steam release, will be less than that cal-culated for the radiological limiting Condition II event of Station Blackout. 15.1.4.4 Conclusions. The analysis shows that the criteria stated earlier in this section are satisfied. For an accidental depressurization 3 of the main steam system, the minimum DNBR remains well above the limiting valve and no system design limits are exceeded. 15.1.5 Spectrum of Steam System Piping Failure Inside and Outside Containment 15.1.5.1 Identification of Causes and Accident Description. The steam release arising from a rupture of a main steam line would result in an initial increase in steam flow which decreases during the accident as the steam pressure decreases. The increased energy removal from the RCS 2 causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity. If the most reactive RCCA is assumed stuck in its fully withdrawn position after reactor trip, there is an increased possibility that the core will become critical and return to power. The core is ultimately shut down by the boric acid delivered by 3 the Safety Injection System. The analysis of a main steam line rupture is performed to demonstrate that the following criterion is satisfied: Assuming e stuck RCCA, with or without offsite power, and assuming a 3 single failure in the SIS, the core remains in place and intact. Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable, the analysis, in fact, shows that no DNB occurs for any rupture assuming the most reactive assembly stuck in its fully withdrawn position. A major steam line rupture is classified as an ANS Condition IV event (see Section 15.0.1). 6 15.1-12 Amendment 3, 2/16/79
STP FSAR The major rupture of a steam line is the most limiting cooldown transient 'l and, thus, is analyzed at zero power with no decay heat. Decay heat would retard the cooldown thereby reducing the return to power. A detailed analysis of this transient with the most limiting break size, a double-ended rupture, is presented here. The following functions provide the necessary protection for a steam line rupture: 1. Safety injection actuation from either: l2 a. Two out of four low pressurizer pressure b. Excessive cooldown protection 2. The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the SI signal. l2 3. Redundant isolation of the main feedwater lines: austained high feedwater flow would cause additional cooldown. Therefore, in addi-tion to the normal control action which will close the main feedwater valves, an excessive cooldown protection signal will rapidly close all feedwater control valves and backup feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves. 4. Trip of the fast acting Main Steam Isolation Valves (MSIV) (designed k' to close in less than 5 seconds) from either: a. High Containment pressure !2 b. Safety injection actuation c. High negative steam pressure rate in any loop (below Permissive P-11) Fast-acting isolation valves are provided in each steam line that will fully close within 10 seconds of a large break in the steam line. For breaks downstream of the isolation valves, closure of all valves would completely terminate the blowdown. For any break, in any location, no more than one steam generator would experience an uncontrolled blowdown even if one of the isolation valves fails to close. A description of steam line isolation is included in Chapter 10. Design criteria and methods of protection of safety related equipment from the dynamic effects of postulated piping ruptures are provided in Section 3.6. ( 15.1-13 Amendment 2, 10/9/78
STP FSAR 15.1.5.2 Analysis of Effea.ts and Consequences. . Method of feelysis The analysis of the steam pipe rupture has been performed to determine: 1. The core heat flux and RCS temperature and pressure resulting from the cooldmm following the steam line break. The LOFTRAN Code j (Ref. 15.1-1) has been used. 2. The thermal and hydraulic behavior of the core following a steam line break. A detailed thermal and hydraulic digital-computer code, THINC, has been used to determine if DNB occurs for the core conditions computed in Item 1 above. The methodology employed is consistent with that used in the Steam Line 32 Rupture topical report (Ref 15.1.4). The following conditions were assumed to exist at the time of a main steam line break accident: 1. End-of-life shutdow. : r;;in at no-load, equilibrium xenon condi-tions, and the most reactive RCCA stuck in its fully withdrawn position: operation of the control rod banks during core burnup is restricted in such a way that addition of positive reactivity in a steam line break accident will not lead to a more adverse condition than the case analyzed. 2. A negative moderator coefficient corresponding to the end-of-life rodded core with the most reactive RCCA in the fully withdrawn position: the variation of the coefficient with temperature and pressure has been included. The k f versus temperature at 1000 g psi corresponding to the negative n,oderator temperature coefficient used is shown on Figure 15.1-11. The effect of power generation in the core on overall reactivity with the most reactive RCCA in the fully withdrawn position is shown on Figure 15.1-14 for nominal reactor coolant flow. The core properties associated with the sector nearest the affected steam generator and those associated with the remaining sector were conservatively combined to obtain average core properties for resetivity feedback calculations. Further, it was conservatively assumed that the core power distribution was uniform. These two conditions cause underprediction of the reactivity feedback in the high power region near the stuck rod. To verify the conservatism of this method, the reactivity, as well as the power distribution, was checked for the limiting conditions for the cases analyzed. This core analysis considered the Doppler reactivity from the high fuel temperature near the stuck RCCA, moderator feedback from the high water enthalpy near the stuck RCCA, power redistribution and nonuniform core inlet temperature effects. For cases in which steam generation occurs in the high flux regions of the core, the effect of void formation was also included. It was determined 15.1-14 Amendment 32
1 STP FSAR that the reactivity employed in the kinetics analysis was always larger then the reactivity calculated including the above local effects for the conditions. These results verify conservatism; i.e.. underprediction of negative reactivity feedback from power generation. 15.1-14a Amendment 32
l STP FSAR l 3. Minimum capability for injection of high concentration boric acid (2,500 ppm) solution corresponding to the most restrictive single failure in the SIS. The flow corresponds to that delivered by two f ERSI pumps, each delivering its full flow to separate cold legs. No credit has been taken for the low concentration borated water, which must be swept from the lines downstream of the refueling water storage tank isolstion valves prior to the delivery of high concentration boric acid to the reactor coolant loops. For the cases where offsite power is assumed, the sequence of events in the SIS is the following. After the generation of the safety injection signal (appropriate delays for instrumentation, logic, and signal transport included), the appropriate valves begin to operate and the HHS1 pump starts. In 12 seconds, the valves are assumed to be in their final position and the pump is assumed to be at full speed. The volume containing the low concentration borated water is swept into the core before the 2,500 ppm borated water reaches the core. This delay, described above, is inherently included in the modeling. In cases where offsite power is not available, a 10-second delay is assumed to start the standby Diesel generators and the additional time necessary to start safety injection equipment (mentioned above) is included. 4. Design value of the steam generator heat transfer coefficient including allowance for fouling factor. 5. Since the steam generators are provided with integral flow restric-tors with a 1.4 square foot throat area, any rupture with a break area greater than 1.4 square feet, regardless of location, would have the same effect on the nuclear steam supply system (NSSS) as the 1.4 square foot break. The following cases have been consi-dered in determining the core power and RCS transients: Complete severance of a pipe, with the plant initially at a. no-load conditions, full reactor coolant flow with offsite power available b. Case (a) with loss of offsite power simultaneous with the steam line break and initiation of the safety injection signal. l32 Loss of offsite power results in reactor coolant pump coastdown. 6. Power peaking f actors corresponding to one stuck RCCA and nonuniform core inlet coolant temperatures are determined at end of core life. The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod. The power peaking factors account for the effect of the local void in the region of the stuck control assembly during the return to power phase following the steam line break. This void in conjunction with the large negative moderator 15.1-15 Amendment 32
STP FSAR coefficient partially offsets the effect of the stuck assembly. The power peaking factors depend upon the core power, temperature, pressure, and flow, and, thus, are different for each case studied. The core parameters used for each of the two cases correspond to values determined from the respective transient analysis. Both the cases above assume initial hot shutdown conditions at time zero since this represents the most pessimistic initial condi-tion. Should the reactor be just critical or operating at power at the time of a steam line break, the reactor will be tripped by the normal overpower protection system when power level reaches a trip point. Following a trip at power the RCS contains more stored energy than at no-load, the average coolant temperatue is higher than at no-load and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed via the cooldown caused by the steam line break before the no load conditions of RCS temperature and shutdown margin assumed in the analyses are reached. After the additional stored energy has been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no load condition at time zero. 7. In computing the steam flow during a steam line break, the Moody Curve (Ref. 15.1-3) for fL/D = 0 is used. 8. Perfect moisture separation in the steam generator is assumed. Results The calculated sequence of events for all cases analyzed is shown in Table 15.1-1. The results presented are a conservative indication of the events which would occur assuming a steam line rupture since it is postulated that all of the conditions described above occur simultaneously. Core Power and Reactor Coolant System Transient I2 Figures 15.1-15 through 15.1-17 show the iC transient and core heat flux following a main steam line ruptuti (ct"olete severance of a pipe) at initial no-load condition (case '.) ofP re power is assumed avail- .able such that full reactor coolant f'.og dets. The transient shown assumes an uncontrolled steam release from only one steam generator. Should the core be critical at near zero power when the rupture occurs, the initiation of safety injection by low steam line pressure will l2 trip the reactor. Steam release from more than one steam generator will be prevented by automatic trip of the fast acting isolation valves in the steam lines by high Containment pressure signals or low pceam line pressure. Even with the failure of one valve, release is limi((d to no more than 10 seconds for the other steam generators while the one gen-erator blows down. The steam line stop valves are designed to be fully closed in less than 5 seconds from receipt of a closure signal. I 15 1-16 Amendment 2,10/9/78 ~ .n e.---,.,.
STP FSAR As shown in Figure 15.1-17, the core attains criticality with the RCCA's ( inserted (with the design shutdown assuming one stuck RCCA) shortly before kg boron solution at 2,500 ppm enters the RCS. A peak core power less than l2 the nominal full power value is attained. The calculation assumes the boric acid is mixed with and diluted by the water flowing in the RCS prior to entering the reactor core. The concen-l tration after mixing depends upon the relative flow rates in the RCS and l in the SIS. The variation of mass flow rate in the RCS due to water density changes is included in the calculation as is the variation of flow l rate in the SIS due to changes in the reactor coolant system pressure. 2 The SIS flow calculation includes the line losses in the system as well as the pump head curve. Figures 15.1-18 through 15.1-20 show the response of the salient para-meters for case b which corresponds to the case discussed above with addi-tional loss of offsite power at the time the safety injection signal is 3l18 generated. The safety injection delay time includes 10 seconds to start the standby Diesel generator and 12 seconds to start the HHSI pump and open the valves. Criticality is achieved later and the core power increase is slower than in the similar case with offsite power available. The ability of the emptying steam generator to extract heat from the RCS is reduced by the 2 decreased flow in the RCS. For the DNBR evaluation, a power and power shape 18 analysis consistent with the fluid conditions was used. The peak power remains well below the nominal full power value. It should be noted that following a steam line break only one steam gen-(. erator blows down completely. Thus, the remaining steam generators are still available for dissipation of decay heat after the initial transient is over. In the case of loss of offsite power, this heat is removed to the atmosphere via the steam line safety valves. Margin to Critical Heat Flux A DNB analysis was performed for both of these cases. It was found that the DNB design basis as stated in Section 4.4 was met for all cases. 15.1.5.3 Environmental Consequences. The postulated accidents involving release of steam from the secondary system do not result in a release of radioactivity unless there is leakage from the RCS to the secondary system in the steam generators (SG's). A conservative analysis of the potential offsite doses resulting from a steamline break outside Containment upstream of the main steam isolation valve (MSIV) is presented using the Technical Specification limit secondary coolant concentrations. A more realistic analysis is also presented. Parameters used in both analyses are listed in Table 15.1-2. The conservative assumptions and parameters used to calculate the activity 2 released and offsite doses for a steamline beak assuming no iodine spike are the following: 1. Prior to the accident, an equilibrium specific activity of radionu-( clides exists in the primary system. Reactor coolant concentrations remain constant following the accident (with the exception of acti-vity added due to fuel damage). 15.1-17 Amendment 18, 5/1/81
STP FSAR 2. Prior to the accident, the secondary coolant specific activity is equal to the Technical Specification limit of 0.10 Ci/gm dose equivalent I-131. This dose equivalent activity is presented in Table 15.A-5. 3. The fuel rod cladding is breeched in a number of fuel rods, which l2 results in the release to the reactor coolant of five percent of the total core gap inventory. This activity is assumed uniformly mixed in the primary coolant. 4. The primary-to-secondary leakage of I gpm (Technical Specification 12 limit) is assumed to continue for 8 hrs following the accident. It is assumed that 0.346 gpm leakage occurs in the defective SG and 0.218 gpm in each of the unaffected SG's. 5. Offsite power is lost; MS condensers are not available for stean dump. 6. Eight hours after the accident, the Residual Heat Removal System (RHRS) starts operation to cool down the plant. No further steam or activity is released to the environment. 7. The iodine partition factor in the SG's is the ratio of the amount of iodine per unit mass of steam to the amount of iodine per unit 12 mass of liquid and is equal to 0.1. 8. During the postulated accident, iodine carryover f rom the primary side in the three unaffected SG's is diluted in the incoming auxiliary feedwater (AFW). The steam releases and meteorological parameters are given in Table l2 15.1-2, with the realistic analysis parameters and assumptions listed j also. If the postulated accident is assumed to occur coincident with an exist-ing iodine spike (caused by a previous power transient), the assumptions l and parameters used to evaluate the activity releases and of fsite doses I are unchanged, with two exceptions. The primary coolant concentrations are assumed to be equal to the Technical Specification limit for full power operation following an iodine spike for periods of up to 48 hrs. These concentrations are presented in Table 15.A-4. Fuel failures due to the accident are not assumed to occur coincident with an iodine spike. If the postulated accident is assumed to result in an iodine spike (caused by the power transient of reactor trip), the assumptions and parameters used to evaluate the activity releases and the of fsite doses are again unchanged, with two exceptions. The primary coolant iodine concentrations are assumed to be functions of time. The spike is accounted for by increasing the source term or release rate f rom the fuel by a factor of 500. Further discussion of this iodine spiking is contained in Appendix 15.A.3. Fuel failures are not assumed to occur during the accident. The thyroid, gamma and beta doses for the steamline break for the various cases analyzed are given in Table 15.1-3 for the Exclusion Zone Boundary (EZB) of 1,430 meters and the Low Population Zone (LPZ) of 4,800 meters. 15.1-18 Amendment 2, 10/9/T8
STP FSAR ( 15.1.5.4 Conclusions. He analysis has shown that the criteria I \\- stated in Subsection 15.1.5.1 are satisfied. Although DNB and possible clad perforation following a steam pipe rupture are not necessarily unacceptable and not precluded by the criteria, the above analysis, in fact, shows that no IEB occurs for any rupture, assuming the most reac-tive RCCA stuck in l'tA fully withdrawn position. Re radiological consequences of this event are in the guide line of 10CFR100. ( '. l t 15.1-19 l
E 1 g REFERENCES SECTION 15.1: 15.1-1 Burnett, T. W. T., McIntyre. C. J., Buker, J. C. and Rose. l R. P., "LOFTRAN Code Description," WCAP-7907, June, 1972, 15.1-2 " Westinghouse Anticipated Transients Without Trip Analysis," WCAP-8330, August, 1974. 15.1.3 Moody, F. S., " Transactions of the ASME, Journal of Heat Tranefer," Figure 3, page 134, February, 1965. 15.1-4 " Reactor Core Response to Excessive Secondary Steam Releases," 32 WCAP-9226, January, 1978. \\ l l l l 15.1-20 Amendment 32 l -}}

ML20090F502

Text

SUMMARY

1.0 INTRODUCTION

5.0 CONCLUSION

6.0 REFERENCES

1.0 INTRODUCTION

6.0 REFERENCES

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