Text

Rev 0 to Auxiliary Feedwater Sys Design Review Rept. Util QA Statement Re Verification of Adequacy of Design & Const of Byron Unit 1 Encl
	ML20076D144
Person / Time
Site:	Byron, Braidwood, LaSalle, 05000000
Issue date:	07/16/1980
From:	Braeger C, Meister J, Wattelet P; SARGENT & LUNDY, INC.
To:	;
Shared Package
ML20076C924	List: ML20076C922; ML20076C992; ML20076D015; ML20076D046; ML20076D074; ML20076D111; ML20076D131; ML20076D135; ML20076D144; ML20076D148;
References
	DRR-AF-001-BB, DRR-AF-001-BB-R00, DRR-AF-1-BB, DRR-AF-1-BB-R, NUDOCS 8305200688
	Download: ML20076D144 (26)
	v • d • e

{{#Wiki_filter:r t _J _ ~ =

t t.

s..D i l m,.

Station: Byron /Braidwood Units 1 r, 2 Project No.: 4391-09 \\ Client: CoNnonwealth Edison Compr.ny Systc.m Reviewed: Auxiliary Feedwater Syr. tem Revision: 0 Dete: July 16, 1980 I have conducted a design review of the above-n'entioned systeu in accor'tance with the approved checklist.

O <J (I ;,df y* t-(l s

P. L. Wattelet i A PMD (Print Nana)- (Signr.ture) (Review Chaircen) s J. R. Meister g__ [ /[f d PMD (Print Name) / (Si nature) (Division) P C. B. Braecer ,,g CID (Print liame) (51 pat'zare) (Division) \\ N I. Wu .3 EMD i e (Print Ner:c) (Signr.tura) (Divicion) l P.s%< 'W... -- f'. _~.~,.', l C. S. Burton

s..

MDDD (Print Name) . -y. l (Signature) (Divisien) 8 g\\ l r D. A. Schroeder i h / EPED l (Print Narae) (Signature) (Divicion) I .. I

Comments as of yet unresolved have bcon listed on pages 2 3 3.

N Byron project must resolve these comments to complete the (q k!, accign revicw. .5 ~ N .o c I WA y O g l be i t a., C.s i < r, l .) d, m y i t. l J, c.: i

= v I

.} O$ 8305200688 830511 M E C1:A ';; C O. F.' P T " " : T - -' ~: oS PDR ADOCK 05000373 -- '" ~. l P PDR 3.gg. j:'4 DESIGN XEVIn: :.io '.T 1 4 - U :~~ ~ ' . ~."C.~ 7.. ~.~* i N ~;RR-AF - 0 7.1 - 5 3 FOR nrFICJ USE

3. t Y - Nt:1 TO Ofi f.

e ~~'".'.# SENT G U T .' D E Or S t. G ~ N T E L J';DV I ' ' ' ..su-

ap 5, e e.,,,

..._...r'.-* a .m.. .s-9*.-.m.....s., .,e .J.= .,,,g. i

r e l DESIGN REVIEW REPORT

SUMMARY

sdh The following unresolved comments must be addressed by the Byron project: 1.) .PMD Review Consideration ll, Comment 11: FSAR Article 10.4.9.3.1 states that the AP pumps have been designed to start on either loss of cffsite power, 1cw-low level in any steam generator or safety injection signal. Westinghouse functional Diagram 10SD685, Sheets 5 & 15 and Sargent & Lundy Diagran.s M-4037-1AF01 do not show the auxiliary feedwater pumps starting on loss.of offsite power. Byron must resolve this conflict between FSAR statement and design. 2.) EPED Review Considerat av 1, Comment 7; The design criteria does not specify IEEE standards for Electrical Class lE comconents as required by the Classification C.citeria CC-ME-01-3B. EPED Review Considerations 3 & 4, Comment 10: 7 Environmental conditions are not specified in F/L-275SC, Amendment 2. 4.) EPED Review Consideration 5, Comment 9: The required equipment qualification documentation must be submitted, reviewed and accepted. 5.) EPED Review Considerations 6 & 14, Coment 2: l Solenoid operated Valves 1AF002A & B, and 1AF024 i do not have control logi.cs, schematics, or cabling l identified in appropriate docu.ents. l 6.) EPED Review Consideration 6, Comment 3: [ C Solenoid valves lAF004A & B shown en Logic Diagram 'Y M-4037-1AF02 and Master Diagram ;$519 do not have E schematics, nor are they identified on the P&ID.

m. =e T

7.) EPED Review Consideration 9, Comment 4: - vs @ es Valves lAF005A-H are located adjacent to each other &J providing insufficient separdion, and the associated S iq Automatic Flow Controls are in tha sa"e area. I r.~ -...~-.- { {_ S-3') "p-'., S. p' r*

_. -3 FGR CF3 ICE USE GNLv - NOT TD Dc i r e n e..n. *. ;. m. c.<

e '--a--. .i .>....u. st:NT OUTSIDE OF 5/.n G E t;T (, LUNDY PD O 2 E '**"s.'. ~ t

c ./ Dl' SIGN R-EVID1 REPORT SLM!ARY: ~ h' 8.) EPED Review Consideration 12, C.7;aent 1:- The design criteria requires automatic start of the pumps on loss of off-site power logic and -~ the schematic diagrame do not implement this directly for Pump 1B4 9.) EPED Review donsideration 15, CO =ent 5: Overloads of Valves lAF013A-H, IAF006A&B, and 1AF017A&B are not bypassed during ESF action per NRC Regulatory Guide 1.106. 10.) EPED' Review Consideration 15, Cc r.ent'6: Transfer to local control panel does not transfer control of 1AF006AKE and 1Ar007A&3 for emergency suction to aux.iliary feedwater pumps. 11.) EPED Review Consider.ation 15, Cornent 8: Transfer to local control panel'does not preclude shorts on the control circuit in the main control room lights, thus rendering the circuit inoperc.tive for Auxiliary oil Pumps lAF0lPA-A and 1AF01PBwA. The Review Status of this system shal-1 be indicated as Coraclete with Unresolved Comments (CU) until Byron project resolves these comments. l l l 2 f. I ,s s g

m. cc e4 V I

en f bh< 4 o e,; ' = ta u_ _- _.=- I j (.-...-.-.. ( [. _....'; L..' I ' t L L.. o. h-l 00- A. F-FOR OFFICE USE ONLY - NOT TO DE,; .f..,-. j l l si

.L..

g j,.,'y pq,, 3 SENT OUTSIDE OF SARGENT & LUNDY 3

ELECTRICAL DESIGN C o'^'"r' ?' - e s REVIEW CONSIDERATIONS YES EO N/A ....,J.,.. b:.. . r.. 5 I. 1. Is the classification of the electrical Q, components of the system, as indicated in the project classification critoria, consistent with the safety. function of ..7 the system? X g. 2. Do'the electrical documen'ts (master diap, rams, schematic diagrams, key diagrargs, cable tabulations, cr.c.) for the systen reflect the correct safety classification? X i I 3. Do the equipment specifications (both mechanical and electrical) establish the i qualification requirements (environmental and scilmic) for the c1cetrical ceaponents in accordance with the project classification X 10 and design criteria? 4. Arc the. equipment qualification requirements. specified in the equipment specifications up-to-date? - X 10 g 5. Has tJie required equip.nent qualification documentation been submitted, revicued and accepted? X 9 I 6. Do the electrical documents (master diagrams, } l key diagrams, cable tabulations, etc.) reflect the latest informr. tion with respect to equip-l ment size, ratings, etc.? X 263 i 7. Are the redundant safety-related components b of the system identified? X h5 l T(', J 8. Arc the redundant safety-related electrical j g components of the system assigned to redundant t j-u-auxiliary power, control and instrument buses i consistent with the divisional assignments? X i .s o 9. Arc the redundant safety-related electrical I. 'o,. components of the system separated and' ~ g segregated or physically protected to l g minitize the probability of simultaneous ,p, failure of redund ntt components from the 6 t', effects of a missile, pipe whipping, I g g4 dischr.rging fluid or a fire? X 4 g % t to N< !W C H A N I C l.L D E P A r Tr.':...! '~,7'** fi EPT. NO. DRR-AF-001-BB D.VI F. 6.,2 7 / 80 oo O

s. r<

TITI.E:. A.u.xi.l_i_a.rv F.e_ed'es t et DES 1GN R:. VIE *J CE7CRLIST 2_ ~----- ----- I (h - k [- - ~ ' ~ - - " r on.os:F I '. E USC UNLY - NOT TO llE SENi UUT5IOC O '- 5 AINC NT I, L U'4 O '.' 23.0 C 1 OI b " ~ k

L

^~ -....-. s a. - -_._......-..~......e.. -~

-- ~.. _. - -...,.,.- _. I,,,,,._.~...-...- S

p...!.J..

UIIX3.;1C.\\1. Di. SIC 1 !'O !!//i...... '. ; i e RE\\'] D' C3;;C) rd"ill0:iS YES .t { - -. ~ = -~ O >lI V ~ '= the dess.' er 'ne etect ric rott ic" < t 1o-r the cycten such that. the safet y f tmetion of th: syst;cr.: c:.n be acec.:plish'ai-in the event i f of c sint.lc fallure? X II 11. /.r o th.: no:-r.nfety related O.ctrical i, cc:poacat.s of the systeh s.7arated anf ( segregated f rom the a:irel y relat ed l' i co:rpor.cnte in accocdnnce wit.h the pcoject l decisn crit.cria? X j i s 12. Do t.he syst eri sc1tenutic disgrans it:plev.ent i I tbc cc,ntrol fenctinas specified in the systcr. logie diagrans, functional det,triptionc I or syctcin doncriptions? X 1 ji .1 'i. Da t he re]r.ted electrical inst alla'icin { c'ravi.r.pa (conduit, cable troy, cabla ~ i i tchule' ion, et c.) rcilect the. ci t. trict.1 I design, e.r. cn53c types, sizes, regre;;aticr. l ccter.:,rics, sepnration reef:it n. at.s $ndicat e.; l on the nsecciated syst..- nact. c c'iagrc:.., ? j X l 14. Can tbc sciety cystcia he tected in accora.t.cc l vith the requirnwuts of IH.E Stendare! 330 cr !O e:..er br<.;<et c::41s:. criterie? x 2 15. De the ir.nt: e;.ntr.t. icv.- and cc. titre] f ca ttirc e. l, of the Eyste::. ec : ply uith the recs 'rcncats of Sectic.m:.,.1.1 (Chr..el typans or Ts Ec. oval fraa Operatien), 4.12 (Operating 3 Byt,assec) c.nd 4.13 (Inficati..c. of hypascon) { ", of IE'm 2/9-l'R1? X S&6&B

w. -

A b-.'.).l t ,4 i I. ..,o e v.3 [, _g

s.

M....- l ,8 t I s ..a l o, ') I t.o I e,

c. * ~.,t c r-t

- - to

. c-

.~ s c':>

= ~:g v I.O.: i a I l g i... n;...,..... ' T r. :' '. i n' P../ F. -001. B.B... pl.TE_.6/ 2 7/.60 _ _. ! 7 1' ' '- "~ 3 ,. :l !!i. Auxiliary Fecdwater DF'. lG:: P..".' I D t'I: EC K '.' !, (. l..,,, O 8'... . ::.;=..:.:.... .. -. -........ ~.. ..._.,l :. r ?!. L',.. l f ti: fil' F i s.f a!J fit;t Y - fi i' 1 Il HC l f'. t i ..'i,. =,g 1 f lx .i.'- .c ( g u. n v. ". I 8 ~* .e..,. ' /.R t.: NI L i1:'n,Y h,L.".-__..,.'.. "r l

irr:1 i;'.s i S I'E im

(.-.-..... l

7- ! C.O::iF.!!T " REVIEb'ER'S CO:01ENTS . ~... _. s txMBER NAME_ D. A._ Schroeder RESOLUTION .I 1. Design Criteric requires auto start of punps on loss of off site power logic and schematic diagrams do not~ implement this directly for Pump 1B.' - - s 2. Solenoid operatt.d valves lAF002A 6 B, and 1AF024 do'not have con-trol logics, schemtics,'s or cabling identified in apprcpriate documents. 3. Solenoidvalves1Fb04A&Bshot.n on Logic Diagran M-4037-1AF02 p l. Master, Diagram AF519 do not have schematics nor a e they identi-fled on the P6ID. 4 Valves IAF005A-H are located adja-cent to'each other providing insufficient separation, and asco-ciated Automatic Flou Controls are in the same area. ~ [5. Overloads of Valves 3A5'013A-H, 1AF005A 6 B, and 1AFOl7A 4 D are .a not bypassed during ESF ection for NRC Regualtory Gbide 1.106. Transfer to lo' cal control panel does not transfer control of s IAF006A 6 B and 1AF037A & B for 3 emergency section to auxiliary b., fecduater. pumps. e E ~ t[ 7. Desig' n Criteria does not specify ( IEEE standards for Electrical Class'lE components as required by the'Chssification Criteria CC-ME-01-33. $.n S. Transfer to local control panel S does not preclude shorts on the -Q5 y control cirucit in the main con- /7 trol' room lights, thus rendering 48 the circuit inoperative for Auxil-lary OilFusps lAF0lPA-A&lAF0lPB-A., f NO. ' ORR-AF-001-BB DATE 6/27/80 M E O f f N ! C A L _ O F p a R T.*ie N T S T.*.

b.*. ~. "

-a = TITLE Auxilicry Feedvater DES 1G'; :DlIE'.i CODENTS j f'on nrrICE USE CNLY - NOT TO BE if-~~~-~~~~~ !y e.f......, v z, ,,r...., 3 t SENis O U T '3 f C E OF SARGENT 6 LUNDY i c ..== ' ' '.., ,.,.n c m __g r _..- -._ J ._.-..J N e e = + s

.r0:0!E:iT REVIEWER'S C0;OIf.NTS NU.'t bel' NAl[E D. A. SchroedE RESOLUTIO:: 9. Environmental Qualffication not for F/L-2738, F/L-2891, F/L-2802, Seismic Qualification not accepted F/L-2819 or F/L-2802 10. Environmental conditions not specified in F/L-2758C and Amendment 2. .a 1 O L. O m s l 0 I f.2 { 1% 's I \\ !lh C I cf ' u i> 3 t I "h. P. I g l 4, O N ia \\ 5 l U *.,0. DRR-AF-001-BB DATE 6/27/80 - CT D t. R P ' F N T 5 T a * ';. ~.' ' gg MECHANAl 3 u r. \\ = - - i ,.rI.i.r - l DESIGN REVIE'.' CO).NFSTS n tn@.i;ff reidw. iter -. ~ - -..... -. ~.. - l I l FOR CIFICE USE ONLY - NOT TO BE . O Q...l}. 18 i{.'* ** * " ~O _ 'v."u r".::'g u l! y. e t.. t SENT UUT's DE OF SARGENT & LUNDY /,

7. f. A x v:

A, N.... - - _m. m - e e -.a-.a.:-- ..~*n""**

L' '. H G C NT E. L * )N DY r u c st.* c c n s .:m n.cca f s vJ2::. ;TF 57 * ^TL / c Tl TJ. Or *.n.!.,.. AT. :.2.-EB 1. C1csnf ~$ cc: ion Crf :c rf n CC--i:~ 01-E3, .e.ic en 5. u r... p...i 9....> e.....1 o. y,,, s, e _., e s.-. . u ' u on E 3. 1:-37 4x : c D 4. 1:-2037

c. '. 1 5.

1:-2337 "h rn t 2 -!.s'en C 6. ? -2037 h ut 3 73rfen 3 ,. io 0.1,.., e r. n.

m.. ac..

1 ~ 7. v. e 'e - :<icn C C. I 4C57.-1A701 t it. inn C 5. 1:-4057-3 /IO2 ,.......Da .,:.r,a c 10. .. n -- se '<;cn A

1. 1.

'. l. '..' - ' m' ~. t).' =' 12. ?I -403:.2 /JiD5 .c in'en B r. '<-fon G l'. 6 / 9.0E--l-3301 ten D 14. 6/205-1-E 92 15. 6/20E 3305 ~ ."? s:':en E 16. ( /20E-1. / C30 AF01

a $ :- :! on E

.r -

en,,.
17, e.f,..

v :. _- v.> u t 1 e ' 10. '/?OD 1~403? //.- 03 N fc3ca L i l + l ' 19. 5/0:-1-4033 'J.3 C 4 =micion A 20.

/' 0: 1 4030

/105 i>1cn A

'1.

6/20E-1 4030 'F06 'er.sica A 1 V'. 5/: 0; ' !MO - r07 Ta. * = b n A ~. :t < : A "3. 6/.;r ' 4D30 _0S e 4 - e.s A r '. A Q s 5 38 6 /.*) f.M... 9

q. = ' ^aU

'C' -w n vs .4. w s o .g -.h

)
J e

,/. .,( s. .i s ( \\./ Page 5 of..6.- A l l L ~ - - .m__

i -l,* CARGENT & LUNDY. l 0 EN GIN C E RS t CMtCAGO j 26. 6/20E-1-4030 AF11 Revision A 27. 6/20E-1-4030 AF12 Revision A f 28. 6/20E-1-4030 AF13 Revision A 1 29. A-229 Revision V l

30. AF-503 Revision 5

31. AF-507 Revision 7

32. AF-511 Revision 9

33. AF-516 Revision 9

34. AF-517 Revision 7

35. m -519 Revision 1 36.

AF-532 Revision 5 i 37. Cable Tabulation - AF - June 9, 1930 38. Cable Tabulation - SX - June 9,1980 ' 39. Cable Tabulation - VA - June 9, 1980 9 1 4 O e o O Page _5__ Of _.*_

~ --..,.~,.m.,-- PMD S Er. CO.5ENT REVIEW CONSIDERATION YES NO N/A gipqgt.g .i.. p 1. Have the Auxiliary I'ecdwater Pumps been provid-g j d ed with sufficient NPSH available? Where is this documen'ted? 2. Has the Auxiliary Feedwater System been sized ./'. 1 to provide einimum flo.i,to unfaulted steam generators, as specificd by Westinghouse, during a feedwater line break? Where is this documented? 3. Do the AW pumps have suf ficient head to deliver 3 '3' 'their rated capacity at the secom generator safety, valve set points? 4. Have considerations been gi~en to testing the ). /[, AW syste n periodically to insure operability? / 5. Have hig'h, enurgy portions of the AW system Ni f f, been analyzed for proper suppcrt and restraint? 6. Has adequate accessabf11ty been provided to f perform the in-service inspection expected to be required during the plant life? Where is this documented? 3 7. Is the AW system properly classified? Are the boundaries cIcacly defined? 'K /' Where is this documented? 8. Have the seismic analyses required for SSE and g y, OBE been performed? Where is this documented? .w N f 9. Has the A W system been designed to minimize K 6' $p j' if water hammer in the steam generators and [ pi ing? P k 10. Is the A W system designed to perform its [ /;t, y, intended safety function while custair.ing a 0 single active failure and Joss of effsite 3-power? * >0kJ; 11. Have the AW pumps been designed to start on ~ g jf, .? ? either loss of offsite. power, low-low level gfj in any steam generator or safety injection jf signal? (/~J 4 4 to. 4. 's. 3. t) I v tn 4 c. 1 I"I C H AI4 I C AL C EP AIIMEfil EIOAAI gg Kl'T. NO. DR R-n, r. -001-BE DATE 5-2h SQ _ ' Auxili, rv feedwatcc Syste::: u T!TI.E: a ~ DES 1CN PEVIEW CHECKLIS'- gy. w~-------_ .. nEv.l tsf.n i FOR O F F l'. E 1 f.E flNL Y - ?!U T TO (1E j g .....,.7 ' l. ' _" *

EASE I OI_

I' '8 SENT 0U15iDE OF ' S A m PJ T & L UllOY =euw*mn2.l l \\ - _. ~...__- m

PMD S t.r - - = : COMM:-::T 'REVIEb' CONSIDERATION YES NO N/A gpgg p h 32. Is the AIV system designed such that operator -( j', action is not required within the first 30 minutes following initiation? 13. lias ti.c AFW system been, designed for uso during y/ 4 '/ C, start up and shutdown? If yes, what considerations have been given to the systcu design to operate in this modo? O k .k [ j \\ p 4 $ Q n N l }- t. W tw d. e4 N dE ~ $~ l f fECHANICAL D E P A F' T M E N T SIA CA4C

f. :.

1 PT. MO. DRR-AF-001-BB l' ATE 5-21-80 0 0 [I},i E:..AP?il_lAU_ Feodwater System DESIGN RFVIEW CHECK!,lSI-s" ( e e I FJr. i :n.i E i ~~ FOR OFFILE USE ONLY - NUT TO lie 'j p, _,, ,.3 p.,.,9 f i" *N '. E850 2 Of - SEN1 Ut> T 51 'DE OF S t. R ', E N T f., LUNDY

t:r;u..: e u sL,,

_ _. = - . _ - = - ....... ~..

...-,-..y-- 'CO:1.!EST' I REVID.'ER' S CO:0 L:iTS l i Nr:tr,T.R NAME_.T 41 n.j h.^ ntSOLUTION h '1. Byron AFJK-1 design calculation None Required shows that with both pumps running and taking suction from the condensate storage tank that the suction losses are too great and that the valves will open for alignment to SX system. DRN P&ID-120 added extra 8" line 1CDC6A8 off of 14" line to reduce head loss to provide adequate available NPSH. Calculation I AFJD-1 shows 23 feet F SH required ~ and 66 feet NPSH availabic with' new routing. / 2. Byron Design Calculation AFJK-1 None Required is the calculation to determine minimum flow required which agre.cs with Chapter 15 of the FSAR. 3. Design Calculation AFJK-1 is ,None Required , performed with steam generator pressure at the safety valve set point.plus 3% accumulation. O 4. The AFW system will be tested on None Required a periodic basis, however, this is a requirement of the Co'mmonwealth Edison Operating s' Department. A flanged connection m is provided to test the system for flow and operability. i y%. (M-37) T. Hottle. 1 o 1 o l* 5. Auxiliary feedwater system 1s None Required i1 a moderate energy system. !k (T. Hottle).

1 i

l3 6. Specification F/L-2907 requires None Required the ISI contractor to perform a ,1n walk-thru of the plant to inspect l&h for accessibility. Modification to the AF system will be made l m l;a 7 if required. (R.. L. Johnson). g 15. ~ ' I' M F C (i t. Pl ! P - -~ ' ": r).s O 7 ;?

N T S T.*, ; * * ". ~.'.

} l s ? l,,0. D"R _.r-M1-E DATE 5-21-c0 ~ ~ ..u s e ae b. b 4. I i f.pc ; ;...,,...,,,. j g!,...,, ran nrr a.c usc a.w_v .m c es SENI OU T S ! DiI GF SANGENT 6 !UNCv ... '..,,. 2.j;, . _J _.,,. 2_ .'[,, L~,_.____ m -~.

I,m--CO."J11'.Nl' ' E.' ' 1 LWER' S COM'!ENTS W.'.95R NAME,_,.db.ic.s b'.*- RESOLUTION 7 P&ID M-37 specifies all b'o'undaries None Required h. of AF system. 8. As of May 15,1980, 39 of 39 None Required auxiliary feedwater' subsyst. ems have been analyzed. (B. Tatosian) 9. Westinghouse has changed the None Required design of the inlet pipJng to reduce water hammer problems. ~ A.F.'line connects into separate l 6" line instead of feedwater line. This was a Westinghouse design modification for water hammer prevention. (T. Hottle). 10. Two trains are provided, cach with Mohe Required its own AFW pump uhich have divers 2 power sources - one being diesel to insure operation with loss of offsite' power. (F&ID M-37) 11. Westinghouse functional diagram Byron project must resolve thic co flic 10SD685 sheets 5 and 15 and between design and FSAR statenent. Sargent & Lundy diagrams Either FSAR is wrong due to change in h M-4037-1AF01 do not show the requirements or design must be checked auxiliary feedwater pumps starting and/or corrected. on loss of offsite power. (G. Sensmeier) U 12. Automatic switchover to SX suction None Required fs provided in valves 017 A and B .'bj and 006 A and B (P&ID M-37). All ( U other valves aliga automatically. s\\ No operator action intended for e safe-shutdown of both units. \\ (T. Hottle) .a n, n. c. o. 4 NR

g v>

f. E. gg 10. DRR _AF-001-BB DATE5-21-80 -MrcpANicts~ nwApforNT S T a ' ~..- j c

n. m DESIGN REVID? C01:4ESIS l

O ,..L5.. p...- gi;,jy,g.g ._j q.gg g n ;. g _ g. I, i ren nrr:cc est cu a v - ren,a or q,.;.g...,.,.. y, l ' ". '

r*e'ta ts t.'~ ~' 7 p!

! M1 UUT5!DE~~Or SM;GE NT 6 LUNDY i

t 7 ') page y L'- g .~ ..., ~,-..

/ lIl l ,, a...- bt.. CC'J.'){ ^* = " v REVIEW CONSIDERATION YES NO N/A .;7.: m 1. Were analytical drawings signed, its information reviewed and approved by 0. P!1D?

2. ' Were there any 'special modeling. tech-

.s niques uced in the analysis for equip-Irent, restraint, pipe, joint, etc.? Explain! 3. Nere the thermal expansion modes, asso-ciated anchor movements and associated number of cycles, specifieri and docu-mentary verified and approved by PIID?" 4. Were all the thermal expansion modes N analy::ed and found with.in Code allow-ables? [ 5. Were the motor-operated valves, weight, and C.G. documentary reviewed and approved by !!DDD/P !D, before used in analyses? / 6. Wha t was the acceptance" criteria used 4 for analyzing pipe under hydro loading,, and were these loads transferred to SDDD? 7.

ce the pinned restraints under hydro

^ testing specified and documented? h 8. Did the design specification state the g 3 peak pressure to be used for Upset, g Emergency, and Faulted condition, and were these pressures used in the stress O>',: evaluation under these conditions? lb "c. i l 5 9. Has the dynamic interaction of nonseismic X l portion of the piping been considered on lA t.he seismic required piping subsystem? l How is this docunented? i*

,y

10. Did the number of modes extracted have

/, l U.,., sufficient frequency to exceed the ' @* R cut-off frequency? <i e r4 n y[ m k< ~

..g c a.: x u g w,. e.

s .u = = i U R."T. 50. DRR-DATt. [ 'e

. !TLE.:

DESICN 9FYlfk' CHECELIST - i n. r - ~. REVi -Ir.Ec.)j:/,-.i l g.., n _,.,

]

~~~ } l FOR OFFICC USC CNLY - NOT TO BE ,,g.,,,,,, !t I P3IC 1 OI 4 l SENT '3 U T C I C E OF SARGENT f, LUNDY y s.sul L

l 6.l.: D ~. - - -. y-g CO W.EN? REVIEL.' CONSIDERATION YES NO N/A .;t y. r p,

11. Were the correct response'spectras with Q,the correct. damping factor enveloped for all points of excitation (anchors

'. and restraints) before analyses (include .the R. S. of steam generator)?

12. Was the interf ace information of NSSS vendor considered and used in the analysis?

13. How are the closely spaced modes 4

combined?

14. Table 3.7.3 of the PSAR states t' hit'under 3

SSE, piping Group B of 8" hnsi lar3er should remain functional; what was the method used to justify th,at?

15. What was the method adopte'd 'in analyzing 4

~ the large continuous subsystem which exceeded the size limitations of the available programs (statically and dynamically)? n ll

16. What was the methodology and criteria 7

'r used in analyzing small pipe subsystem Q (instrumentation tubing) which is seismic qualification required? 'l. L7. Did the maximum length between any two I 'N data points in the systems analytical model have a minimum frequency content larger than the 31 i:v.? s. et ! J 18. Were the seismic differential anchor' s 'd movements considered in the analyses? ifere they furnished by SDDD? ca s )

19. How the differential. anchor movement 3

combines uith the other loads, and what. s were the.allowables used? I's

20. Were the equipment reactions within the

( , l, R allowables provided by vendor? '#*e.1.'Were the valves G level within the 7 l" vendor's specified allowables? 4 v t !oIIRE

e E c n.w i : r. _ ;E Ar-"E.T r-.

?.r!. 50. DRR- ' DATE ~ DESIGN ?.E'.*:E*? C$ECKI.ISI R R':_ .. ~.. ..) .b! I - j?

l' "

.r, n,.,.. I,,: b,....,. j! S - o FOR OFF1CE USE ONLY - NUT TO BE. g l i t G J'.J ..uf j. r l s t. SENT-OUTSIDE O F' SARGENT f, LUf!OY .,,y,.,. . * ~ ' ~ '~ 0-

&MO SLE COW NT REVIE'J CONSIDERATION YES NO N/A g;Qg

22. Was the Snubber' rating checked to verify

.d that they can carry both Upset and h Emergency condition,s? 23'..Whatwasthe$ct$odologyusedtoeval- / unto SSE response, and is it justifiable?

24. Were the pipe reactio'ns on valves with3n

~ the allowables, or'sent to manufacturer for confirmation?

25. Were the flanges within the systems

4 O

checked against.NC-3647?

26. Were br'anch,ed piping reviewed, modeled

'N properly, checked as per NC'3643.3, and results furnished to MDDD for implemen- .tation?'

27. Were there hydraulic analyses performed, X

and are the piping responses within ~ allowables?.

28. Were there hydraulic transient analyses g-performed, and is the piping capable of /

. withstanding the transient forces? h9.-Herethe'RVsizedappropriately?. Fere -" X the forces at the RV exits considered in the analyses? Were the supports designed for static, dynamic and/or wind loads ' reviewed? (.;30. What was the criteria used to account for N 'f .g the flexibility of the piping restraints y for dynamic and static analyses? . a $ 31. Were the reactions at the penetratioris X. (in and out) been tabulated, and docu-I mentsry forwarded for use of penetration I design under Upset, Emergency, as Faulted s condition?- ~ 2 a .n o

n. n

. n

\\ c4 a.y n I U 1t< RPT. 50. DRR- -..,DATE t.:ec E: c a t osn.m ca.T s g"ITLE - - - DESIGN P.EVIE'* CHECKLIST - =, .=1-;E /.

b. \\T:. I I

FOR OFFICE USE ONLY - NOT TO DE l ] n,,'.~.,.,y,'j' l " *l "'"'b of4 .SENT GUTS 1DE OF SARGErD 6 LUNDY j cu=manouG; ?ne 3 s ... - ~

j. ' _ ' '. I:. i i...\\ i Shi. C ^ '. "~. '~= "y, ' ~ - y,].. REVIEU CONSIDERATION YES NO N/A

32. Is'there" piping'in this system requiring j'

i,.e pipe break or pipe whip analysis? ,If not, continue question 33). ( Was the cEiteria used in postulating a. break location, break type and restraint location, proper and correct? g b '. Was the criteria used in designing pipe whip restraint. proper and correct? c. Were the pipe break loads, pipe whip loads, and jet. impingement Joads calcula'ted correct and proper and transmitted to the pertinent depart-ments?

33. Did all documents examined in the course o.f review appear consistent with each other?

~ m 1) - l,i i. .u s u

t i

s AJ o.

rJ .g.' m 4 r n eh i .. N, i.s-e, N e g

$cwAN:ca 09;r7 ant

~ s m. : _.

PT.: :0. D:'R. _.

DATE . ' ' ' j ~e. .g.EV: DESIGN F.E'JIEW CF.ECELIST - gP l - e.. f],,,,GG,..,ll..n,,,,.j[ ,.. 6 FOR OFFICE USE ONLY - f40 T TO B E' u.si Jiw g 9 4.., 4 nc.' .i SENT OUTSIDE O r~ S A R t; EtJT 5 LU JDY, ..:.:.uru a. ' :-

F U. 'C0YMENT REVIEL.T.R'S COMMENTS 10M5ER NAME IESOLUTION c.:. u te m..A ' :EId N A '"* 'f.}.. I @.d : T u d u ;li. ' 2 * '- "~"'

ia ;1..

~% y c w ~*-'--' ~~- y*n . ~ s.+s us.. ' ~~ ud!&i2 S R1's Un o:; c.nr tan) ,. u i7.. *.. b'Ir' C.Q 0:1.t5 c;o d O c M.<A.,.!j je .] l - { ts ( (( g C. . 'l~ .Cb. # 'l = = t. ,,b Y%d ,4.'.Le '. s b ,4 - J.. v.r 6 p ~'c.ts., cw.w. n c, -{ u; J'I -DJ LLW ... b .s .J, /. C.. x. ; '.,u,.'s '. t

v" ;f

f. "..

i V- '.d.:+ (.:((r C, <.S s., ('.2w L c: w tt . i-. J l ,,%. L si. Q $) s'~t k$ faj tz-e* s ** < f ~ (, = 5 CH *,.i t2 E!.2 llJ <* l i e i e L.: 4 l Id M ! ,f". , '/./ [3 c' c ;,%.( t l' b a 'l s m.g d [f t s4-l o. 2.,.

:,' s s 's 3

u ,r. ;. l,,f, sss*.- -% -c I 7 h JS1E Ca/nise U" d6 M .y s. l dLs !L ks . [fs e;9 bL' < l! R l*$ ! y .j s. e i s,t.

s. X l s...'o ':-

vf>'Ci V*s l, e CcMCtaw[f..c-s..t % f'v. /- n m a . OJ tw.w.u 3:kr / ' ' ' '-.. a. tJ er. s \\ l'I <t & M g L .;. u....., i: :i ;.- d u L.A %.,: ^ ' ' ' k + 4 - a i u<.. alt.,; w Y.;. I ' ) t 's : e ~. t. M., ^-- ., u. , a> .y 9u f T..j-x C u > s<u ll' L s<.Cnt:0:-J' .a n a s t 4 s,' r. v - e g < r- .m I -\\ r [j$

-ly -

o N (G ~.c .'.* sti c +-, t.J & I roi (.'-t.. s. - -r' 'J ff s En,,

t. f.

J.,, ..,,,x p..qy): jw

;.y s a

g<

M C H t. ' ~ : - ~' : " A P ~ ' * ~ '.' S__~ _ ' ' F : : E DATE u > NO. ORR-r - + - 1 I DES.... .s.I_.,,Cv........n s

1...,i

.s .v:. r.: ,.I._ t.... [ 4 4.:. w e I i FOR OFFICE USE ONLY - NOT TO DE g,,,.,,.., m,,a -..T,,,,....,, p, %..,.. SEN1 O tJ T *.> I DE OF St.RGENT f, UNDY _, s m a a r. u l n.r g 1 ~ ~ -e===.-=."

==.-.--..-=-*-e.

.. - - 8t).3ON !ijI)Y ~ 3 s ta.n en s. -/{2r >r u. r y r.c Rev. O [Date :: a,: gl

- - - C^*

Safety. Rol:ted l Non-Safety-R eiated Pa :e of Cf:ent CC"C O Prepared by

2. th Date 8-l'o.L rnini DYRml /BRAtfzytopp nevined bv Date P;nj. No.

4 3g /-O(> Equip. No. Approved by Date / I ~~2bcukiMT~ hlo.

*T 171 E ib V. ~ " 34 7N - "

W E W in K S ~ ~ WHD ~oo &4 /7 b -of ..... O_ //.-G.78- 'Oo84.1f--... l.A8=07.--. lo-. -. o 7 'o 7577 - -- ----- j N 020o24-.. .-..A.B -(3... o. .. to -3h7f l i -.Of-t6.-j7.-. l 1 = ' i l--) !.o 0.f 7 / 6..-... .8R-of R. _ Ofo45c.--. .. BR.-- I'd - Lo.. -..10 -31..--7.7- ' ' O f 04So. .-. B R. --J $ I 0.. .. /o -3I.77.... l l-l ' ooat43.-.-...c.c-of.-.J.. _... o.. --.of=28:76..-. _{l_. ) -..l. 'o /.7 27 {.. . cc. -tI.---- L l. : 03-2' :7f. -. V 1- - -l l b oCo74-- .._jcc.-l8_... . 2_ 07-(o 7 _._ l 4 --... of 9 ys'.. .: cc - J..- -..o$..28.=7L-. .---l ._ --. o t/ 20// ; 'CS-ef ' / I2.. -At.-17.-. L.- --. - l '.o. /$. 3_/ d 1)(..- og_. /.. - _.. tt-.m.. 78...... i -.i. = j g Q .P/f.432. .-. M. - 3 ll -- J. -. 03 Af.:7f.. .. _ i Oib4o3........ Q -o$_..L --.f.... c3-o2-7f. ...7- . c c N L-2.... 4 l= - tA..-- -.o --. of - to : 72-i --... o o 03 7 .3D-A3 -..-. o. - -. 09.-d.. 7. 7.- ..-...a /..t-L 8 ' pgd-oJ' ^ O to... 30. 78 . -.. o I 3 1 & I. fl#-0 S O' -- oS'-.ii~72... ..:..-.. 0/d / 2..[. -.1/d -- /L -.-...o.-- -.07-1.f i e. a . o I 3 4 0 f4.. IO:'-!l. /.- o.?:-!f.-]2 i ..-.-.013/ o.. $ VD -62.-.-- .-.l- -..o V -!E. 78 - A IL fl .. O /!!S'8. Wb -o f.. 7 .. O. a -. 0 8 .23.-78.. } ... 0 /5 8 / 3...- .... b9 - 0 2...- -.

o...

12 - 7eP.--. E ...Othb b. -.. ) 6 - l Q -..-..-.. D.- -_o2 %S.2 7f._.--- -.: 7 .oI t.ol 3. . 3X - lL.. .3. . /2-20 ~ 77 b ---, 0 05'8./3*.... .. C3 --02.. ..l. .//.-os-;4.- )' .,.... :01 y42 -... .- F c.-- 4 _. ...o, __ ...- cs-2.o.7.y-f d.I b. ';I . __. _ :... _.L ?... ..l... O rwamu xa re air, alta.: 3. T /\\ T o.5 ( M t .J~. I. dcL C.S c R C. U!.t:* R. J. RA [(Op)yk[ _..L...._ .T. 4. W lh T E A J N 8.L r q

) . _. _ ~... .l C'":,d. ;."'"..' .l RE, VIE'.I CC'ISIDERATIO'l (CID) YES NO !!IA .Z.1.O c ..e --- < ~ g a.-- - l ~ 1. Have "in service" tes't requirenent pro-X. 1 visionc been met, dnd accoss clearances J Q and appurtenances neen provided for- . test? 2. Have the seismicall'y qualifi'ed instru-X ment.s been seismically. supported? 3. Do instr.ument sensing lines penetrate X 'the containment? If so, does the ~ in'stallation satisfy the requirements of 10 CFR 50 - Appendix A and Regulatory Guide 1.11? 4. Have the sensing lines of nuclear X 2 s'afety-related instruments been~provided - ith the identification tags as called w for in the " Process Instrument Piping and Installation Design Criteria," DC-ME-02-BB?

5. 'Have redundant instruments and the X

'3 respective sensing lines been prov.ided with the necessary separation or bar-riers as called for in the " Process ~ Instrument Piping and Installation I Design Criteria," DC-ME-02-BB? O G. Have the instrument sensing lines been X designed to acco:modate the process ) pipe thermal. expansion, the seismic vibration, and the thermal expansion u' , e: of the sensing lines during blowdown? h(, ~ 7. Have seismic qualified instruments been X-t,. located such that the seismic response w 3 spectrum values are not exceeded during a Design Basis Earthquake? } 8. Do the equipment performance specifica-X tions (instrument data sheets) agree. l f. with the latest calculations (process, g, parameters)? c, ~ r#<y 9. Is the equipment selected suitable to X ", j r;' perform the intended function? , s, v m f< i = c. f'!'C H A ta I C At O F AFTidi.*:T 5i A: T w> '. I'T..w. DRR _A.F-OO1-o,B DATE 5/23/30 -'D =- o,. ! I TI.E : Aux 13iarv Feccwater Sys:em Ge. DES IC.. REVI E., Ci!ECKLISi - Mx: 0 .--3.-. ' E E - lI\\IE L. _ c--,,,--- FOR O!:FitE USC UNLY - NOT T O lj E j t p:,f.,,.,,,...,.n s.l,.e s e c.: <,.u k. o ng,g., set 2T uvi'; TOC U r' S A n t,C N T ' f,' t U N D Y 1 n.~c 3' c r s i ncm.m.w u ...._...n.

\\ . J.. .r. 8 C"'..".".' . -.'. '.".. '.. ;/. '.,'.T R[VII.'.! CO::SWI:! ATIO t (CID) YES i:0 ?!/A- ..l.. 4 s = - . i ~..... -. o 10. Are the materials of construction X compatible with the environment in C..j,. which it.must operate? r. 1 t i s ) .~ n. o . ) t.\\ t.1 c. y :u .~ v.t x. 'k ) 5 ';',a u c., o a g >s s .a i ac, s '. c , *a .%A 1 !sn AN 1 <; s l t e

ca I 7 L) e t l

' t V 1 vs if < ra i :..5..=a ~- l m c c>u n : c v. I c. >. Kl'T. ';0. *1RR-AF-001-I,3 DATF. 5/_23/80 s oe n 't Iit I': Auxilinry Feeci.ca ter Sys tem L D_.Sl, .......,, b.., t.... L.., l... - r .s:... r r. :. o i O_ i. r.. -.!I'

N.--

b.\\T.~ l... - 'I FOR OI-r!CE USC firJL Y - tJU T TO tlE i ~ - - g..e e t e ; (' l -- ~~ ~' I* a 0.. :. r. t i.... :.... n,.. 9 o'r o I a.N I Ut> I S I DL OF S A8 t CrJT f, '. Ut;D Y l Cis '..) da I i 4 r l i 3 is '-(.- ~ _..s,.. , n,..,, a } l -..........._......~.m...,....

pe 3 r, April 19, 1983 COMMONWEALTH EDISON COMPANY QUALITY ASSURANCE STATEMENT REGARDING VERIFICATION OF ADEQUACY OF DESIGN AND CONSTRUCTION OF BYRON NUCLEAR POWER STATION UNIT #1 The purpose of this statement to the United States Nuclear Regulatory Commission by the Manager of Quality Assurance of Commonwealth Edison Company is to describe the broad range of project design and construction quality activities which enables him to affirm the quality of the design and construction of the safety-related and ASME Code-reldted aspects of Byron Nuclear Power Station Unit 1 and its associated facilities. A. GENERAL Commonwealth Edison Company is ultimately responsible for the assuranhe of quality in all phases of the design, procurement, construction and preservice testing of the Byron plant. The Commonwealth Edison. Company Quality Assurance Program, which has been approved by NRC and found acceptable by the American Society l of Mechanical Engineers (ASME), addresses the requirements of the l 18 Criteria of Appendix B to 10CFR50 and Section III of the ASME Boiler and Pressure Vessel Code as well as the applicable ANSI Quality Assurance Standards and NRC and other Federal and State Regulatory requirements. l I l J

[' ~ Page 2 The quality assurance programs of the various vendors and contractors, involved with ASME Code-and safety-related items and activities, are also required to adhere to the requirements of the applicable Appendix B Criteria. Further', the Commonwealth Edison Company Program requires that its Quality Assurance personnel review and accept all such vendor and contractor quality assurance programs (including those of the Architect Engineer and the NSSS vendor) prior to award. Commonwealth Edison Company's Mat.ager of Quality A'ssurance directs the quality assurance activities related to the design, procurement, construction and operation of the Commonwealth Edison Company's nuclear plants. Commonwealth Edison Company's Quality Assurance organization is completely independent of the Engineering, t Construction and Operating organizations, has freedom to identify l quality problems and reports directly to senior management. Since the Commonwealth Edison Company Quality Assurance organization was established in 1973, the Manager of' Quality Assurance has reported, successively, to an Executive Vice President, a former President and presently the Vice Chairman of the Company who all have'given, and are giving, full support to the Quality Assurance Program The people that formulate the Commonwealth Edison Company Quality Assurance organizations consist of graduate engineers of essentially all engineering disciplines related to. nuclear and non-graduates with years of hands-on experience such as involving maintenance, construction, engineering, and operating Navy nuclear and other technical activities. Having qualified personnel is a g

Page 3 basic requirement for a Quality Assurance organization. To assure i that qualified personnel have adequate knowledge to perform their work and maintain their prof 4-;iency with respect to such quality associated items and activities as Codes, Standards, materials testing, nondestructive examination, welding, auditing, inspection, coating, fire protection, etc., formal training and refresher training are routinely given. Generally, the training of Quality Assurance personnel is an extension of their formal education or work experience. In addition, the Commonwealth Edison Company Quality Assurance personnel are required to be qualified and, in l turn, certified for specific functions involved with inspection, l surveillance and audit activities such as covering the items listed above as examples for which training is provided, and are required to adequately demonstrate their qualification to perform specific I functions before they are allowed to perform the quality assurance work activities for which they are qualified. Furthermore, Commonwealth Edison Company Quality Assurance personnel are l required to be recertified on a two-to-three year cycle. l Qualified Quality Assurance personnel are assigned to all nuclear construction sites and operating stations and the Commonwealth Edison Company's nuclear testing and engineering l organizations. A Quality Assurance Coordinator, who reports to the cognizant Director of Quality Assurance (Engineering-Construction), has been assigned full time, since 197S, to the Commonwealth Edison l Company engineering organization which has overall design l l l L 1

s. O Page 4 i responsibility for the nuclear stations. The Quality Assurance Coordinator reviews procurement documents, specifications and procedures and performs periodic audits and surveillances of project activities to verify compliance 'to the Commonwealth Edison Company Quality Assurance Program and applicable procedure requirements. Each Commonwealth Edison Company corporate department having nuclear related responsibilities, such as involving engineering, purchasing and testing, is audited annually (Exhibit B) to the requirements of the Quality Assurance Program, their respective department procedures and other requirements by Corporate Quality Assurance using qualified and certified Lead Auditors and Auditor personnel who are not directly involved with the organization being audited. Also, the nuclear construction sites are audited by Corporate Quality Assurance on about a four-to-six month cycle. In addition, audits of the other l construction site organizations and' contractors are performed by ( Commonwealth Edison Company Site Quality Assurance in accordance I with an approved schedule ranging from monthly to annually ' depending on need or established requirements. Identified deficiencies are required to be committed to be corrected or dispositioned in a timely manner. Audit deficiencies, involving Commonwealth Edison Com'pany Departments, which are.not corrected or dispositioned in a timely manner are referred to top corporate management for their' handling. Where serious problems or inaction l l l

4, Page 5 to solve problems are identified, stop work action is initiated. The Manager of Quality Assurance has responsibility and authority to stop unsatisfactory work or stop further processing of unsatisfactory material during design and construction of the plant. Conditions found by quality personnel, which are adverse to quality and require prompt action and corrective action but cannot be resolved at the respective locations, are required to be promptly reported to the Manager of Quality Assurance or his designated alternate for action. Furthermore, resumption of work under a stop work order requires concurrence of the Manager Quality ~ Assurance. There have been nine cases of stop work at Byron plus two at a vendor plant due to Quality Assurance Program implementation and welding problems (Exhibit C) ranging from a very short time to many weeks. The extended stop work cases, for the most part, involved administrative, training, inspection and/or procedure. problems which had to be corrected before work activities could proceed. l Currently at the site, a stop work order is in effect l against Reliable Sheet Metal, the HVAC contractor. All new work on l safety-related systems and attachments to safety-related structures was stopped on September 17, 1982. The stop work was issued because of inadequate and incomplete inspections, inadequate 1 procedures, lack of some procurement documents and open audit deficiencies. The stop work will remain in effect until these items have been corrected and the corrective actions have been accepted by the Commonwealth Edison Quality Assurance Manager. A

Page 6 At the construction sites, the Commonwealth Edison Company construction Quality Assurance organization functions independently of the Commonwealth Edison Company Construction site organizations under the direction of the site Quality Assurance Superintendent. He reports off-site to a Director of Quality Assurance (Engineering-Construction), who reports to the Manager of Quality Assurance. The Byron site construction Quality Assurance group has expanded over the years to keep pace with the construction activities and currently consists of about twenty qualified' people augmented with an Independent Testing Agency contractor work force of qualified inspection and test technicians. The site Quality Assurance group performs audits (Exhibit D) and surveillanc;s (Exhibit E) of site construction activities. This group checks for implementation of contractor and Commonwealth Edison Company Quality Assurance Programs and procedures, and. assures that on-site design functions are controlled and performed to applicable requirements and that the plant is constructed and tested in accordance with approved drawings, specifications and procedures. The-Independent Testing Agency contractor, consisting of between forty-to-fifty qualified inspector type technicians, performs normal independent inspections, tests and examinations of contractor work at Byron under the~ control and direction of site Quality Assurance (Exhibit F). Site Quality Assurance also uses the Independent Testing ' Agency to perforn special inspections as deemed prudent or necessary to check construction work performed and inspected by contractors. L J

Page 7 Also, to cover plant operation activities at the Byron Station, the Commonwealth Edison operations Quality Assurance group was established early on. It has ov,erall responsibility to ensure the activities and functions performed by the Station staff are carried out, as required, during the formulation, pre-operational testing and startup stages of the project as well as during plant operation after issuance of the plant operating license. The operations Quality Assurance group consists of seven to eight persons who perform inspections, surveillances and audits of plant operation activities covered by the Corporate Quality Assurance Program such as maintenance, procurement, pre-operational testing, plant start-up activities and associated administrative and other functions related to plant operations. Special emphases were placed on performing surveillances and audits of each system pre-operational test to ensure requirements were fulfilled. In addition to the normal auditing performed by each site l contractor (Exhibit G) and the surveillances and comprehensive audits performed by Commonwealth Edison Company site and operations Quality Assurance, site design, construction, testing and Quality Assurance activities were also audited by Commonwealth Edison Company Corporate Quality Assurance (Exhibit H) on a four-to-six month cycle throughout the course of the Byron Project. The Corporate Quality Assurance audits were performed to ensure that the Commonwealth Edison Company Quality Assurance and other organizations and the contractors on-site were performing their 4

Page 8 responsibilities acceptably as required by the respective Quality Assurance Programs, design documents and procedures for the ASME Code-and safety-related aspects of the Byron Station Project. Audits of architect-e.ngineers' -and NSSS vendors' corporate activities wicit respect to design are given special attention by Commonwealth Edison Company's corporate Quality Assurance management. All such audits are conducted under the immediate direction of either the Manager of Quality Assurance or the Director of Quality Assurance for Engineering and Construction. Furthermore, one of these individuals has participated with each audit team in almost all of these design audits. The audits covered all aspects of the Quality Assurance Program and in many cases went into considerable depth in specialized design areas. The methodology utilized by Commonwealth Edison Company Quality. Assurance to achieve an appropriate confidence level regarding adequacy and accura.cy of the design and construction includes 1 f inspections, surveillances and audits (Exhibits I L J). Although l inspections, surveillances and audits do not in themselves result in specifically checking the design and construction of each and every item in the plant, the results from the implementation of. the Commonwealth Edison Company Quality Assurance Program has proven to be sufficiently flexible and extensive'to cover the various areas of design and construction. In addition, this technique has resulted in Commonwealth Edison ' Company Quality Assurance Engineers and l l r ,.r

Page 9 Inspectors examining, in great detail, a wide range of selected' items and areas, including those reported to be of concern at other project sites such as troubles with vendors or contractors, construction, equipment and operating problems, failures relating to implementation of Quality Assurance Programs, design and design control problems, welding, nondestructive examination and inspection deficiencies, etc. In particular, special attention has been given to items-reported as identified problems by other projects across the country and to items brought to our attention by NRC Region III Inspectors, particularly involving design and inspection activities. Nuclear industry anb NRC reports of problems are reviewed by the Corporate Quality Assurance staff and pertinent problems are brought to the attention of our Quality Assurance personnel at all locations, by separate memoranda, who in turn are required to investigate whether similar problems exist in connection with any of our nuclear projects or operating stations. l Furthermore, because of the comprehensive coverage achieved through i this program, Quality Assurance Engineers and Inspectors have reviewed an'd verified the design and construction of a broad scope of items and components installed at Byron. These results are in addition to the reviews, checks, inspections and audits performed i by in-line organizations responsible for design and construction. Also, the NRC and ASME, starting in 1972 and 1974 respectively, have conducted inspections and surveys of Commonwealth Edison Company, to assure that the Commonwealth Edison

Page 10 Company Program is correctly and properly implemented. ASME conducted surveys of the Commonwealth Edison Company corporate offices in 1974,1977 and 1980 (Exhibit K) which resulted in ASME awarding Engineering Organizat. ion and "N" Certificates plus NA and NPT Certificates to Commonwealth Edison Company that grant it authority to design, fabricate and install items that must meet ASME Code requirements. In all cases, the ASME Survey Teams reported that the procedures and management controls were sufficiently effective to im'plement the Commonwealth Edison Company Quality Assurance Program. Overall, not only did the Commonwealth Edison Company successfully pass all these corporate ASME surveys but as a further benefit all departments became more involved in Quality Assurance and gained significantly in the understanding and application of the Qualicy Assurance Program - a significant contribution to the overall quality effort. The Commonwealth Edison Company Quality Assurance efforts during the course of the Byron Project have been actively, energetically and intensely performed to ferret out problems, obtain corrective action, prevent recurrence of problems and assure that the proper engineering information was used in generating the design documents; the proper design documents were used in the construction of the.pla'nt; the construction was performed to approved procedures and is in accordance with the design documents; and the inspection and testing of the facilities was performed, .,,. -... - - ~

O Page 11 reviewed and accepted by qualified personnel. These verification efforts have been carried out with multiple tiers of inspections, surveillances and audits by contractors and Commonwealth Edison Company. Quality Assurance and other outside organizations such as NRC and ASME plus consultants, such as Energy Inc., who have performed management audits (Exhibit L) of Commonwealth Edison Company Quality Assurance every other year as to the implementation and effectiveness of the Quality Assurance Program in Commonwealth Edison Company. Also, top corporate management of Commonwealth Edison Company has been kept fully informed by Corporate Quality Assurance about the quality aspects of the Byron Project as well as of the other nuclear construction projects and the operating nuclear stations on our system. A copy of each Commonwealth Edison Company Quality Assurance audit report is sent to the Manager of Quality Assurance and to other corporate management having responsibilities involving the Byron Project. In addition, detailed monthly reports covering the Commonwealth Edison Company Quality Assurance activities, audits and problems are provided to top corporate management in order to keep them informed. Also, items that need upper management attention are identified in this monthly Quality Assurance activities report for their corrective action. When such identification has been deemed necessary by the Manager of Quality Assurance, prompt action is required by senior management.

Page 12 4 i Finally, an annual Quality Assurance assessment report, J covering quality assurance activities and problems involving all - Commonwealth Edison Company organizations and locations checke~d by ~, ~ - ,~ Quality Assurance during the year is provided to top mancgement'- having nuclear related responsibilities. This report provides'~a -cts candid overall assessment regarding the implementation of the Commonwealth Edison Company Quality Assurance Program including contractor construction activities. Problems experienced during. the year are addressed and where more attention is deemed needed by upper Commonwealth Edison Company management,.bsually relative ~ to attitudes, approaches and resistants; such problems'also are presented in the report for corrective action and dicposition. The reporting concept is similar to that (1 sed with financial audits; that is, bring problems to 'the att;$ntion of the ~ responsible manager for correction. A difference under the Commonwealth Edison Company Quality Assurance Program, however, is that stop work action by the'Ifanager of Quality Assurance is exercised where a serious problem needs immediate attention. In addition to the above verification activities which ,a x have been performed regularly thcoughout t,he course of the T s construction project, a special intensive evaluation of the Byron project was performed in late Octeber,1982. The basis for' th6 ~ ~ evaluation was' the Institute of Nuclear Power Operations Performance Objectives and Criteria. Commonwealth Edison Quality Assurance coordinated a self-initiated evaluation which was se k

,s 'c \\ v s- - s_ w. t' Page 13 s- ~ 'r carried out by a twenty-man team. The team consisted of senior N 'l management personne,1 with broad backgrounds in construction, n engineering and operating <along with five consultants. The team looked,.in depth, at the plant facilities, work under -construction, construction practices, design input, design output ~ and design r'eview. The Byron project was found to measure up well \\\\' 4 s against the IHP0 Criteria. Corrective action for the deficiencies 1 ~ identified by-(the Team wdre readily undertaken. It is felt that i ~ this industry evaluationiadds ponsiderable credibility to the view O that the,Byfo,n Unit,Qne facilities -are adequately designed and 7 1 s constructed and' are pf good fuelity. 's'- B. DESIGN' VERIFICATION f-N commonE'ealih Edison has placed heavy emphasis on the '~ s e i s s s-s s s s, ' ' evaluation and verification of design. Engineers qualified in i \\ 4 -

& ^1 specificfdesign. disciplines ar\\e used in these design verification s;

\\ audit ef fc.rts.- This is accomplished in two ways. Engineers from a .l\\ i L ~.. - - consulting firms are hired to perform independent design verifi-cations aSd Commonwealth Edison Quality Assurance performs its own i ' .a \\. ve'ri'fication of design with audit teams of Edison design engineers. y vs t,- l'. _ Verification by Consultants ,st .Q 7, Energy Incorporated has been involved in three ( s... recent technical assessments of Sargent & Lundy (SLL) for y Commonwealth Edison involving design control and design \\ verification. In June of 1982, Energy Incorporated s participated in a design audit of SLL, the architect U, 4 .3 \\ 's s 4

Page 14 Engineer, and did a thorough assessment in the design control area involving the Byron Project. Again, in October 1982, Energy Incorporated participated in the INPO Self-Initiated Evaluation of the Byron Nuclear Station Construction Project. In this Evaluation, Energy Incorporated was responsible for assessing the effect-iveness of the design control covering design input, design output and design review. In both efforts, Energy Incorporated found that S&L is doing an effective and thorough job in the design of the plants and has an effective design document control system. During these two efforts, Energy Incorporated's assessment covered the following specific areas: Engineering administrative procedures Design criteria Drawing Equipment & System specifications Stress analysis reports Design document control system Design interface controls (internal & external) Class lE equipment qualification program Design review and verification program Computer program verification The approach used for the assessment was to review the S&L Quality Assurance Manual, various Department Standards, and the Project Instructions covering the design process to determine if the'se documents were thorough.and, if properly implemented, could result in a well controlled design

~ Page 15 process. The specific areas named above were then evaluated through document review, personnel interviews and observations to determine if the engineering procedures were effectively implemented. It was found that the S&L design control program is very thorough, well understood by the engineering staff, and effectively implemented. Energy Inc. found that S&L's Quality Assurance Manual gives the overall direction for the design process quality program and requires the development of Project Instructions to provide additional direction to the staff members assigned to each project. These instructions contained design interface and review requirements and contained tracking requirements to assure the various design interfaces occurred. It should be noted that all the project instructions reviewed during the evaluations contained the level of detail necessary to assure continuity throughout the program. It was found through interviews by Energy Inc. that the project design staff was very familiar with these instructions. In addition to the Quality Assurance Manual and Project Instructions, it was found that each department within S&L has its own department standards. These standards contain instructions covering the engineering work being done by the given department. Again, the i s.- ~.-._.. -, -,m,..... z ,,,,,.,.--_..__.-~m._,,_ ,_.-_s-m

Page 16 department standards reviewed during the course of the evaluations were found to be excellent training and instructional documents well understood by the department staffs. Also, to aid the staff members in applying the correct design criteria, the projects had developed general project design criteria documents. These documents were not necessarily a program require-ment but.were generated to aid the staff during the design and analysis activities. These documents are excellent reference documents and will be invaluable during the life of the plant for future reference. In addition, the engineering documents (drawings, specifications, analysis reports, and criteria) were well done and cross referenced. Specifically the stress analysis reports were found to thoroughly reference the drawings and criteria documents by both number and revision. This thoroughness seemed to be consistent in all documents reviewed. At the conclusion of the evaluations it was Energy Incorporated's opinion that S&L has a very thorough scogram in the Design Control area. They found the program to be well documented, understood by the staff, and effectively implemented. (See Exhibit P) 4

Page 17 In the third and most recent audit of Sargent and Lundy in March 1983 by Energy Incorporate, a five man team of experienced design consulting engineers performed an independent design verification audit of the Byron /Braidwood projects with particular emphasis placed on design of Byron Unit 1. The review included, but was not limited to review of stress analysis reports for proper input and use of output, compatibility with design l drawings, completeness of the model, adherence to codes and standards; review of pipe whip analysis for modeling methodology; review of fluid system thermal-hydraulic analysis for modeling; review of a fluid system for classification, operating modes, thermal-hydrauulic operational analysis and safety analysis which establish design input values;. PLID's for proper regulatory and code requirements; control requirements for operational and safety considerations, design drawings for proper carry-through of P&ID and functional requirements, control loop diagrams, completeness of drawing package and design l specification. 1 Energy Incorporated, as part of this independent design analysis type review, similarly verified that Sargent & Lundy is doing an effective and complete job in the design of these plants. -n. ~ -.. - ---c--.r-g ---, y -. -~. - - - -,, -.,...,.- .w.. -,-.,,--,,--,.wm,..v_-,,, ..,-,w ,-,,,n, - - - -,,,.,.,n-

Page 18 2. Verification by Commonwealth Edison During the course of the Byron project, extensive audits of the organizations having design and construction responsibilities, have been and are being performed by Commonwealth Edison Company's Quality Assurance Organization. Quality Assurance has conducted audits, surveillances and closeout surveillances of audit deficiencies of our principal Architect. Engineer, Sargent & Lundy, Westinghouse Electric Corporation, and of o'ther organizations performing design work involving the Byron Proj ect. Sargent & Lundy has responsibility for the overall control of the design, and of the distribution of design information and documents, which is controlled by distribution lists and memoranda. Two methods are used to verify calculations. In the case of design calculations performed by computer, the' computer programs are verified by Sargent & Lundy for adequacy and accuracy of the design process for safety-related and ASME Code-related structures, systems and components prior to permitting their use; and thus, sample calculations of subsequent computer runs are not required. Calculations involving design not performed completely by computer are documented ~ and their review, including verification of sample calculation's, must be documented. ,w-~ - -.,,. - -. ~, - -. - - y

J Page 19 Audits were performed by Commonwealth Edison Company of both Architect Engineer firms' corporate and on-site design activities for Byron. Examples of some of the technical deficiencies which were identified and corrected as a result of these audits are: calculations l for cable pan loading did not properly reflect FSAR requirements; field changes were not properly translated into design drawings; provisions were not made for local annunciation and alarm for fire protection in the battery room; insufficient accountability for concrete test data and reinforcement bar damage data; calculations for electrical conduit supports did not consider the eccentricity of the applied load due to the width of the tube steel; improper horizontal acceleration factor was used in electrical conduit support calculation; a horizontal force component was omitted from an electrical conduit support calculation and calculations for pipe rupture jet forces were not properly documented. In addition, the Architect Engineers perform audits of their assigned design responsibilities, which Commonwealth Edison Company Quality Assurance reviews and l l assures are performed acceptably and, when deficiencies are identified, are corrected satisfactorily. These audit and surveillance activities will continue until i completion of the design of the two units at Byron. l i

Page 20 System design at Sargent & Lundy for the Byron Project was initiated in 1971. A majority of the system design was conducted between 1973-1976, with many revisions being made since then as a result of changing design requirements, changing NRC requirements, TMI lessons learned, design improvements by Sargent & Lundy and design changes suggested by Comconwealth Edison Company Operating and Engineering Departments. Commonwealth Edison Company Quality Assurance has assured by audits and surveillances that systems design was performed in accordance with procedures in place at the time of the design work. Early in the 1970's, an approved Sargent & Lundy Quality Assurance Program was l not in effect, but formal design control procedures were l in effect which required independent review of design l work. Design work since the implementation of the l l Sargent & Lundy Quality Assurance Program in about early l 1974 was performed to procedures in effect at the time of I the actual work performed, which ensured that design work was independently reviewed and approved. All l of this design work was audited by the Quality Assurance l l organizations of Sargent & Lundy and Commonwealth Edison Company and by independent consultants for Commonwealth l Edison Company. In addition, Sargent & Lundy further reviewed the design of the various systems as a result of l l

Page 21 the NRC review of the system descriptions within the FSAR, which adds further assurance of proper system design. Finally, as to final design verification, Sargent & Lundy is required to do independent design reviews. System and structure design reviews are performed in accordance with Sargent & Lundy approved Quality Assurance Procedures and Project Instructions to verify the design of the Byron Station Project. Design review teams of qualified Sargent & Lundy design people, not involved with the project design, were used. In addition, an overall integrated design review of the total design of the Byron Project is being performed by independent senior Sargent & Lundy design people in accordance with the Project Instruction. In the case of both types of review to independently verify the design, documented design review reports setting forth the results of these reviews are issued by Sargent & Lundy l and are reviewed by audit for acceptability by Commonwealth Edison Company Quality Assurance. C. QUALITY ASSURANCE COVERAGE The scope of audits conducted over the years by Commonwealth Edison Company Quality Assurance covers the various aspects of design. The questions contained in the approved checklists for the design audits are comprehensive and cover not

Page 22 only the 10CFR50 Appendix B,18 Criteria basic requirements such as design control, document control,' auditing, corrective action, training, personnel qualification, etc., but also the complex and intricate aspects of the plant electrical, mechanical and structural design. The checklist questions for each audit are developed from established requirements such as those contained in J procedures, applicable design documents, N tional Standards, the a ASME Code, the FSAR, NRC regulations, and the applicable Quality Assurance Program and its implementing procedures. In addition, more extensive design analysis audits were performed in selected areas. Some of the areas were selected on the basis of reports of problems at nuclear construction sites not on the Commonwealth Edison Company system. Some of the design areas covered with special " analysis type" audits were pipe supports, whip restraints, environmental qualification of electrical equipment and cable tray loading. These special design audits were performed by senior Commonwealth Edison Company Quality Assurance engineering 4 people with design experience in the related area. In addition, audits were performed of ~ architect engineer organizations which are located at the Byron Site (Exhibit M). In addition to the basic Quality Assurance 10CFR50, Appendix B Criteria requirements and the verification of implementa-tion of the Quality Assurance procedures and controls, the scope of 1 design audits of our architect engineering organizations (Exhibit N)

'T Page 23 by Commonwealth Edison also covered such items as: (1) technical review and evaluation of calculations; (2) performance of work; (3) computer program validation and usage; (4) piping design and analysis; (5) controlling'and processing design changes; (6) stress analysis; (7) design support activities by other design organiza-tions; (8) verification of qualification and ability of design and engineering personnel; (9) incorporation of design criteria and parameters into design documents; (10) design interface and data control with other design organizations; (11) technical aspects of electrical, mechanical and structural design including instrumen-tation and control; (12) design of supports, hangers, snubbers and restraints including applications of loadings; (13) seismic and environmental qualification requirements; (14) effectiveness of l design control and design review including data verification; and (15) overall design review, all to assure such items are being performed correctly by the Architect Engineers. A sample of some of the technical questions from our design audit checklists, which ranged from ten to as many as 233 questions l per audit, are listed in Exhibit A and are provided to demonstrate i the comprehensive, complex and in-depth nature to which these audits were performed to review and analyze the adequacy and accuracy of the designs. i As can be determined from the listing of items covered above and the sample questions from our design audit checklists listed in Exhibit A, some of the design items being checked are highly

~ l i Page 24 technical and require the use of an engineer with substantial design 1 experience. Thus, when it was deemed necessary to augment our audit teams' technical capability and credibility, independent design personnel with technical expertise in a particular design activity were assigned to the audit team as consultants. Examples of technicel items for which consultants were used to assist audit teams involving design were (1) evaluation of calculations; (2) design of hange,rs and snubbers; (3) computer program validation; (4) applica-tion of loadings, etc. The technical audits performed by Commonwealth Edison Company with outside consultants has proven to be very effec- ~ tive in identifying problems and verifying design acceptability. D. On-Site Quality Assurance The Commonwealth Edison construction Quality Assurance group l l at Byron monitors quality related construction activities through l audits and surveillances. The staffing of the group has increased as j construction activity has increased to its present number of approximately twenty-eight (28). This number includes qualified technical personnel from the Independent Testing Agency who supplement the activities performed by the twenty (20) Commonwealth 1 Edison Quality Assurance Engineers and Quality Assurance Inspectors. This site group is composed of a site QA Superintendent, and two (2) QA Supervisors, with the remainder of the personnel assigned under i several lead people who report to the supervisors. In addition to l the site Construction Quality Assurance group, an Operations Quality Assurance group is a1so in place. Prior to receipt of the operating r

Page 25 license this group consisting of 7-8 qualified Quality Assurance people under the. direction of a Station Quality Assurance Supervisor, is continually involved in performing surveillances and audits of the preoperational test program leading up to plant start-up as well as the other operating station activities covered by the corporate Quality Assurance Program. During the course of the project, the site construction QA organization has performed extensive audits of the contractors performing safety related work at Byron. In accordance with approved audit schedules and to approved checklists. The scope of the audits conducted through the course of the project cover not only the eighteen criteria of 10CFR50, Appendix B such as design control, document control, inspections, tests, personnel qualification, corrective action and audits, but other areas such as ANSI Standards, ASME Code, applicable NRC and other governmental regulations, the requirements of the Commonwealth Edison Quality Assurance Program and the respective contractors' Qualicy Assurance Programs, procedures, and plant installation adherence to specifications, drawings and procedures requirements. As of December 31, 1982 the site QA group has performed 409 audits of contractor activities at an average interval of each three months (see Exhibit D). Where additional coverage was warranted, the frequency of audits was increased. The audit process provides for identification of problems, obtaining appropriate corrective action including action to prevent

t Page 26 recurrence, f llow-up and close-out by the Quality Assurance groups. l ~ This audit program has been effectively executed during the course of the project.and has been instrumental in the identification of problems at Byron. In addition to audits, the Commonwealth Edison site Quality Assurance performs regular surveillances of construction activities. Surveillances are less formal than audits in that they are generally not c,onducted in accordance with approved checklists but instead are conducted to verify adherance to procedures, codes, standards, etc. Documented surveillances are performed at a greater frequency and independent of audits although they cover the same types of activities as audits. Deficiencies identified during surveillances must be corrected.in a timely manner or they will be elevated to audit deficiencies. To date 4107 surveillances have been performed by the site QA group (see Exhibit E). Each site contractor also performs audits and inspections of their work activities in accordance'with their own Quality Assurance Programs which are approved by Commonwealth Edison Company. To date the' site contractors have performed approximately 1080 audits of ~ l their construction and quality control inspection activities (see Exhibit G). The performance of such activities are monitored by Commonwealth Edison Quality Assurance to assure adequate implementation by each ' contractor. Contractor Quality Control (QC) inspection personnel are required to be qualified in accordance with ANSI N45.2.6. This requirement has been in place since the beginning 3 i -r,---, ,--a.----,,,.mr-- - - ~. .-----n.*,;.. -n.,-.., ~. -,,,, ,w,-, -,e- ,, - -n ,,,. n ,,,,--,-,-,-,n,.,,n- ..-..,-.,--en,an--

Page 27 of the project and has been monitored by Commonwealth Edison. Due to changes to interpretation to the ANSI standard, it was required to periodically review and upgrade site contractors inspector qualification requirements mainly after Til. In addition to the audits and surveillances conducted by Commonwealth Edison and site contractor Quality Assurance groups, the Independent Testing Agency performs over-inspections of construction activities. The Independent Testing Agency has performed, on a sampling basis, over-inspections of welds previously inspected by contractor QC inspectors as a check of the inspection and to assure j compliance to requirements. This group has also performed surveillances of many contractor activities in the Structural, Mechanical and Electrical disciplines. Since September 1982, another group consisting of 5-7 Independent Testing Agency qualified inspection personnel has been established. This group known as the " Unit Concept Inspection" team conducts in-depth inspections of work performed by site contractors to the basic design drawings. The unit concept inspections are in addition to the normally assigned inspections and the specific overinspections performed by the testing l agency as directed by site Quality Assurance. The inspection performed by this group is an independent reinspection of work and inspections performed by the contractors during plant construction and a verification of the quality of manufactured equipment installed in the plant. The purpose of the Unit Concep t Inspection is to perform complete check and reinspection 9 -n --,=v-r-en.,-c -n, .----.r-----nan-, ,,y--nn~-.w.~-- ~~aw-,,- ,.---,ww. w,-

~ Page 28 1 of all items of an element of the plant each week, such as all items in a volumetric space within four columns and the floor and ceiling of the plant, a specific item, etc. against the final basic design and vendor drawings to provide the Commonwealth Edison Company an additional level of assurance of the construction quality of the facilities and equipment in the plant.at the essentially completed

tage of installation.

In addition to the audits performed by the Site and Station Quality Assurance groups at Byron, extensive and comprehensive audits are performed by General Office Quality Assurance using qualified Quality Assurance auditors not directly involved with the Byron Project quality assurance activities. These off-site audits are performed two or three times each year. Such audits provide an independent check of the Quality Assurance implementation of all organizations on site as well as provide an evaluation and assessment of the adequacy and effectiveness of the Commonwealth Edison Quality Assurance groups at Byron for the Manager Quality Assurance and the other involved top management of the Company. Also, a management audit is performed every other. year of l the implementation of the Quality Assurance Program and other requirements, by an independent outside the Company consultant organization, covering all nuclear associated activities in Commonwealth Edison all of which provides additional assurance that the construction quality is good. ~ g

Page 29 As a result of the most recent Management Audit and Assessment perfo.rmed in 1981, it was the assessment of the Energy Incorporated Audit team that Commonwealth Edison Company's Quality Assurance Department and the other Edison organizations and contractors, demonstrated a high level of capability and had a good attitude toward implementing the Edison Quality Assurance Program. This Program was comprehensive and therefore, more than just adequate. It was current with resp'ect to the applicable regulatory requirements and codes and effective as evidenced by its implementation such as: the timely responses and corrective actions taken by the audited organizations, as well as the recognition and understanding of the program requirements by personnel outside the Quality Assurance Department. E. QUALITY ASSURANCE COVERAGE OF ON-SITE DESIGN l Three organizations, Sargent & Lundy, Westinghouse and Nuclear Power Services perform design activities at Byron Site under l the direction of Commonwealth Edison Company Project Engineering l Department. Sargent & Lundy, our principal Architect / Engineer, is involved on-site in piping support design, electrical design, resolution of field problems and gathering of field information to be used by their Chicago office in design activities. On-site design activities are performed utilizing engineering standards and hand calculations. Resulting changes to design documents are controlled through the use of field change requests, engineering change notices and subsequent drawing revisions. 1

~ ~ A / j,l' ' ' ' Page 30 _gm Nuclear Power Services (NPS) is basically an engineering ) adjunctto Sargent & Lundy's piping and support design activities on site. Work procedures developed by NPS are reviewed by Sargent & Lundy to ensure they have adequate design and design control provi-n 'sions. Sargent & Lundy maintains a(, vigilance over NPS activities through on-site management, site visits and periodic audits. NPS activities include field walk downs, analysis of insta,11ed piping, design of piping supports and issuance of approved design drawings. To accomplish their design, NPS utilizes standards and hand calculations, or if applicable, computer analysis is performed. "STRUDL" and[" BASE PM re the computer programs WJ widely used to analyze complex structures and resolve complex problems. The piping analysis and piping support design work being l i performed by Westinghouse is a continuation of the design effort initially performed by Sargent & Lundy. Westinghouse's site l activities also include resolution of field problems. The piping analysis performed by Westinghouse is done using automatic interactive graphics piping analysis methods. Piping models are developed from fa'brication isometric drawings, checked by l an independent engineer and then transmitted to computers in ~ Pittsburgh for analysis. The results of this analysis provide piping stress levels and piping support loads to the support designers for i incorporation into their analysis. ,3b e piping supports previously designed by Sargent & Lundy are verified to assure they are capable lp. N;,y ~ J'! l l

O g/ Page 31 ,/' of supporting the loads generated by the piping analysis [j If so, the drawings are released for fabrication. If not, the piping support drawings are modified to reflect needed design changes. In the event that new. support designs are required, these are made by the piping design group and original support drawings are prepared, reviewed, approved and distributed. In the design of new supports or resolution of field problems, simple calculations are performed, where applicable, by hand with computer analysis utilized on the more complex structural frames and baseplates subjected to multidirectional loading. ~ To augment Westinghouse's manpower, NPS personnel aid in performing design tasks. Westinghouse maintains overall responsibility for the work performed by NPS personnel, and all the work is performed in accordance with the approved Westinghouse QA Program and applicable procedures. On-site design activities are delineated in appropriate interface documents, instructions and procedures. These interface documents are reviewed and approved by the involved engineering organizations including Commonwealth Edison Company Project Engineering Department. Procedures are reviewed and approved by the applicable engineering organizations and Commonwealth Edison Company Site QA. At the Byron Site, design activities performed on site are being reviewed and monitored by site Quality Assurance on a continuing basis by scheduled surveillances; also, design activities

~ Page 32 are being audited to formally approved checklists in accordance with an approved audit schedule. Through the end of 1982, sixteen (16) audits and eighty-four (84) surveillances have been conducted by site Quality Assurance. These audits have contained numerous questions covering the following general areas: Organizations Indoctrination and Training Design Control including Design Changes Control and Approval of Design Inputs Design Interfaces Design Status Control Design Reviews and Verification Reviews of Design Calculations Document Control Instructions, Procedures and Drawings Corrective Action Records Audits Specific design areas examined during these audits.have included such elements as: 1. Review, approval and updating of design input by the responsible design organization. 2. Documentation of design input, objectives and assump-tions used on specific calculations. 3. Verific'ation that field walkdown information and calcu-lations are correct. +- ,S..

Page 33 4. Verification that penetrations and attachments on safety-rel.ated walls have been properly documented. 5. Verification that design calculations are being performed to acceptable methods and results conform to acceptable criteria. 6. Verification that approved as-built conditions are supported by design calculations consistent with design criteria. 7. Verification that systems have been checked for thermal expansion using the appropriate flexibility check rules. 8. Verification that supports and restraints are designed I to an emergency loading condition and that the emergency condition includes: (a) pipe weight, (b) self excitation l l emergency dynamic load, (c) weight of restraint, (d) l l thermal expansion pipe loads and (e) frictional loads. 9 Verification that resultant deflection or deformation of the aux steel piping anchor due to weight, thermal and dynamic loads does not exceed.25 inch at the centerline of the pipe.

10. Verification that only current design standards and l

design drawings are used during the design process. l l These audits and surveillances have been effective at identifying and resolving deficiencies promptly with the involved site design organization. While there have been minor problems noted and resolved, these audits and l l l

~ ? age 34 surveillances have shown that no significant design - problems exist in design activities at the Byron Site. These actions and results add to our confidence that Byron Unit #1 design is acceptable. F. NRC Inspections The NRC has performed many inspections of the Byron Project. a Since 1974 approximately 120 inspections have been performed of essen,tially ASME Code and safety-related construqtion and design activities. The scope of the inspections concentrated on implementation of both Commonwealth Edison and site contractor Quality Assurance programs relative to procedures, performance of inspections, qualifications of inspection personnel, and documentation verifying acceptance of installations and materials in the plant. During the course of these inspections, approximately 208 items have been identified by NRC inspectors; approximately 123 infractions and 85 unresolved items (Exhibit 0). The items identified are tracked.by Commonwealth Edison site Quality Assurance to verify completion of corrective actions. The corrective actions in many cases have resulted in improvements to the quality, programs being implemented at Byron through more comprehensive procedures, additional audit and surveillance activities and more detailed inspections, for example. Presently some items remain open, though for most NRC items the corrective actions have been completed. It is anticipated that most of the open items will be closed shortly.- v,. , -,~., _.., ,n-. ? - -. - - ---.r. , - - -., - - - - -,. ~ v

Page 35 G. CONCLUSIONS To the best of my knowledge and belief and based on the overall results of: (1) the comprehensive auditing and surveillance activities including resultant corrective actions of deficiencies I covering design and construction; (2) the quality control inspections and audits performed by each site contractor; (3) the additional assurance inspections of the contractors performed by the Independent Testing Contractor at Byron under the control and direction of Commonwealth Edison Company site Quality Assurance; (4) the comprehensive inspections performed by NRC; and (5) the various ASME surveys of ASME Code certified contractors and Commonwealth Edison l Company that were performed during the construction of the plant, involving both design and construction, Commonwealth Edison Company's l Quality Assurance is quite confident the ASME Code-related and l l l safety-related aspects of Byron Nuclear Power Statica Unit 1 are j acceptably designed and built. l Shewski l Manager of Quality Assurance t 1779Q W ..~,,,~.._-.._--._,,...-..,,,,-_,----,.y ..e,y.--- c -,,,, _,. -,,,..,,, _ _ -.,, ,,,._,~--o

BYRON STATION STATDIENT ON ADEQUACY OF DESIGN AND CONSTRUCTION INDEX OF EXHIBITS

A.

Design Audits - Sample Questions B. Corporate QA Audits - (Edison Departments) C. Stop Work Actions D. Site QA Audits E. Site QA Surveillances F. Inspections - Independent Testing Agency G. Audits by Site Contractors H. Corporate QA Audits of Byron I. Quality Assurance Coverage J. Quality Assurance - Interface Flow Diagram K. _ AS\\fE Surveys of Edison L. Management Audits - By Independent Consultants M. Audits of Site Architect Engineers Corporate QA Audits of Sargent & Lundy K.. NRC Inspection 0. P. Independent Assessment of Design by Energy. l Incorpoated I i I 9 m. .---..r. y

March 29, 1983 EXHIBIT A BYRON PROJECT COMMONWEALTH EDISON DESIGN ANALYSIS AUDITS SAMPLE QUESTIONS NUCLEAR POWER SERVICES AUDIT June 4, 1982 Question 1 Verify that supports and restraints are designed to an upset loading condition and that the upset condition in cludes : pipe weight load a. e b. upset pipe dynamic load c. self excitation upset dynamic load d. weight of restraint structu.re thermal expansion pipe loads (if applicable) e. f. frictional load (if applicable) (Ref: Consultant Spec. #109, Section 302.12) Question 3 Verify that resultant deflection or deformation of the auxiliary steel piping anchor due to weight, thermal and dynamic loads does not exceed.25 inch at the centerline of the pipe. l (Ref:. Consultant Spe.c. #109, Section 302.12)

o i

SARGENT & LUNDY June 11, 1982 { Question 9 Verify that for each amendment or revision to the l Safety Analysis Report an abbreviated internal div.ision of responsibility has been prepared. l (Procedure GQ-3.05, Rev. 3, para. 4.0, pg. 6) f Question 64 Do the mechanical piping and design groups interface so that when revisions are made to piping, the cor-l responding hangers are also revised.. Is the field notified of such upcoming hanger chadges? (Criterion III)

EXHIBIT A Page 2 Question 82 Have all non-conforming cylinder breaks for Byron / Braidwood been acounted for in the mix design. (Criterion III) Question 84 Verify that electrical valve operators located inside the containment have been tested in accordance with IEEE 382-1972. (FSAR pg. 8.1-12, para. 8 1.13) Question 85 Verify that calculations were performed to demonstrate that there will be no failure of cable insulation of the bottom layers of cable due to compacting (plastic flow) over to design life of the plant. (FSAR pg. 8.3-11, para. 8.3.1.4.1.1) Question 86 Verify that the electrical penetrations for Division 11 cables are physically separated from Division 12 cables. (FSAR pg. 8.3-12, para. 8.3.1.4.1.1) Question 87 Verify that the two groups of electrical penetrations o.f the same ESF division are spaced approximately 40 feet apart. (FSAR' pg. 8.3-12, para. 8.3.1.4.1.1) Question 88 Verify that the electrical penetrations meet the design requirements of IEEE 317-1972. (FSAR pg. 8.3-13, para. 8.3.1.4.1.1) Question 89 Verify'that nuclear instrumentation signal (triaxial) cables are run in steel conduit that is limited to 25% fill. (FSAR pg. 8.3-18, para. 8.3 1.4.2) Question 90 Vehify that for nuclear insErumentation signal (triaxial) cables the following criteria has been met: (FSAR pg. 8.3-18 para. 8.3.1.4.2.2) a. Tha rax16um straight run of conduit is 100 feet, or l b. where a 900 bend occurs that straight run is 10 feet maxiuum. Question 91 Verify that one battery charger for DC Bus 111 is rated to supply its associated DC loa'ds while fully recharging the battery within :a 24 hr. period. (FSAR pg. 8.3-24, para. 8.3.2.1.1) j l 9

EXHIBIT A ~ Page 3

i Question 92 Verify that the 125-VDC batteries are sized to carry the loads shown in Table 8.3-5 for the indicated time period. (FSAR pg. 8.3-24, para. 8.3.2.1.1)

Question 93 Verify that redundancy within the inverter equipment ~ ~ is established by a normal 480-volt, 3 phase, 60Hz input feed from a Class IE motor-control center, and a reserve 125-VDC feed from the respective ESF division's 125-VDC distribution bus. (FSAR pg. 8.3-7, para. 8.3.1.2) Question 94 Verify that a. reserve d-c feeder is provided to each ESF control bus so that, in the event of a d-c feed cable failure, the control power can be manually transferred to the reserve feeder. (FSAR pg. 8.3-26 para. 8.3.2.2) guestion 95 Verify that control power for each 4160-volt ESF bus is taken from the ' respective ESF division's 125-VDC distribution bus. (FSAR pg. 8.3-4, para. 8.3.1.1.1) Question 96 Verify that power circuits fed by the 4160-volt switchgear are protected from phase-to-phase and phase-to-ground faults by relays within the swit chgear. (FSAR pg. 8.3-4, para. 8.3.1.1.1) Question 97 Verify that the 4160 volt ESF switchgear has been adequately sized to serve the safety a-c loads. (FSAR pg. 8.3-3 para. 8.3.1.1.1) Question 98 Verify that a diesel generator is automatically started by any one of the afollowing: a. Automatic safety injection signal b. Manual safety injection actuation by push buttons,'or undervoltage on the 4160 volt ESF bus served by c. .the diesel generator. Question 99~ Verify that physical separation of redundant Class 1E circuits and devices has been pr.ovided within the Main Control Boards. (FSAR 8 3-20 para. 8.3.1.4.3 L IEEE 420-1973' para. 4.2.1)

i l EXHIBIT A Page 4 l { i 4 i Question 100 Verify that the fire suppression for the upper } cable spreading room consists of: (Fire Protecion i Report pg. AS.2-3) a. Automatic Halon 1301 system l I b. Manual carbon dioxide system, and c. Manual water hose stations j { 4 Question 101 Verify that fire stops have been provided whenever f cables penetrate fire barrier walls or floors. (Fire Protection Report pg. A5 2-3) i Question 102 Verify that ionization detectors are provided in-the battery room which ' annunciate and alarm locally l and in the control room. l (Fire Protection Report pg. 2.3-57) l j Question 103 Verify that motor assemblies located inside the t containment have been qualified in accordance with j IEEE 334-1971. (FSAR pg. 8.1-12, para. 8.1.7) Which specs need this requirement? f l a. b. Did the specs include this requirement?, Did SLL verify that the requirements were met [ c. j t Question 104 Verify that Class IE cables comply with the testing i requirements of IEEE 383-1974. j (FSAR pg. 8. 3-14, para. 8.,3.1.4.1.2) Which specs need this requirement? a. i b. Did the specs include this requirement? ~ l c. Did S&L verify that the requirements were met. Que'stion 109 For piping analysis, verify that the modal response of' the piping system in each direction are combined by the double sum method. (BY/BR FSAR + Sec. 3.7.3.3.2, 3.7.3.7 and 3.7.3.6) t Question 110 For piping analysis, verify that the stresses are determined by 'taking the square root of the - sum of the squares of the individual responses i in the three directions. (BY/BR FSAR Sec. 3.7.3,.8.2, 3.3.7.3.7 and 3 7.3 6) i

S EXHIBIT A i Page 5 j f t j Question 111 Verify, that when equipment or components are i supported at points at different elevations, the envelope of these elevation response spectra is used for the seismic qualification of the equipment. (BY/BR FSAR Sec. 3.7.3.9)' Question 112 Verify for piping analysis, that the torsional effects of ~ eccentric masses such as valves and valve operators are_ accounted for.by the following modeling techniq)ue: l (BY/BR FSAR Sec. 3.7.3.11 t a. Items are modeled as massless members with the f mass of the item concentrated at the center i of gravity. r b. A rigid member is modeled connecting the center i of gravity to the piping. i ~ Question 113 Verify that ASME Section III Classs 1 piping has i been analyzed for loading combinations shown in the table 3.9-11 in the BY/BR FSAR. (BY/BR FSAR Sec. 3 9.3.1.1) Question 114 Verify that the stresses developed in ASME Section III Class 1 piping ~ do not exceed the values listed i in BY/BR FSAR table 3.9-12 (NB-3600). (BY/BR FSAR Sec. 3.9.1.2.1?) Question 115 Verify that design loading combinations used for ASHE Section III Class 2 & 3 components and 4 supports are in accordancehrith those combinations listed in BY/BR FSAR Table 3.9-5. t (BY/BR FSAR Sec. 3.9.3.1.1) i Question 116 Verify that th.e stress levels for ASME Class 2 & 3-active and inactive valves are in compliance with' the stress criteria given in BY/BR FSAR Table 3 9-9 (BY/BR FSAR Sec. 3.9.3.1.2.2) r Verify that, for active valves, a seismic anal of extended structures (i.e. valve operators) ysis l Question 117 is performed and that the resulting stresses and j deflections demonstrate op(erability and structural integrity of the valve. BY/BR FSAR - Sec. 3.9.3.2.2) i 'i w e --s- -n,-- -+.,,---n-,~-..,- --w, w- +..- -,, - -, -.,, ~ ,.e,.,,-,-,..,,.-,,w ,, ~

9 EXHIBIT A Page 6 Question 118 Verify that, if the natural frequency of a valve operator is less than 33Hz, the operator stiffness is included in the pi ing analysis. (10CFR50 App. E, Criterion III Question 119 Verify that the loading combinations for structural steel elastic design are in accordance with BY/BR FSAR Table 3.8-9 (BY/BR FSAR Sec. 3.8.4.3) Question 120 Verify that the loading combinations for reinforced concrete ultimate strength design are in accordance-with BY/BR FSAR Table 3.8-10. (BY/BR FSAR Sec. 3.8.4.3) Ouestion 121 Verifv that stresses and strains for reinforced concrete design are in accordance with the i ultimate strength design provisions of ACI-318-71. 1 (BY/BR FSAR Sec. 3.8.4.5.1)- Question 122 Verify that stresses and strains for structural

s. teel design are in accordance with the 1969 l

AISC Specification except for specified abnormal i loading conditions where 1.6 times the AISC allowable loads are used. (BY/BR FSAR Sec. 8.4 5 2) Question 123 Verify that the AE4E Class 2 & 3 piping (except piping containment penetrations) breaks are postulated at at the following locations: (BY/BR FSAR Sec. 3.6.2.1.2.1.1); a. At terminal ends of the run b. At each location where the stresses under the loadings resulting from normal and upset plant conditions and an OBE event as calculated by. equations (9) and (10) in Paragraph NC-3652 of i ASME Section III exceed 0.8 (1.2Sh + Sa)- i NUCLEAR POWER SERVICES AUDIT September 20, 1982 j Question 3 Verify that Class C systems with an o'perating te=perature greater than 1500F have ;been checked for thermal expansion using the appropriate flexibility check rule. (Ref: Bid Spec.109, pages 25-29) e .g,,.

EXHIBIT A i Page 7 work is to the latest revision of BY/BR/pansion anchor Verify that design input for concrete ex i Question 4 CEA. 2 (Ref: Bid Spec.109 std. specification for CEA work, l page 1) 1 Question-5 Are small piping systems designed using the seismically qualified small piping restraint rule for horizontal runs with axial restraint? i j (Ref: Spec.109, page 16 of 62, Small. Piping and i i l Tubing Procedure) 1 l Question 9 Is the piping subsystem field verified for conformance to isometric routing, line number, valve identifica-i tion (s) and locations of previous supports installed i and for selection of' preliminary support locations? t I (Ref: Work Procedure 3.0.4, Section 4.4.1) 4 i j WESTINGHOUSE ELECTRIC. CORPORATION AUDIT j February 15, 1983 1 i Question 13 Are "as-built" drawings incorporated into the pipe j support loading analysis? (Ref: WCAP 9625, Secticn - 2.5.1.3) { l Question 14 Has a Design Report been prepared to show that applicable stress limitations are satisfied with the 4 component is subject to the loading conditions specified in the Design Specifications and ASiE Section ' III? (Ref: AS4E Section III Summer 1982 addenda l 4 NCA 3350, WCAP 9625,~2.5),,. i i Question 16 Do Load Capacity Data Sheets state the load capacity of i the component support and identify:the tests and calculations used to establish the. load capacity? i (Ref: ASME Section III Summer 1982 Addenda NCA 3551.20 l SARGENT & LUNDY AUDIT i March 11, 1983 Can hand' prepared design _he purpose or objective be 4 calculations be easily-Question 8 reconstructed (i.e. can t r determined, data and assumptions listed, all calculation steps recorded, and source (s) of formulae i or equations not in common-engineering usage referenced)? Ref: Procedure _ GQ-3.08 Rev. 4 Para. 3.0.B.1-Criterion III l 4 n ,, > ~ ~, e- -,e w er--r-se r-

EXHIBIT'A ~ -Page 8 j 1 4 Question.10 Do the design criteria and procurement specifications contain the requirements for environmental design and i qualification to assure the capability of performing y design safety functions under all normal, abnormal,- [ i accident, and post accident environments and for the l length of time for which i'ts function is required? -l P Ref: Procedure GQ-3.04 Rev. 6 & GQ-4.01 Rev. 9 NUREG-0800 l (formally NUREG-75/087) { l Section 3.11 Subsection II Criteria III & IV 1 l Question 14 Is a review of computer-aided design calculations made to determine if input data is consistent with design i j input, it correctly defines the problem for the program, and is sufficiently accurate to produce ) results within the numerical limitation of the program? f i Ref: Procedure GQ-3.08 Rev. 4 t Para 3.0B.2 Criterion III i 1 I } Question 16 Verify that AS4E Design Specifications have been certified to be correct and complete and to be in i J i compliance with the requirements of NA-3250. [ 6 i j Ref: Criteris III & XVII 2 l Question 40 Verify S & L reviews as-built drawings and reconciles [ l those ' modifications with the stress report i calculations. .a. Ref: 10CFR50 Appendix B Criterion III l j Question 41 Verify that S & L is re-analyzing' structural integrity l as complete support as-built data is furnished by. J Westinghouse and NPS. l i l Ref: 10CFR50 Appendix B Criterion III Question 42 How does S & L incorporate final loads _ in their-structural ' analysis? r Ref: 10CFR50 Appendix B Criterion III i i 9 6 s -, _ ~,. _

i EXHIBIT A I Page 9 a l 1 l Question-43 How is support as-built information: transmitted from i i Westinghouse L NPS to S & L~for final structural } analysis, and once received how is it processed? i i j Ref-10CFR50 Appendix B Criterion III h j Question 69 Verify that as-built information has been - evaluated for system safety function. (Ref.10CFR50, Appendix B Criterion III) t j Question 72 Verify that R4D has identified high. energy ( lines based on the following: a.' Regulatory Guide 1.46 l b. Byron /Braidwood FSAR Section 3.6.1.1.1. i l3 (Ref.: Regulatory Guide 1.46) l l (Byron /Braidwood FSAR Section 3.6.1) l (SLL PI-BB-38) i Question 73 Verify that EMD checks high energy lines determined by PMD for postulated breaks using the criteria of j Regulatory Guide 1.46 and Byron /Braidwood FSAR i Section 3.6.21 for the following areas: 1 e 4 a. Reactor coolant loop piping j i b. High Energy Piping other than Reactor l Coolant Loop Piping. l f (Ref.: Regulatory Guide le46) (Byron /Braidwood FSAR Section 3 6.2.1) } (SLL PI-BB-38) i Question 74 Verify that piping in the containment { penetration ar'ea meets the requirements of i J FSAR Section 3.6.2.1.2.1.2.2 for "no break" conditions. ^ (Ref.: Byron /Brdwd FSAR Section 3.6.2.1.2.1.2.2) l Question 75 Verify that pipe whip restraint design includes a check of ring to pipe gap in I the cold position. l (Ref.: 10CFR50' Appendix ~B Criterion'III) (Byron /Braidwood FSAR Section 3.6.2.2.2.2) i

O' EXHIBIT A' Page 10 Question 76 Verify that the design of pipe whip restraints in'cludes a check of the pipe to ring gap in the hot position using the.most significant thermal movements from the piping analysis. -(Ref.: 10CFR50 Appendix B, Criterion III) (Byron /Braidwood.FSAR Section 3.6. 2.3.1.2) Question 77 Verify.that the blowdown forces on primary [ loop components are computed from equation 3.6.1 in the Byron /Braidwood FSAR. (Re f. : Byron /Braidwood FSAR Section 3.6.2.2.1.2) Question 78 Verify that the impinging jet force for piping other than reactor coolant piping is determined in accordance with Section 3. 6. 2. 2. 2.1.4 o f the Byron /Braidwood FSAR. (Ref.: Byron /Braidwood FSAR Section 3.6.2.2.2.1.4) Question 79 For blowdown calculations, outside the reactor coolant boundary, were the most significant ressure-temperature conditions utilized by EMD? p(Ref. : 10CFR50 Appendix B, Criterion III) t Ouestion 80 Verify that the structural design of reactor coolant whip restraints is adequate for the l postulated load based upon the values for working i stress allowable stresses in the 1969 AISC speci-fication increased by 50%. i (Ref.: Byron /Braidwood FSAR Section 3.6.2.3.1.1.3) Question 81 Verify that the structural. design of whip restraints (for other than R. C. lines) is adequate for the postulated loads based upon the criteria listed in Sections 3.6.2.3.2.4 and 3.6.2.3.2.5 of the Byron /Braidwood FSAR. [ (Ref.: Byron /Braidwood FSAR Section 3.6.2.3.2.4 ~ and 3.'6.2.3.2.5) Question 82 Do the embedments and supporting structures have sufficient strength to accommodate the design forces of the whip restraints as wellias any non-pipe whip related loads? (Re f. : 10CFR50 Appendix B, Criterion' III) (Byron /Braidwood FSAR Section 3 6.1.:3) i ? I b 9

t EXHIBIT A j Page 11 Question 83 Verify that structural-items important to safety that may be impacted by unrestrained pipe whips have sufficient strength to with-J stand the impact loading. (Ref.: 10CFR50 Appendix B, Criterion III) (Byron /Braidwood FSAR Section 3 6.1.3) (Regulatory Guide 1.46 Section 4.6) Question 84 Verify that the design of pipe whip restraints has 1 included the proper. selection of energy absorbing materials in accordance with established guidelines l based on testing. (Ref.: 10CFR50 Appendix B, Criterion III) Question 85 Verify that an analysis of the performance of absorbing material has been made considering both the solid state and plastic deformation modes. (Ref.: 10CFR50 Appendix B, Criterion 'III) e Question 86 Verify that when primary loading forces are applied i to energy absorbing material in.a direction other than normal to the primary face, the effect of the j angle of the force has been accounted for in the determination of the energy absorbing material performance. (Ref.: 10CFR50 Appendix B, Criterion III) Question 87 Verify that criteria for acceptance of l damaged energy absorbing material has been i justified by test or, analysis. (Ref.: 10CFR50 Appendix B,i. Criterion III) Question 88 Verify that field installation tolerances for i i pipe whip restraints have been justified based upon the original design. (Re f. : 10CFR50 Appendix B, Criterion III) l l Question 89 Are "as-built" conditions reviewed and analyzed by S&L and reconciled. with the original whip restraint design. (Re f. : 10CFR50 Appendix B, Criterion III) l Question 90 Verify that design calculations for whip i restraints are' properly documented in accordance with S&L Standards. l (Ref.: 10CFR50 Appendix B, Criterion III) l (Sargent & Lundy GQ 3.08 Section 3) i i -e.. ,. -, - -. + - e--,, ,, - ~,

t EXHIBIT'A Page 12 / Question 91 Verify that drawings and calculations have re'ceived independent and documented review. ) (Ref.: 10CFR50 Append'ix B, Criterion III) (Sargent & Lundy GQ 3.08 Section 3B.2) Question 92 Verify that procedures have been written and-established for the interdepartment control of whip restraint design. (Ref.: 10CFR50 Appendix B, Criterion V) [ (Sargent & Lundy GQ 5 01 Section 3) Question 93 Verify that design changes for ship restraints I are subject to the same control, review and approval as the original design. (Ref.: 10CFR50 Appendix B, Criterion III) (Sargent & Lundy GQ 3.08 Section 4) ^ r f I i i j. t i l 4 I l i l ] l e , ~, _,

.{- . EXHIBIT.B i o t I CORPORATE QUALITY ASSURANCE DEPAR*IMENT AUDITS OF COMMONWEALTH EDISON DEPAR*IMENTS CECO Departments Deficiencies Identified In the Annual Audit-i 1974 1975 1976 1977 1978 1979. -1980 1981. 1982 Total SNED/ PED 17 11 14 8 15 4 1 2 1

73 (Engineering) i OAD 5

5 1 7 4~ 5 0 1 1 29 (Testing) ,1 17.. } Purchasing 7 2 1 2 1-3 0 0 t Nuclear 0 0 0 Licensing

2 3

Total 29 18 16 17 20 .12 1 3 3 119 4 W-9

[. EXHIBIT C BYRON STATION STOP WORK ACTIONS-INITIATED -BY CQ4MONWEALTH EDISON COMPANY 1. 1977 - May 6,1977 concrete placement ' stopped due to improper-placement of reinforcing. Corrective action was completed and work. resumed -5/11/77 (NRC-#2C) 2. 1977 - System Control Corporation - May 23 to June' 10.; System Control Corporation work on cable pans and hangers. was stopped as a result of ' deficiencies noted during a CECO-- QA audit of their plant activities in Iron Mountain, Michigan. 3. 1978 - Ceramic cooling tower - stopped 7/28/78 due to unsatis-factory implementatien'of QA program.- Inadequate documentation was received prior to the release of materials for construction. Work resumed 7/31/78. 4. 1979 - System Control Corporation - June 7 to June 28. Commonwealth Edison Company Station Construction Department conducted an in-plant inspection on- ~ containment isolation panels, cable pans and hanger welding. Welding was unsatisfactory and work was stopped. Work was resumed after an inspector: from Pittsburgh Testing Laboratories was assigned to the shop to inspect equipment prior to shipping. 5 1979 - Blount Bros./Inryco - stopped post tensioning on 11/27/79-after anchorhead failures. Subsequent distressing and replacement of Unit I anchorheads. Work resumed in spring of 1980. 6. 1980 - Hunter Corporation - stopped installation of safety related component supports on 3/27/80 due to design.and inspection inconsistencies. Work was resumed on 4/3/82 after procedures.were revised. 7. 1981 - Hatfield Electric - stopped all safety related cable pulling activities on January 10, 1981 due to program-matic inconsistencies. Work was restarted completely on March 16, 1981. + 8. 1981 - Johnson Controls - stopped all safety related welding activities due to procedural deficiencies. on March 6, 1981. Work was resumed on March 9,1981 after correc- } tive action was taken. ,2108Q ~ 1:.

EXHIBIT-C Page 2 r i 9. 1982. - Reliable Sheet Metal '- stopped allinew installation activity on safety related H.V.A.C. Systems and HVAC { attachments to safety related structures on 9/17/82. Stop work caused.by inadequate and -incomplete inspec-l tion and inadequate procedures. Stop work is still in affect' as of 3/30/83 I i h I I 2 [ i 4 i; d i I i l i 5 g. i t 1 i 5 I I t e 6 A i t ,2108Q - - +. -, b ,g

EXHIBIT D i SITE QUALITY ASSURANCE DEPAR'DiENT AUDITS OF - BYRON CONSTRUCTION. ACTIVITIES i hbmber of Deficiencies I Year Audits Findings Observations Totals 1974 4 0 5 5 1975 1 20 0 20 1976 37 102 99 201 l 1977 50 108 120 228 1978 58 113 83 196 4 1979 68 76 77 153 1980 '56 89 72 161 1981 66 90 i" 74 164: ~~ 1982 59 55 56 111 j TOTALS 409 653 586 1239 i i e (0222.4)/0053q -m ,w r--

EXHIBIT E i SITE QUALITY ASSURANCE. DEPARTMENT SURVEILLANCES 0F BYRON SITE' CONSTRUCTION' ACTIVITIES Year Number of Surveillances 1976 44 1977 486 1978 550 1979 576 1980 703' r 1981 945 1982 3 *- -803 Total thru 12/31/82 4107 (0222M)/0053q t-e

- EXHIBIT F ~ s BYRON STATION-INDEPENDENT TESTING AGENCY TESTS AND INSPECTIONS Test and Inspection Area Concrete Aggregate . ~ 36,500 Inspection Grout Cadwelds Reports (approx.) Concrete Expansion Anchors Bolting Welding Non-Destructive Examinations Magnetic Particle Liquid Penetrant 36,800 Inspection Radiography Reports (approx.) ~ Ultrasonid g. e (022Z4)/0053q 4 e e .,..,e -,n.,, - ~~-, --e

t J ' EXHIBIT G a BYRON STATION AUDITS BY MAJOR SITE CONTRACTORS ~ 'THRU 12/31/82 1 [ I-Contra ctor Number of Audits Hunter 88 Blount Bros. 561 . i Hatfield 97. Powers-Azco-Pope 79-Reliable Sheet Metal 7 Johnson Controls 3 PTL' 245 TOTAL 1080 $0 7 e 0 i 'h (0222M)/0053q W i '9 e e = ge, . - -.. + ,%e y

m f - EXHIBIT H l' T i CQ4MONWEALTH EDISON CORPORATE QUALITY - ASSURANCE AUDITS OF BYRON SITE T Deficiencies Number - Findings Ob servation s - E ls Year of Audits - 1 1975 1 20 6 26 i l 1976 3 26 4-30 i 1977 2 17 10 27 1978 2 Id 12 22 1979 2 13 6 19 ' 1 4 1980 2 20 5

25-1981 2

9 7 16 1982 J 23 29 52-Totals 17 138 79-217. i 5 9 O 9 --+ ..r r is. p-e.- ,,y

Quality Assurance Coverage of Commonwealth Edison Company Construction Projects Construction Sites Design Activities Other

1. Contractors 1.

Architect Engineers 1. Corporate'QAL Vendors Audits i Inspections (QC) Audits (QA) -Audits (QA) Other Other CECO Orgs. Surveillance & Audits Independent Review

System Design

Overall Design

2. CECO Quality Assurance 2.

Ceco Quality Assurance 2. QA Coordinator Engrg. G Tstng. Site QA Site Design Activities Surveillances Inspections Site QA Audits Independent Testing Surveillances (Documented) -Agency Continuously Audits 7 us< r: Surveillances (Documented) Corporate !1 Continuously (- 800/yr) Audits go., Management Audit Audits Outside Consultant Corporate- . Outside Design Activities c Audits - By Corperate QA Corporate.QA Of all' site organizations Audits of AE Organizations ManagementLAudit - (Supplemented Team with-By outside Consultants design Engineers having Special design expertise) QA Reporcs'to Commonwealth Edison Company Corporate Management

1. Copies of 'all CECO < surveillance and audit reports 'are sent-to Manager Quality - Assurance and Director.

~ Quality Assurance for. Engineering-Construction. -2. Audit-reports are sent to responsible corporate management. .3. Monthly _ composite reports are - prepared and provided to respons b e' top management by QA. il

4. Annually' an assessment activities and objectives report is provided to senior management.

mn m

Quality Assurance Surveillance Audit & Inforination Interface Flow Diagram For Comonwealth Edison Construction Projects ONT g 1tQ I f fe cio,M ICE'OihiRMAN V PRESIDENT .anm.1 sqwe N ~ 4 /\\ /\\ /\\ /\\ 3} } kY y$ 13 } / YY Y Y ~ MANAGER SMOR AND W

N ameniy Res=t=

, @ ITY N MAN /GI N N ~ x rg DEL T Q.A.l 6, 9 h ] y sinceries 1mne SITE Q.A. 00mDINNIUt n Q.A6. m& M b OEo Senior L Top Nrugenmt Interfaces and Inforiention Flow TESTIE N ' T,.eg,, Wh

MfMT L Nrwillere Reports ME
Bi-wrekly Activities Ecports M

-= MMEL Sirw111nrres 's -[ SITE [VE'NDORS l [A.E.l l C.E.Co OCNIRACIW S / } lDEFIS. -*- - - - - Inspret ima h Inspections L /udits Insmtima L halits Axtits

EXHIBIT-K-j t t l t i - h b - i i ASME SURVEY i: OF l COMMOWEALTH EDISON CORPORATE OFFICE' I Year Citations v '1974 None r 1977 None f 1980 None t O e a* e ) f l I I t e k o l 7 i l I ? - I 1 t .O m E g 1 l

e + EXHIBIT L i i s. r 4 k MANAGDfENT AUDITS OF CQ4MONWFALTH EDISON COMPANY l B3 Ton Deficiencies Ye'ar Findings Observations 1977 1 1 1979 (General Comments Only) 1981 1 7 2-8 ' Total j g. e i . 1 l L l l l t l l I i

EXHIBIT M BYRON STATION-CECO QUALITY ASSURANCE AUDITS OF SITE ARCHITECT ENGINEER ORGANIZl.TIONS~ I Number of Deficiencies Architect Engineer Audits ~ Findings Ooservations - Total Sargent and Lundy 7 3 10 13 Nuclear Power Services 8 4 11 15 Westinghouse SAMU 4 0, 3 3 Total 19 7 24 31-8 h e i k [ O i t l

EXHIBIT'N I f CECO QUALITY ASSURANCE DESIGN AUDITS Number of De'ficiencies Architect Engineer Audits Findings Observations Total ' Sargent & Lundy 15 113 38 151 Westinghouse (NSSS) 11 29 22' 51 1 Total 26 142 60 202 f r ~ h h i 5 l Page 1 of-4 l i 2108Q l

~ EXHIBIT N CQ4MONWEALTH EDISON CORPORATE QA AUDITS 0F SARGENT & LUNDY DESIGN ACTIVITIES (BYRON PROJECT) Number Date Major Areas Audited of - Ques tion 2/20/74 - Organization 136 - Design Control- - Procurement Document Control- - Instruction, Procedures & Drawings - Control of Purchased :laterial - Corrective action 1 - QA Records - Audits 2/25/75 - Organization ^ 164 - Design Control - Procurement Document Control - Instructions, Procedures & Drawings - Document Control - QA Records - Corrective Action - Audits 2/17/76 - Organization 137 - QA Program - Design Control - Procurement Document Control- - Instructions, Procedures & Drawings - Document Control - Control of Puchased Material, Equipment and Services 3/7/77 - Design Activities 132 - Approval and Control of Documents / Design 9/7/77 - Implementation of QA Program 1111 - Compliance w/10CFR50 Appendix B - Follow-up of March Audit - Design / Document Control l Pa ge 2 o f 4 ' i 2108Q

EXHIBIT N COMMONWEALTH EDISON CORPORATE QA AUDITS OF SARGENT & LUNDY DESIGN ACTIVITIES (BYRON PROJECT) Number Date Major Areas Audited of Questions 3/1/78 - Design Modifications 233 - Field Change Requests - Engineering Change Notices - Nonconformance Reports - Drawing Changes 10/30/78 CECO Field Chan 16 - Follow-up of 3/ge Request Procedure 78 Audit - Internal Audits - Audit Checklists 2/21/79 Pipe Support Design 46 . Field Change Requests - Cable Tray Loading Overall Design Review 2/27/80 - Oper. Design Change Requests 41 2/11/81 - Design Spec. Requirements (NCA-3250) 46 - Interface with Other Design Groups Cable Tray Occupancy Vendor Drawitag Control Design Review 5/11/81 Design Personnel Qualifications - and 16 Training Computer Program Validation Generic Industry Design Problems Reviewed 9/29/81 Environmental Qualification of. 11 Electrical Equipment: NU-REG.0588 1/18/82 Nonconformance Reporting 74 Field Changes "As-Buil.t" Drawing Processing Cable Pan Design Overpressurization Design Block.W'all Design Post Tensioning System Design Review and Calculations e-Page 3 of 4

EXHIBIT N CCHMONWEALTH EDISON CORPORATE QA AUDITS OF SARGENT & LUNDY DESIGN ACTIVITIES (BYRON PROJECT) Numbel Date' Major Areas Audited of Questioi 6/7/8? - Stress Analysis 146 - Mode Combination Technique Load Base Combinations Seismic Analysis (More than one Response Spectra) - Valve Modeling - Equipment Qualification - Structural Steel Design - Reinforced Concrete Design 3/7/83 - Equipment Qualifi ation 93 c - Design Criteria - Design Speciications - Revies of Calculations - Field Changes - As-Bailt Drawing Reconciliation - Design Change Control - Whip Restraint Design Page 4 of 4 l l 2108Q .}}

ML20076D144

Text

SUMMARY

System Design

Navigation menu

Search