Text

Rev 2 to Nuclear Svc Div General Reactor Vessel Setting Procedure
	ML20238A390
Person / Time
Site:	Comanche Peak
Issue date:	02/13/1979
From:	; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20237K807	List: IR 05000445/1985007; IR 05000445/1985014; IR 05000445/1986008; IR 05000446/1985011; ML20237K804; ML20237L110; ML20237L125; ML20237L136; ML20237L151; ML20237L159; ML20237L179; ML20237L203; ML20237L268; ML20237L298; ML20237L321; ML20237L367; ML20237L407; ML20237L453; ML20237L464; ML20237L493; ML20237L577; ML20237L797; ML20238A290; ML20238A390; ML20238A485; ML20238A535; ML20238A548; ML20238A573; ML20238A583; ML20238A596; ML20238A609; ML20238A642; ML20238A654; ML20238A667; ML20238A719; ML20238A749; ML20238A782; ML20238A833; ML20238A839; ML20238A868; ML20238A883; ML20238A903; ML20238A923; ML20238A966; ML20238A974; ML20238B015; ML20238B131; ML20238B210; ML20238B251; ML20238B352; ... further results
References
	PROC-790213, NUDOCS 8708200474
	Download: ML20238A390 (10)
	v • d • e

.

WESTINGHOUSE PROPRIETA Y CLASS 2 REV. (

)

. Wes11nghouse Power Systems Company "o urse.a o. e

, rn Electric Corporation u s:. pp r se w w

- February 13,1979

)'

WESTINGHOUSE NUCLEAR SERVICE DIVISION PROCEDURE FOR SETTING

.. .a OF MAJOR N 1

NSD' General Reactsr Vess'el'Settin'g Procedu're 4M ?

REF:

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. . . t - ps. ([% d

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.u

., 'V n >

. .-o ! ek rM rs h >, :.

conoonents Accuraev in <*ttina aMor ' Nuclear Steam Sucoly Sys".amanRSSS eaa"racta At j 5 the sole retnantib' ity of the cut +aamer's oract1 m tett' M i

d

+ . ,

,,,, ,ts suggested that said contractor provide a"" 2 t !

(

, b, .for nereview usebyo"Westinghouse this documenand/or JiSD site personner. ourp specialists, as cons/

_to Westinchousem ' Good erection' practices'mus,t,prevai,1. .

's .3In both orientation and elevation, the reactor'. vessel (RV)RV is elev-the key

) ' 3;g., component' to ;the setting of' all,other ~ major' NSSS '

~

8

s. . " n ' b '" '

. piping to the other.components. n y.. ./r'- f 4 . . . . nW ' .. ,$

.g,The following 'is 'a summary ':o(b,asic. requi

. i e G o v.i. 'M. ....s sequence the Iff, fuel transfer system'(FTS).RC pumps, steam gen-

.'gara, sy for' n's'tallingiHQrtz,er,;and,,RCAe tors,,f}h ig pg. u.. ggg .tna.83g {

w

.-r

...n a ..

ch; marks ';g. .

,,,,,d. levatio,n referenceh,ta'inment&f,$

. .a , .

. r.y

.',.., t ..,.u';

% PRE.,RE. u QUI,S,,ITE I, Azimuth an .emust'91 sube' . aVailable' irith ' theleh .

g pg,g.gy.g.. . p fyS.lf.9jy..

", ,O C"qt.{* 2 %'o.,

g Lay out the'th' ecrettcalg p3.'lj,, g containment p g,g axes from.the above

'4 i

)

"' * .,_' f0;'. Af&% ". . Q ' 1 indicatedbenchmarks."@Y*)Q!t&"fN_

. _ ;,:3.lc q .ei ,gy. .;

The reference points used for' sunoort' arid the' horizontal J?

. t '. gdshall be the cent'e'rlia* af the,aetting ,the

~

. c, '

9 .* * ' 9M of the ' Support.d: u, ..u n w W: nn M..

~

' < .'v, . ' bearina elatm !J a t !

I

....a

3. The setting tolerances'for theLsupports sh'all'be: u eg y(r .

. r , .

.A M. a: .nw .

M.p1.'.yp j

( - ,

.- p 's. ...

'I '

i

'C v . :.;nc .,.

. ,.i v. . . ...s. o e v

% .' q' t&:N.

y..y',. ...

The supports shall be grouted in'pg4' .

~

, J' The reactor vessel support cooling channels shall,.be ' hydrt, .

tested. .

B708200474 B70312 PDR ADOCK 05000445 IV-1-1 0 SFPbR.

WESTINGHOUSE PROPRIETARY CLASS 2 REY. 2 t.. \

Westinghouse Power Systems m e se.a y. w

] ~

Electric Corporation Company ,, ,,

Pme.':g1Fe cyt.sa a 14% *

4. Within the equipment support system's design flexibility, ,

calculate the apth RV elevatinn from the RV, RC pump casing, SG and RC pipino "as butits". This optimum elevation

,is.the mean horizontal plane required to average any variances in the equipraent nozzle elevations for maximum RC p'iping fit-up.

During this exercise, also detennine the SG and,,RC pump.' casing ,

i,..

azimuth' location' and crien'tatio6' for' the'same' reason..e Insta11' and functionally check out the neutron detector.

5. .%

4 m- ' $::

A

\.:.r.p.an positioning u s.qdevices.

w e es t %$ '[.t' .s t ~ . '

r.na:A -@ .-

l req'uired,or'at'leasPbeEeficial, 6.4 Due' tipl ant' l ayo~u't,' it"ma'

')to position"s'ome '

"the shield Wa11'/all sleeves"the'RC# prior to" vessel pipingsettiri adjacent'/V Positioning 'to"theRV *through f y of,,'the bottoin a'ounted instrume'ntation pipe 'ma ~ likewise, be j.,

deem ~ed"riec'esf ar'f.T,f*j'92f0.? "/ R8 ,

W**d?

tr' k el is %a'r$h s a b c'hecked, Ina

'I verified,and newly'aiirked if in'arkiri ffo'und 'in'c~o'rrect W.'- @'@

/ 'v. ' ' . . W '). ,2

d d .. "or nit ' visit,le'.Tt,PW '^ AM '

.l;,y' he.reacto:.rlt+:QbsW-h.:%y&

.TPAn

'circumfe all ! .f:;fs:yM e

~-

Q, .

' e c.c

-. *: ~* ' 'i !. .

vesse % "loca y

"' . .Setthe'y

.erse '

. b RV, to .

o'up

.. J num el

~$

9 W-

$T:

7. re etermine .

reference lanejfor,,, ele 's't atioh' Tt evelness'ljs M . M ?.9:1% r,' ..

ie'ma'tinti flang M

ver" plates' allow

access'tW.d[til

.' ola'fe'd,)t'e~ctih FTK ablelro i

..., .. the .supportiled e or.A terminin '. actual., ele ation~and. levelness W ',1 [.

' ' " l

dur'i 'etti'~ vei .bekithintf4!.,'.A.@.i+'

,,w ,,;p,g. ,y, .

$$kk N,

u. .v n

'drTe"nt .the ,vess'el,jsuck

-proper. s' aligned . .P

' l,.j,?paia11el't'o'the same containment theore,ma orp'xis tical,a ' r." " utilized

' for layout of .the FTS 't'ube 'and ' manip'ulatof ifane guide rail. 3:. -

f'

. Scribe or punch90', marks are located on the snain: .,'LE flange vertical

.180' .and 270' axes,' accurate to hr. .

. Thes e . marks cE ,be Napl oyed 'for:orienti theveise1.

~

N.' J]- l j

ge's f erenc

, 3 the iRV gi s be s t ., loc a ted ,,c i rcunfe ren tj al ly' n,AMthe four' barrel The kepaysAre '

,$ Mos't . future elrprotection* Nes,s,kaywaysL4 coversalso will!pos{tione p ojide' . M M

to the keyways g ,' , .. Q , g ' Q ,y / g.'f.'].7 >.

" e ' ". , .gh ;

9 1

I

.<?, -

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WESTINGHOUSE PROPRIETARY CLASS 2 REV. 2

" "'I m t05c" Westinghouse Power Systems Electric Corporation Company ,

- ei. me .~. .

P

, :,.,p.{.ffs{t*I.3'.3..a'MT; g

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)

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e) ,

. ,*e* . ., - r q ' s . 3 ', e,

- 8. Final set the vessel per the referenced " General Reactor Vessel Setting Procedure".'.~The ' vessel should b5 insulated;during 3

-the period ofitime.whef the~ support bottomihiis}are!beino 6.r.,- ~- e n g ,.] -'.? ,g 4 C.; } ,\ . .

-) N., machined. - q.J$

-(c..w .c t h M t P 'yioMh ,.. U M it%It.r,1p 16 3 'M 4 .

9.' Rough' set the SG's.(archi.tect engineer su installed within supports) and RC p~ ump 'c'asin , din cold ?

e '

CV. position, to the..elevati.on., azimuth and orientation'g5 calculated

' W .;'during determination of,:the' RV 'opti,m'u'm' aliva' tion;*'fp,,r makimum s

ab pipe fit .up.>p Axes. marked on' SG's71f {

.e.: t1 and thus , ;should only,,be,,.. mployed a,s.'.,'any," ~""

a gui '

. J h.e ,r'adial dis-nozzles'.sh)

-l' O. rtance afrom sthe ;RV .cente aid nt i

' pT

% i..,NO Rf .

r:s ep I

. f a WW.1 Mih43Cf : . .,, (i.- I

.. ; tis f.MJGeneralInforma on y g33.j -

, .: peg . ...

.) t..:v:/A 6ktiWit4htetab The. steam; generators istial pr*f6/'l'j be se corancewith.-}I, N W-T 5 pre-detennined woHi" point brJThi'se rh,oints.shallbe '

r,g$$ igdeterminedfusjng,the *js' "c ' 'onenyd,i f ensions' and .;M(..

..-;.mfy3 3.4thec pointsgshall be '

mum reactor vessel .

q ' . Consideration.must f g unt for[the elastic 4c g.4e ceflection c levationi.should be g,'.

ij'gfj'

' .5 NWctd % vu m .

t .cd'.e:/,gT.a,dju4tedfor'6the

v. a k m hi gf o.no-,

ss quired

. suppoi

~ c..sv to set NiHrd <..41 p,hc ! ?.r.ior,in.,'p'ection(of/ng d'h,a,vP soul the..suppp, e been*pe r.ul tra-a' fonned > -

'ics . w4t 50nic., s 1w.e J.y

. ~.'.%, :\.,i#::. 4, .y., .;pyp,,.:.a'nd .7 : q y y,a,ccep,te'd

.g,y ,g, - ,1*?.; 93y.Qp.

he3:E >g .

p< 'd. , To insure that"the s equa ted on qd ;,.dt-I U.1.suppor,t system, some load peasu, ring 'deji,ce,'such ,'as strain

. J,d 7. 3 - . < .. gages or some other method sh'oul'd be a reed on'by 'construc-t k.' f,$ t it$on d.[t (e yfj.(ch,[.,d #j[ j

[* '. F l'Y . h;, . .

)

~. ,.:

. : i '. . I V:w',,,; r d- m '

Q.. . 8 n p.' co[

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y. ,

y . ,

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1 J

t REV. 2 ..

WESTINGHOUSE PROPRIETARY Ct. ASS 2

10. The pressurizer may be final set at this time and is to'be ]/

vertical within one-half degree. It is advisable that the

. ', N ,

plumbness be less than one-half degree so asf to' facilitate e~

fit-up of auxiliary ilnes. ,

11. Utilizing the RV nozzles as a reference check'the 59 ano ,3 RC pump locations via instrument and adjust the equipment '

./

accordingly. The SG's must be' vertical. within one-half degree whereas the casing's main flange' must be level across ;

the OD to within.0.125". It is also advisable that the . !. l olumbness of ~the SG be less than one-half degree so as .to

~ facilitate fit-up of main steam ano feedwater piping.__. .. f ., .,

(.

reactor coolant. *. @. .y#

NOTE ' presented bel'ow is just' one 'of. s'everal . . ....., .s.w *

. pjpe'weldout'seqbences successfully employed 4to date.1 ,,,iq (.

.g..y n 3 3 e... .pp4y.uv g. :g g

<c ,

12. I Align hot'and.' cold leg pipe'.'or..vipe' assemblies(with.'stosi

ge#F.5/f; , 4 valves) between. Units and es' tallish' reference marks tto detect;.4 - .J. !

46 '. .

pipe' draw. t Win'g' proper ' welding" technique land 'sequ'encing < ' < ,i , .

,to maintairtpip,i,n in aligned ' position. attach. pipes to the

' RV*r . Elbows' can a s'o 'be , rot'ated 'slightlyto'.a(sist " E e.in alignesn dol

(.9?.2R.

Wi m .

If,. loop stop'.7v'alves%'are.d.3? t6 Dns' tilled.%r6cee@Q4 sFe$$,d 1 '

itikweldt of ' valves to pipe sections already*a'ttache'dJtd thelvesse1N '

-e and then iemaining pipe pieceltoYdlYe.'/iG t.e line* alignment be mai ne thro, ugh g y':;g.g.,n 7 '

b

of eachsi,pm,in,,the assembly.mustiously.' rev M g,;4 s . .N 9 g$. .TlH , @4P 'l;lili% ,'. E ,-

, ..i, reference f.J 1.i,;O MM,@ marks asifndicated

'?ngNfsevera CT v

..e NOTE - $1 Itaneovs we ter11ne' M C T

""i'N.[$ snyo6e71oop,'Is%ccepta le?is,

~a ali

. #'.. '3' ~ n

~

!f?gnment 3sodone'.;anaddjtional4/32"clearahce I ^ " @{ . , . .

% ;W .

fy

. I' allowed for'at the va'1v'e2 4 bao rease , ap., tween :, .

13 h nk e t d

he'5Gand s 'and pumsi,ye.,.'m 1

. w..the ,instaAdj$stthe led ' pip'.e.'ior. pipe ' asp' lies docat'idns* *i 7 l

[.dcasings. -

. j ian inwar .directiohf,equ',ipme,,aralle' To*'In' inal RV2 d al ,11ne %.'- ?

'.throdbh

@the connecting no:Ile cente O ch7t hitJthe' ' ~

is 4W #* L.M., ~v.s .? .r ,

i,y'+,c,l.earance

, .n, -:.. s . gsegeestablished.

p.stfuw M.f.hM ,,, ^

[,5 '

' '" M ;, '..

14. '.deTh'ot a'nTYo1'd$1egsNoks a'nd p'unisi"cIsings' beinp sure .

~ 'not to , draw'any major N55$ eqdipment, 'otherjthan rad' ally i 'ji,s.'.. . . . . '

"inward Qf. c ;'p. g,e.:

1,.previo,.towards the NwRV due.to,~ weld shrt,6Eige7 rom I ,

usly se ~~#

k p;ositions.'

> K:.%'<

15.y ;\yg ' . 1 dou t...of . Q- i

\

mAlignment:an rm n ead outlet

duri feciding ( l Ee11sito)5G',s}can3e:done anyg{pe]aryc j atthiti nd's' eld pre i W welding ,.i opera'tlin'sle Be' sure J

' l forms a horizontai piane. u p : y. w. r. g ., g...t,w .e .. , e e . .g

v;.7

_-.____----_-------------------.--------------------------------------N

l I

REV. 2 I WESTINGHOL. SE PROPRIETARY CLASS 2 '

l I

10. Measure horizontal and vertical RC crossover pipe clonrc J

's dimensinns tn within 1/16" and sulciit through the Westinghouse

) Nuc1 car (.onstrut.tlon Department Site Manager. SalJ dimensions '

are not to contain an allowance for wcld shrinlage.'. .The . . . il #

3 j

erector is urged to separately advise desired excess material for shrinkage, otherwisc the closures *as received" will l contain an additional icngth per Westinghouse standard.

.} 6- 1 A three-week advance notice of data submittal date will '

considerably decrease closure' delivery time. z.With said advance notice, the first closu're can normally be shipped ,.s...

within nine weeks of data receipt. An additional closure Mt'sv . .

can usually be shipped during each of the folloiting'weekt ' Q.& j

-.) toyallow - i if data was likewise submitted. gon gy, a timely" .'},h,. basis;if Q 1 four weeks shop time each.:) : ,

. j, , . ,'g.

u t . . . . . g. . f, . g< yc . . , f.,. . . , , , ,

c- . .

e' NOTE - Closures are normally supp1ted in,two sections.,per.. :g !

initiate an' agreement Systems Division Projects' with Westinghouse.WRD-PWRifp'for,one. piece e closures'.?r H AMy

A-

.,c.1,?% G:%s %mj%c:hl%i&W!%'

2;

'Q"p.. m' :

17. Utilizing weld.. techniques as employed,for hot and cold, ,

'y .d piping'ftt-up,' weld out the.sur e line~between hot.. M 'leg :d

' t ,

nozzle and 'v,: pressurizer.$g.? MS {," '%

j s, i.ut.W . . . .f# v. ,

)

18.'

Rig' same RC fit-up pipe techniques, closures weld closuresito 'into posttion,,an',d M ellsi pump casings agi 9 :;:

n' utilizing (

and .if.~in two sections, weld togetherMa$$.' (.L~ %.f'h ' .

?!.?Y %&.%'..'e.48kh$ M i&%.

." %vstm 19.f Install equipment support seismic restraint :- ' steel l . c.a .. c '

,r. ....

' % "$i y X ,i . .? O,4.l .'<9 W.. V. ?. w E'I

.p . w?'e*..'., y

..]y , , , , &p%'ifqf' .

'_.'y < - . y , qr.1 & n v. :! -., .,, hs:, ;. .';;. r;

u y, . , *

.w W. j v c, w ,4/ tr2 . -

. . , . , . d.ecy. . . .] t ' ,- - us' a....--

.. ,a. '

. .J. . G * ,. ,. .... .i.i'

'. \ - (, i' i ,,.fff v,. . .

k. . . . . m.<,u.

b. . . , :, ,. +. . c; v . ,, . . .g.ug p . . , ..3;. %"- ,. L e .M t

F: ,.

mc a'

.. : .1 n -w

NOTE : P.. . 3 ' a.I i @ ,. Y.

9.bo 4.-. ' .w, s Yi,.?.<, % s.r lC$n.b".p,i N:

tr, o S s ,' . '. W*..k

..J.'".-

m '. e . . . r rH ss t r% q.,

s l

, a fece. losure >

Only leg center,weldctWelds 'two' welders at theare puup requiredc'asing 'and steam to' start welding the., .

"'ge should not be started until sufficient centerNeld shrink'has .$ . , , occurred ;,"Qf ,. 1 to accomplish proper fit.to the, pump and s, team generator h ',f.$

S , .*, .

L+

.b

?

,a

".I:f N .e ., t:,

r.v.A<. , . r.g t

) : , ., . ,y ;.W w: ,r. u.-. - m . 3

.; @ y .

u :.n

tW! ,,

'k k W g.4 qy

~

,.?*.. co .. e .,w *

. * -y, t' ..,j- * ,' . 7 ,,

. 9.' i

~r . _y m ..gp .~ g ., 9 p ., f. g : - j n..W . .g.,'*, ,

. CQs .a :'i;... . . *4 i- p~.

. ;.lc.

' #,. .M' J ' i: l>&.;.

M.b'yfy%

i IV~l*6 VOL.I :

I i

l

_--------O

WESTINGHOUSE PROPRIETARY Ct. ASS 2 Rev. 2 8/30/74 l

. FES ,

.. GENERAL REACTOR VE3SEL $ETTING PROCEDURE _

c.; . e .e 1

- r**

s. ,.,

1. 0b.iective ,

,, . g ,. ,, g., . , ,. ,g g j

. :. . . .: . u . y, ,.. w,,: ,

Toprovideasafeandefficientmethodfor.3nstallat{on.,ott'entation.

f T ' and leveling pf, the reactor vessel.4'as.The

'the.Batum' plane ieacter for,)e'veling. vessel,must to be le

. .using the.' intern.als seating fla e:

)

~

'Thi vessei azimuth M(

~

within a tolerance of, tion mus ; ) Itolera'6ceallowance orientation and eleva

.- of ,the reacto(, coolant piping and t, he fuel ,transfergq jpnent. ,y

v i n t n p f = g:q .D ~ M.O r,,~ .: i. v jg.: s y.as ip foyfy II. Conditions

,, , d :. g.m .3g , .

ljng screws, and mis-

A.: .The reaet. or. vessel support shoes. ,. . shims, .leve., ,, , ,. gy ,, g., jr .

. m ce11r.neous. hardware are on hand. a

. a t- M .n.it . u-

..~. m .n :. .r - y;:.q I Acalibratedjigtransitortiltinglevelwiths(calhyor targets I

.s i

B. m' ..iis available for* service.p,The level .has a sere sensit

a. i

,j .than the jig (transit, but cliise' focusing"c6 bE F Rulem.40-

~

.t. e',.just;prjoy,t,ousein 1

-)

c:

- .t r setting Calibration'should the vessel.

+ . ' pe,-

jg,p;.;

checked,,at .

,the;.sg/.J2 iN.s'h

. c.

III. Precautions y,fgy w g g.,, p,,3 g; y yg'jf ~ g '

j';N N

Sc w..M . .M t b6 h a t.v P" ken tu . W.i dW 9 et t pn an dropping of Extreme care must be ta o prevent, con A.

tools, etc., onto the vessel to vessel *Reaif ma i g"Tufface or.into e . . -j. the , vessel..af ter .the ess ers ve e ved.3 ^

..yhd; W ,

. . a. n-u s.- w.w s.Nik.s*d 'Q. <

% w 'st B

ng. fixture to ssel ve&Q$'.G4%4W) pe bo ting for

't fl

, 3 '.,,, Recheck'.the. vessel 3l .. : dtightness prior *.to . lifting o l

.T ' " . . pads are' tight against'the ED[.cfy,f,,1,a, f fg. ,.,.,,

,.fa , . .

r:.. ' .. c ;w.. ::,;. . .w

.J -

. , C. j When vessel . shipping . skid is ,'used for,.thejspending pperation~

. ' . J , '". 'section botto verify that theof vessel' shipping' skid per:

to skid reta.in.i.n.g . vlite'

.. ,.'stiaps' ..,. o reactor,

- .. .ves e '&ge.m g

assembly is tight. 1 .

.-; -r; * *

.3.g. gg7, . ,,.g,,tay.i.N: p tr. q.. ; .,. . ,g.

1

] IV,. Instructions y p , % *y: :yyuf ,:. .

y .~ g$,., .

uppo ds. .

m;,.

A . -, .,p. a;., .n .. . e ... Set the reactor.vesse.l g,upport mm, hoes .. . w

. . . v.u .

.. .y g,-. . . a .< g,, ,. ,. w 3 eT ..

(e . .s pport pads.L.

.,3 ... .

cj rea a1 l . ~:.BC,',' Io

' ' ' ' Insta114the.;1eveling screws, n .the.The t' ris*of 'th'e levl s ximately

. spanner wrenches

-) .,

ww-w r ,.

.ws.

g.,

s .

YOL !

' IIV-2-1 1 I 4

f j

1 * \

WESTINGHOUSE PROPRIETARY CLASS 2 1 Rev. 2 8/30/74 WNES-50 NPS

-) j GENERAL REACTOR VESSEL SETTING PROCEDURE.

C. Set to the up',th'e vessel ,butnotonthevesselcover).

jig (transit or tilting' level Thisinsiust a conveniant be such area close

.s ,

that accurate focusing is possible at all scale access locations. ) )

D. Set leveling screws with tilting. level and optical scales at all.

supports,to.the established vertical target elevation. The leveliri ~g screw' elevation'is'to be' established by actual themfield '

h'oe elevations ~ a'nd reactor) vess as-erected" 10 s

as-built" diinensions'pport"'such' i

."'c.e.ntde s c v a. .

Witibn'w1jn.e, lsl sea.5 e'a' w,ac; chi m:eved.tissT4 '

. u.i.1"b:thit'.the'lorrect O +

kO91 ngupow;oA 1:I4th..? ; ,, -a ,. Jreact J

E. "ReinNE' sci 1E*i$cis's#c'oveEs (r'om*th'ereYe'to'r"visYeFihipYing' e through cover.-

Level the access readings ports. are ' Q taken g., on

.y::'.. the internals:..:ggida.0. seating flanyL .*1 x ..Q ,

u .. .

Noti! R!tf ishe'ciinnended ithatatil tilg' leve1Y. w.,. c . .. . . ., m . i.,

. readings due to the rathei' stiin' gent' lev'elnes's'teqVirements. .q.

re to bes:tached at o atvia verticality control.

"O.5 %pirit levels

.m.o w ig a.s.as .p.s to , scales s g j s > "A .,1 6 ' .

Gith the' leveling

F. $716wiPthT.FeactbVvEs'sil i to'the' fWin't-i..,f:i. 'c 6 on

.. * #screws'Jaking ,

'f rei that"th'e':vEsselfe,1gh't Ll's';' ully,japoseds#d; the'1evelinf sereks' befo're1t','is'a'ssured 'tha't "c'rews . are',f n - ,

' y . . ..,~.@gd.jpt;*f.g:+ '2 contact per gStep G.- .

.:. .!892p. fig,,t~g:

q. . W 1. . : c . w ' p v.f . . .

Note: Using'i' transit,^ keep"theivessel orientat .lbfft:hich"{. pthe' ' N.c n

vessel axis marks with t.he cardinal ixis"1 ons the.'a ^ ' ' '

l M ao geactor,TtWeYuvi?ea'r"n'"c.t 4 OW#.'U6:istesv: J U" *ut ens' um, .s 4 4

sc-

"' . m. .,p%Rd . yirs fw.per..,

c. s T

'G. italse7aYyj/v'eiln T'cr'ews s N a'ti re Ndt'1 O on"(idt' M 5e u der-0 ide the'v ag nt ined.f . . -

' .T.{.p1 %p' qQ,,

.N . I5: esse . . . 'shp,p,ojr,tj.y

.. . _,?$ntQ. ,sgP;.jl g' firm .

Q ($ l. con Q ?:.,. u. '.

' Level. ,the, reacter vessel to with1_nf A , tolerance 'of,0. 5",per, foot ,

H. ~

-id, justing 3:

rhi'sfng#thereactopvisse1%id I '

~1.

O"'U nf,f.14,nge diastaf pf'Regh'eckTa'11thE% brit 16n pt cal eyel To

~

the levellng verifyethat no' errors screwsef(have roduc besn M o tri{we' l'.

.W .' 'M ' rh

~ 3acnmn: betwpen fak'e heigh ( Measurements'r s ters w'the u support vers

.I. 4'g

' l

...,.Jpadsand..thesupport] hoe,"bearingssface"atp6 ints,on each " l

'supgiort 'ahd ~ record.',' .{.}y ,

. j'

.NoteY Telescoping gagis~"and kinicromeur,

' ,#' f i

60%"dy*N5$Mk p J, .a achine.a.11~ reapent bo g

. ,,:. J.fh.,f Step'J.'Mp

" T'jare ates 4 $% bje$t @j$. .

,,Wp

' ;.g'g. . bottom.sh* Note:. atha.g,,ished latis, smachin ,* 1 ,S d'th

, fin Wr .. 4m / s not be less than one (1)' inch.. . . E ' < +" !'. E 3. . '

. Qj.);.gf. p.

3.g-::.y}y{MV-2

, . ' .. , J. ~

2 YOL.!

l

. g WESTINGHOUSE PROPRIETARY CLASS 2 g h. /74 g l

FNES-SDNPS GENERAL REACTOR VESSEL SETTING PROCEDURE (continuedl K. Lif t the reactor vessel cnd back off the leveling screws until '.

they are .375" t .010" above the be.aring surfaces of the suppnrt shoes.

L. Install the holddown nuts.

}

M. Insert the pemanent bottom shim plates. Do not allow the bottom

' shim plates to be held off the bearing surface of the support shoes g:t;/,. ~

. by the leveling screws. ,v:e. .,,,,. .

.c. ,

N. Coat t ' top surface of s

.p,.; .: e .~m . > 2. ..

9.: .

Lower the nactor vessel .to a centered position on the shim plates.

O.

. . . . m :-

. ' ":' .e .:.s' . ;" & , ~.. :'. n.h.% ,.-

P3 Lift the reactor vessel and ' check,'for#s6rface contact betwe'en the

,t top of.the bottom shim plates and the underside of the' vessel. brackets.

., _ . giy:.4- .%4.y.y p.,... 394.:.;.y ..

*' y.

~ is' required between'the bottom ,

. Note: .A surface contact ac e s... -machine or hand dress, as I shim plates and the vesse necessary, to!.obtain required y .,

contact . area.y&f+,g':;pM-:yd'b.ht':r.k.:..

4. ; , '

W:l2..O , . * %<$1%ggfg i Blue' from' top ',, ..

.'Q. After' required surface co.ntact ~l is mett; remove Pruss ansurface~bf the' brac

~ -' surface of the shim plate's and' '

ry' film lubricant.d I

' ' ..}. '

top surface of shim plates' wit

. .(This treatment is improved by buf ng ac th. %:A<f.. f 4 L

.o %. - .. .

.'. c /.t % t i H; f n ,Q .. ;, G p u.(.},.:

.. l<3 . : -

R. Lower the reactor . vessel to a centered position on the . shim plates. :h

, -sh;; . ' , ,. : * .5

.> y :.\W&:.W.M:WWl-}%b; : , .wk S. Check reactor vessel levelness which should be within . tolerance.-

(SeeStep1.) Record final results. O!N.&9;OK 4B,&...,

. w. m. 9 : O. : y e . .g ntation.%.. .in.i.,.n. .;..n: ,g -g'the~

. . D. . .y.s .o ,. .

.orie. . . . us 4.

T.. Check.the. reactor vessel . X.g .(fgT,

,[,e^! .

. ]. Recor'd,..e.inal,results.'),g.

f fQ$j.@.y s y .~,. r.t , - ,- .,

for. axis,.

g4,.5 9. .. . , j.g:

. g.se ,2.- 3..

c.

'U.' Take gauge readings of the . spaces on a r si o- supportc.T,'~

I for ide shims '.- J. . J bi record . n Step k. ,},,,1

~\ . brackets '

'.'e . '; Mach.ine %:.TRG'to . hthe' support . shoes and l .:. dpemanent Q d iQ W Q Nshin' d { ^ sensionsyeco U, 1

J 'V. side) less" clearance as follows: 'platesTwo and three-loop vessels'. e

' coldclearanceof.015"andfour-loopvessels.020fo6)eachs

., of the . vessel support lug. " .c( d W.,$

s - ' j.3. ,g 2 g

U. . V: :

..) Q2 .i.~:L i. h. M .y .0,*.

port pad, f,. q-

\ i j:conta

/ . ' W. Coat surfaces of side shim plates .above ;. .

ddb ;with Molykote Type-Z{ry' ila ubri cant 9 nW,Me;c.a.s.sy.g .atd on .t

M.

X. '

^E 8 T> side.h,hims'.l.'?~ Lift AneEreactor%vesse I Wre.v

',~,

.m.

Wr 1' my'!. ,.,:.,, p

?

j .8 Insert' side shim plates u:from ; q., . above 'thE support' h

.- Y. ,

+1Qx:- -

~'

C' *- ..~~,f ~ '.;...),.

My : .,W,., ~ W Z. Repeat Steps R, 5, and T. . . . . m;;. - ..

Ok r fV :!V-2-3 YOL.! . .

+

g Sfo7 85-o b 4, 4 4 m h m s c 4 T l i

I

/

[Ph tzE I

g ,7 e 8 e

~

....- i

, .*;; . . , % '~.' : ,

./s ~.s*,, UNITED STATES 8 r, NUCt. EAR REGULATORY COMMIS$10N l

q

7. ]' tuASHINGTON. D. C. 2GSS$ l 1

JUL 0 31985 Docket Nos.: 50-445 ,

{

I and 50-446 i

e-F, !

Mr. M. D. Spence President Texas Utilities Generating Company 400 N. Olive Street Lock Box 81 Dallas Texas 75201

Dear Mr. Spence:

1

Subject:

Use of ASME Code Edition and Addenda for the Comanche Peak .

Steam Electric Station (Units 1 and 2) !

The NRC staff has received your request for approval to use a later ASME l Code Edition and Addenda than that currently in the Code and Standard Rule (10 CFR 50.55a). identified in a TUGC0 letter dated December 21, 1984 )(

Before addressing your specific request, the staff has aske.d that we clarify to TUGC0 the applicability of ASME Code Editions and Addenda, l l

as well as the scope of the 10 CFR 50.55a rule in order to resolve any misunderstandings on the part of your staff.

The most recent version of 10 UR 50.55a. dated March 30,1984, approves 4 the editions of Section I!! of the ASME Boiler ar.d Pressure Vessel Code, through the 1980 Edition and Addenda through the Summer 1982, and is only applicable to Code Class 1. 2 and 3 components. The rule does net address i Section III components such as: ClassMC(SubsectionNE); Supports (Sub-section NF); and Core Support structures-(Subsection MG). The NRC staff is currently in the process of approving additions.to Section !!! of the Code through the 1983 Edition and Addenda through the Summer 1984. that are also written, Jon1 applicable tMs new revisiontoofCode Class the rule does1,not 2 and 3 components.

address As currently such componehts as those constructed to Subsections NE. NF T 6iI W of Section III, as noted above. Caution should be exercised by your staff in the use of Code Editions and Addenda for which approval is pending. Use of aty such Code Editions and Addenda is at TUGC0's own risk.

%-4 yy . -

i l

Mr. M. D. Spence JUL 0 31985 While NRC staff approval is not required by 10 CFR 50.55a to utilhze a later Code Edition and Addenda of Subsection NF (Section III) of the Code, staff approval of the standard used for the construction of the component supports is required by General Design Criterion 1 of 10 CFR 50. Implementation of a later Code Edition and Addenda must be in conformance with NCAs1140 of both the later Code Edition and Addenda of Section III. and the Code of record for the facility. It is the responsibility of TUGC0 to epsure that all related provisions of the Code are adhered to, particularly when selected paragraphs and tables of a later Code Edition and Addenda are used in conjunction with the Code of record. To that end TUGC0 shall prepare a listing for NRC staff review of those Code provisions TUGC0 considers related and with which it will comply. The list shall be compiled from the later editions and addenda speci-fied in the December 21, 1984 graphs or tables requested to,be letter used.corresponding)to the eight NCA-1140(f requires plantselected owners topara-

detennine the acceptability of both the Code Edition and Addenda established in the design specification, and in later revisions of the design speciff-cation with regulatory authorities such as the NRC.

Since your request for approval to use the later ASME Code Edition and Addenda was received when the design of Comanche Peak Units 1 and 2 is largely com-pleted, provide the reasons and basis for requesting the revision to earlier design conrnitments. Since there is extensive testimony by your representatives and the staff.on the use of Section III of the ASME Code for Comi.nche Peak, the staff requests that you perfonn a careful review of the hearing record to identify whether (and in what manner) each of the request items may affect the In additions the staff requests that you earlier identifytestimony whether and(before themanner) in what ASLB. the request items may relate to:

(a) The Motions for Suninary Disposition submitted by TUGC0 concerning piping and pipe supports; (b) Matters which were reviewed by Cygna in the Indeperadent Assessment Program; (c) The conrnents of the NRC's Technical Review Team; (d) The comments of the NRC's Special Review Team; (e) The coninents of the NRC's Construction Appraisal Team.

With respect to your December 21, 1984 request, the NRC staff finds the use of the listed portions of ASME Code Section III Subsection NF acceptable when the following have been implemented:

(a) Identify in the FSAR affected component supports (i.e., by systems and supports) to which the later Code requirements are to be applied;

____________s

. _ _ . . ~.

'. .)

- Mr. M. D. Spence JUL 0 31985 (b) TUCCO shall assure that all code related requirements have ;

been met; a listing of the related requirements shall be '

submitted for NRC staff review as discussed above. ,

(c) Applicable design specifications and required design !

documents shall be revised. /

Further,werequestthatyouconsiderwhetheraConstbettonPermitamendment is necessary. If not, please provide your rationale. {

Should there be further questions concerning this subject, please contact the Project Manager.

4 Sincerely, for Comanc e Pe Mr Project Division of Licensing

)

cc: See next page i

I e

_ _ _ . _ . . . _ . _____-__A

5 l

. l i

APPENDIX A ,j NOTICE OF VIOLATION -

Texas Utilities Electric Company Docket; 50-445/85-07 Comanche Peak Steam Electric Station ' 50-446/85-05 Units 1 and 2 * -

Permit: CPPR-126 CPPR-127 1985, violations t During of NRC an NRC inspection requirements conductedInon were identified. April 1 through accordance June 21, General Statement with the of Policy and Procedures for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1995), the violations are listed below:

, , 1. Failure to Promptly Correct an Identified Problem with RTE - Delta Potential Transformer T11 tout Subassemblies -

10 CFR 50, Appendix 8, Criterion XVI, as implemented by Texas Utilities Generating Company (TUGCO) Quality Assurance Plan (QAP), Section 16.0, Revision 0, requires that measures shall be established to assure that- <

conditions adverse to quality, such as failures, malfunctions, deficien- {

cies deviations, defective material and equipment, and nonconformances are j promptly identified and corrected.

Contrary to the above, a potential problem with RTE - Delta potential transformer tiltout subassemblies, which are used in the emergency diesel generator. control panels, was identified to the applicant via a letter, dated June 15, 1983, from Transamerica Delaval Inc. This letter also provided instructions for correcting the potential problen. However, the

~

applicant did not take the corrective action. The NRC initially reported this item as unresolved in NRC Inspection Report 50-445/84-40.,

' This is a Severity Level IV violation. (Supplement II.E) (445/8507-01 446/8505-01).

2. Failure To Follow Procedures l 10 CFR Part 50, Appendix 8. Criterion V, as implemented by the TUGC0 QAP, Secticn 5.0, Revision 2 requires that activities affecting qulity shall be prescribed by documented instructions, procedures, or drawings, of a ;

type appropriate to the circumstances and shall be accomplished in accor- ;

dance with these instructions, procedures, or drawings.

a. Drawing 2323-51-0550, Revision 4. Section 6 6 specified the use of Class E" concrete for the Unit I reactor coolant pump and steam generator supports.

A. ,y h ~$ ) I: g . _ . _ _ _ _ _ _ _ _ _ _ -

b Contrary to the above, comercial nonshrink grout was us)d to grout the Unit I reactor coolant pump and steam generator supports in lieu of Class "E" concrete. (445/8507-02)

ThisisaSeverityLevelVviolation(SupplementII.E).

b. Brown and Root Procedure QI.QAP-7.2-8, " Receiving df Westinghouse-Safety Related Equipment," Section 3.1.d.l. requires a QC inspector to verify that the Westinghouse Quality telease (QR) document l checklist items be filled out completely and accurately.

Contrary to the above, the voltage recorded on Westinghouse QR 41424 checklist, attachment 1 step 4.1, was outside the specified i tolerance, but the QC receipt inspector accepted QR as satisfactory.

(445/8507-03) ;

This is a Severity Level IV violation.

S c. Brown & Root Procedure 35-1195-CCP-10 Revision 5, dated December 4

_ 1978, requires that central and tru.ck mixer blades be checked quarterly to assure that mixer blade wear does not exceed a loss of 10% of original blade height.

Contrary to the above, on May 31, 1985, the NRC inspector detemined ,

thattherewasnoobjectiveevidence(records)thatthemixingblades !

had been inspected quarterly since the trucks were placed in service in 1977. (445/8507-04;446/8505-02)

This is a Severity Level V violation (Supplement II.E)

d. Brown & Root Procedure CP-QAP-15.1, " Field Control of Nonconforming Item. " states that nonconforming conditions shall be documented in a Deficiency and Disposition Report (DDR), Procedure CP-QCP-1.3, " Tool

~

Equipment Calibration and Control " dated July 14, 1975, states that out-of-calibration equipment shall be identified on a DDR.

Contrary to the above, on May 31, 1985, the NRC inspector reviewed the calibration file for scale (MTE 779) used for weighing cement and found that a 24-48 pound deviation from the required accuracy was ,

encountered with the water and cement scales during a 1975 calibration '1 of thisthe backupand condition plant scales, require however,ofno disposition theDDR scaltwas and issued to identify (if concrete any) produced. (445/8507-06;446/8505-04). j ThisisaSeverityLevelIVviolation(SupplementII.E).

i 4

l

3-Pursuant to the provisions of 10 CFR 2.201. Texas Utilities Electric Company is hereby required to submit to this office within 30 days of the date of the letter transmitting this Notice, a written statement or explanation .in reply, including for.each violation: (1) the reason for the violations if admitted, ll the corrective steps which have been taken and the results achieved, I the corrective steps which will be taken to avoid further violations, and I hthedatewhenfullcompliancewillbeachieved. Where good cause is shown, I consideration will be given to extending the response time. l Dated at Arlington, Texas,

'his

. 3rd day of February.1986 P

me g de

l APPENDIX 8 U. S. NUCLEAR REGULATORY COMISSION '

l REGION IV NRC Inspection Report: 50-445/85-07 Permit: CPPR-126 l 50 446/85-05 CPPR-127 !

Docket: 50-445; 50-446 ,

):

Applicant: Texas Utilities Electric Company ,(TUEQ) l Skyway Tower 400 North Olive Street Lock Box 81 Dallas, Texas 75201 1

Facility Name: Comanche Peak Steam Electric Station (CPSES) l Units 1 and 2 Inspection At: Glen Rose, Texas Inspection Conducted: April 1,1985, through June 21, 1985 i

Inspectors: M J/ /# !If '

H. 5. Phillips, 5enior Resident Date Reactor Inspector Construction (pars. 1, 2, 3, 8, 9, 10, 11, 15, 16, 17, 18, and 19) b ht b- r" 2Ab" /WAlt5 d d. E. cummins, senior Resident Reactor pati

) Inspector Construction (April 1 - May 10,1985)

(pars. 1, 3 and 19) l

c. E. norman, Reactor Inspottor AkNI pate (pars. 1, 12, 13, 14, and it)

Ak A '-A . Wikr pate D. M. Hunnicutt. Section chief .

Reactor Projects tranch 2 (pars.1, 4, 5, 6, 7, and 19) l rf po QG r

4 ) (,;

2-t Approved: b LW D. M. Hunnicutt Section Chief, bbf

' Date Reactor Project Section B I

Inspection Sumary /

Inspection Conducted April 1,1985, thmuch June 21.I1985(Report 50-445/85-07)

Areas Inspected: Routine, announced and unannounced inspections of Unit I which included plant tours and review of plant status, action on previous NRC inspection findings (violations / unresolved items), review of documentation for sitedams,andreviewof10CFRPart21and10CFRPart50.55(e) construction deficiency status, The inspection involved 77 inspector-hours onsite by four NRC inspectors.

- . ~

Results: Within the areas inspected, five violations were identified: fail-ure to promptly correct an identified problem with RTE - Delta Potential

- Transfonner T11 tout Subassemb11es, paragraph 3.a.; comercial non-shrink grout was used to grout the Unit I reactor coolant pump and steam 9enerator supports in lieu of Class "E" concrete, paragraph 3.b.; hydrogen recombiners out-of-specification voltage recorded on quality release document but QC receipt inspector accepted, paragraph 3.c; failure to provide objective evidence to show that central and truck mixer blades were inspected, paragraph 8; and failure to issue a deficiency report on cement scales that were out-of-calibra-tion, paragraph 9.c.

Inspection Sumary Inspection Conducted April 1.1985, through June 21, 1985 (Report 446/85-05)

Areas Inspected: Routine, announced and unannounced inspections of Unit 2 which included plant tours and review of plant status, action on previous NRC inspection findings (violations / unresolved items), ritview of documentation for site dams, review of documentation for voids behind the stainless steel cavity liner of reactor building, observation of NDE on liner plates, inspection of concrete batch plant, review of calibration laboratory records for batch plant, review of concrete laboratoryconstruction review of reactor pressure vessel (RPV and pipin testing,) inspection of leve deficiency status, and review of violation and unresolved items status. The inspection >

involved 335 inspector-hours onsite by four NRC inspectors.

Results: Within the sixteen areas inspected three violations were identified:

failure to correct RTE-Delta transformer problem, paragraph 3.a; failure to provide objective evidence to show that concrete central and truck mixer blades were inspected, paragraph 8; and failure to issue a deficiency report on cement scales that were out-of-calibration, paragraph 9c.

i l

3 DETAILS l

1. Persons Contacted Applicant Personnel M. McBay Unit 2 Reactor Building Manager

- 8. Ward, General Superintendent, Civil .'

D. Chandler, QA/QC Civil Inspector

.W. Cromeans, QA/QC, TUGC0 Laboratory / Civil Supervisor

dJ. Merritt, Assistant Project General Manager

fP. Halstead, Construction Site QA Manager

C. Welch, QA Supervisor TUGC0 (Construction)

J. Walters. TUGC0 Mechanical Engineer

- - K. Norman, TUGC0 Mechanical Engineer ~

J. Hite, B&R Materials Engineer G. Purdy B&R CPSES QA Manager i .

Denotes those present at May 10,1985 exit interview.

Denotes those present at' June 10, 1985 exit interview.

- The NRC inspectors also interviewed other applicant employees during this inspection period.

2. Plant Status Unit J.

At the time of this inspection, construction of Unit 1 was 99 percent

- - complete. The fuel loading date for Unit 1 is pending the results of ongoing NRC reviews.

[ Unit 2 At the time of this inspection, construction of Unit 2 was approximately i 74 percent complete. Fuel loading is scheduled for approximately 18 ;

months after Unit 1 fuel loading.

3. Applicant Action on Previous NRC Inspection Findings

a. (Closed) Unresolved Item 445/8440-02: Potential Problem with l Potential Transfomer T11 tout 5ubassentlies. .

By letter dated June 15,1983 Transamerica Delaval notified the applicant of an RTE - Delta 10 CFR Part 21 report to the NRC i

reporting a potential problem with the primary disconnect clips of the potential transfomer tiltout assembly used in the emergency diesel generator control panels at CPSES. The Transamerica Delaval

(

r W

q l i

i lette also provided instructions for correcting the problem, i However, the NRC inspector could not determine if the problem had been corrected at CPSES and made this an unresolved ites. 'The applicant determined that the problem had not been corrected and >

subsequently performed the recossended corrective action. The Unit I corrective action work activities were documented on startup work )

permits 2-2912 (train A) and 2-2914 (train B). The Unit 2 work l activities are being tracked as master data base 008) item 3003-31.

The failure to promptly correct this idevitified problem is an apparent violation (445/8507-01; 446/8505'-01). I

b. (Closed) Unresolved Item 445/8416-03: Coeunercial Grout Used in Lieu ;

of Class "E" Concrete !

The applicant determined that the use of nonshrink commercial grout in lieu of the Class "E" concrete specified on drawing 2323-31-0550

-. was acceptable. Design Change Authorization 21179 was issued to ,

drawing 2323-$1-0550 accepting the use of the commercial non-shrink-

~

> grout. However, the failure to grout with Class "E" concrete as !

.- specified on the drawing at the time the work was accomplished is an apparent violation (445/8507-02).

c. (Closed) Unresolved Item 445/8416-04: Hydrogen Recombiners -

Out-of-5 specification Voltage Recorded on Westinghouse Quality Release Document i Quality Release N-41424 was revised by Westinghouse changing the.

specified voltage from 10+-2V to 12+-2Y which put the questionable voltage within specification limits. However, the failure of receipt ;

inspection to verify that the QRN-41424 was filled out accurately as I required by Procedure QI-QAP7.2-8 is an apparent violation

_ ,_ (445/8507-03).

- d. (0 pen) Unresolved Item 445/8432-06: 446/8411-06: Lobbin Report Described Site surveillance Program Weaknesses During this reporting period the NRC inspector reviewed the status of this o en item several times and interviewed TUEC management and site q survei lance personnel. The Lobbin report stated that the scope and i objectives of the site surveillance program were unclear, lacking I both purpose and direction. )

i There is no specific regulatory requirement to have a surveillance d program; however, TUEC committed to have a surveillance program and has established arocedures to implement'such a program as a part of the 10 CFR Part 30, Appendix 8. QA program. This extra effort is a strength; however, the NRC inspector also observed, as did the Lobbin Report, that the surveillance program lacks both purpose and direction to be effective and complimentary to the audit and l

l r

inspection programs. Since the TUEC audit group is not Ip' cated on site, the TUEC surveillance program on site takes on edited

significance.

This item was discussed with the TUEC site QC manager who described a reorganized site surveillance function and changes that have occurred. New procedures which describe this organization's duties and responsibilities art forthcoming. r.

TUEC has elected to defer responding to the violations pertaining to the audit function in NRC Inspection Report 445/84-32; 446/84-11, but rather to have the Comanche Peak Response Team (CPRT) respond to this report and other QA matters. The surveillance issue is closely tied to the audit deficiencies in NRC Inspection Report No. 445/84-32; 446/84-11. This item will remain open pending the review and imple-mentation of the CPRT action plan. A special point of interest will

- - be how audits and surveillance work together to evaluate the control -

of all safety-rvlated activities on site to assure gus11ty,

" especially the overview of quality control effectiveness.

4. Document Inspection of Site Dams The NRC inspector reviewed documents describing the inspection activities p(erfomed on the SSI) for impounding Squaw cooling waterCreek Dam for the two units(SCD) and at CPSES. Thethe safe shutdown A secondary i purpose of the SCD is to impound a cooling lake for CPSES.

reservoir (SSI) is fomed by a channel connecting the SCD impoundment to the SSI.

Three documented inspections have been perfomed since 1980. The inspections were:

l

a. Relevant data for SCD is contained in Phase ! Inspection, National l

. Dam Safety Program Squaw Creek Dam, Somervell County: Tene, amos {

River Basin, inspection by Texas Department of Water Resources. Date 1

of Inspection: June ID, 1980.

b. Inspection on August 25,1982, by registered professional engineers from Mason-Johnston & Associates, Inc., and Freese & Nichols, Inc.

c. Inspection on September 19,1984, by a registered professional engineer from Mason-Johnston & Associates, Inc.

The inspection activities consisted of visual inspections by inspection teams that included accompanying Texas Utilities Service, Inc. (TUSI),

and Texas Utilities Generating Company (TUGC0) representatives.

Photographs were taken as a part of the documentation. The data for the 1

I l

' eine

l piezometer observations and the data for the surface referencg monuments )

1 were reviewed by applicant personnel and Mason-Johnston engineers.

No items of significance were observed or reported by these inspection teams. Slight erosion areas were observed and reported. A cracked area on the service spillway upstream right bridge seat was observed by the inspection teams and continued monitoring of this area was recommended by Mason-Johnston and Associates. No signs of cracks, settlements, or horizontal movement at any location within the SCD or the $51 were reported. ,

The NRC inspector reviewed the applicant's records and the Mason-Johnston inspection reports. These documents indicated that the SCD and SSI were

. structurally stable and that the applicant was performing inspection activities to maintain the structural integrity of these dams.

The state of Texas requires periodic inspections of these dans (principally the SCD) due to inhabited dwellings downstream. The applicant has met these inspection requirements. -

No violations or deviations were identified.

5. Voids Behind the Stainless Steel Cavity Liner in Unit 2 Reactor Buildino In review of previous related TRT concerns, the NRC inspector reviewed applicant records, including NCR C-82-01202; NCR C-1784, Rev.1; NCR C-1784, Rev.1; NCRRev.C-1824, 2; NCRRev.

C-1766, Rev.1; NCR 2; Significant C 1791, Deficiency Rev.1; Report Analysis NCR C-1824,(SDAR)

- 26, dated December 12,1979; DCA-20856; and Gibbs and Hill Specification 2323-SS-18. The review of records and documentation and discussions with various applicant personnel indicated the following:

Structural concrete was placed in Unit 2 reactor building at

- -- elevation 819 feet 6-3/4 inches to 846 feet 6 inches on June 21, 1979. This concrete was placed adjacent to the stainless steel liner

' wall s. The concrete forms for this pour were not removed until October 1979 due to subsequent concrete placements for the walls to elevation 860 feet 0 inches. When the forms were removed, hone and voids were observed by applicant personnel. The applicant'ycombss review of the extent of unconsolidated concrete resulted in the l issuance of SDAR-26 on December 12, 1979. Investigations were begun and Meunow and Associates (M&A) of Charlotte, North Carolina, were contracted to perform nondestructive testing on in-place concrete.

M&A performed these tests on a two foot grid pattern on the compartment and liner sides of all four steam generator (SG) compartment walls. The selected test locations did not include the locations where the voids were later found to be located. 4

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In August 1982, preparations were made to pour the concryte annulus around the reactor vessel. When the expanded metal forwwork was removed from the reactor side of the compartment walls, voids were observed and NCR C-82-01202 was prepared. DCA 20856 was prepared as a procedure to repair the void area. DCA 20856 indicated that the voids were not extensive (a surface area of about 28 square feet by 8 inches maximum depth) and that the repair procedure assured that the ,

.} total extent of voids had been identified. One half (0.5) of a cubic I yard of concrete was used to complete the repairs as indicated on grout pour card 261. .

i The applicant's review and evaluation of the gird pattern and a comparison of SG compartments 2 and 3 to 1 and 4 indicated that voids did not exist in SG compartments 2 and 3. The review of test girds i extended down to elevation 834 feet, which is the floor elevation of the liner. The liner walls of SG compartments 1 and 4 were not tested at elevation 834 feet, but at elevation 836 feet which is

' - above the area of the identified voids. No testing was done on the.

A' liner side of the area of the voids below elevation 836 feet. The program also included removal of 2 inch x 2 inch plugs from the stainless steel liner at locations where test indications raised questions concerning the concrete. The inspections of the concrete by applicant personnel after the plugs were removed confirmed that there. were no additional unconsolidated concrete areas (voids).

In accordance with OCA 20856, the applicant removed stainless steel liner plates from three areas (one area about 1 foot by 1 1/2 feet and two areas about 3 feet by 1 foot, excavated or chipped to sound concrete, and cleaned the concrete surface area. One and one-quarter inch (11/4)diameterprobeholesandgroutaccessholesweredrilled in the liner plates to determine the extent of and to assure full definition of the void arca. Air access holes were drilled in the

- - stainless steel liner plates to assure that grouting would be accomplished in accordance with the procedure.

The procedure (DCA-20856) specifed that the grout was to be cured for 28 days or until the grout reached a compressive strength of 4000 psi. Repairs to the liner plates were specified in DCA-20856 and G&H Procedure 2323-5S-10.

DCA-20856 required that under no circumstances was cutting of the liner across weld seams, across embedded weld plates, or into leak chase seal welds or drilling through the liner at leak chase channels, embeds, or weld seams permitted. Documentation review indicated that DCA-20856 was adhered to and that no cutting or drilling occurred in prohibited locations.

No violations or deviations were identified.

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6. Nondestructive Testing Observations of Liner Plates in Fuel Iransfer Canal The NRC inspector observed portions of non-Q liquid penetrant examinations (PT) being performed on liner plate welds following re-installation of the liner plates in the areas of the fuel transfer canal removed for inspection and repair of the concrete. The inspector performed the PT on the welds as required by the repair package and the proegdure (QI-QP-11.18-1,

Liquid Penetrant Examination').. Scatt e ed weld porosity was identified by the inspection. The porosity was greeund out and a repeat PT was performed. The final inspection is scheduled to be performed by QC inspection personnel. The liner plate areas to be inspected by PT were identified in DCA 20856.

No violations or deviations were identified.

7. Cadweld Splice Observations and Records

a. Calibration of Tensile Tester -

1 The NRC inspector observed the calibration of the Tinus-Olson Universal Testing Machine (Model Number 600-12 Identification Number M&TE-784) on April 2 and May 7, 1985. The machine was calibrated just prior to perforsing tensile testing of cadweld splices and subsequent to completion of tensile testing each day that tensile testing was perfonned. The machine calibration date for April 2, 1985, prior to start of tensile testing was observed by the NRC inspector and recorded as follows:

Nominal load Calibration Reading Error Error Remarks (lbs) (lbs) (lbs) 5 0 0 0 0 0 machine on 4/2/85 100,000 99,750 +250 +0.25

+0.2 200,000 199,600 +400 300,00 299,450 +550 +0.18 350,000 350,300 -300 -0.08 400,000 401.200 -1200 -0.03 l 500,000 501,350 -1350 -0.27 600,000 602.450 -2450 -0.40 The NRC inspector reviewed calibration data for March 4. March 8 April 2, April 3. April 30, and May 7, 1985. All calibration data met within the +/- 15 accuracy requirement specified by Calibration Procedure 35-1195-IE!-37. Revision 3, dated March 11, 1982. The reference standards were identified as follows:

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10 No. Manufacturer Calibration Due Date _i RS-75 BLH Electronics January 27,1987 RS-75.3 BLH Electronics January 27,1987

b. Observation of Cadweld Splice Tensile Testing (1) Qualification Tensile Testing ,,

On April 2,1985, the NRC inspector sbserved the following tensile testing of cadweld splices for cadwelder qualification:

EBD Q8, GBH Q1, GBH Q2, GBV Q1, 8FD Q4, 8FD Q3, 8FH Q4, GAH Q1, e

GAV Q1, and GBV Q2.

Each of the above qualification cadwk1d splices was tensile tested to 400,000 pounds (100,000 psilandmettherequirements

- . stated in the procedure. ,

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> (2) _ Production Tensile Testina .

The NRC inspector observed the tensile tester calibrations and the following production cadweld splices tensile testing on May 7, 1985: FXD 3P, FYD 4P, FYD 8P, FAD 87P, and FUD 6P.

Each of the above production cadweld splices was tested to 400,000 pounds (100.000 psi)andmettherequirementsstatedin the procedure.

(3) Installation of Production Cadweld Solices The NRC inspector observed installation of robar and catbeld splices at frequent intervals (five or more observations per week during the weeks of April 8 and 15; May 6,13, 20, and 27; andJune3,1985). The rebar installation for the Unit 2 closure was performed in the area identified as elevation 805 feet to elevation 875 feet and artmuth 300 degrees to 335 degrees. The installation activities observed included rebar spacing, location of cadwelds, observation of selection and removal for testing of cadweld splices for testing, and determination of location of rebars and cadwelds for the as-built drawings.

(4) Documentation Reviewed

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The NRC inspector reviewed the following documentation for the rebar placement and cadwelding for the Unit 2 containment (reactorbuilding)closurearea:

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N Drawings DCAs NCRs k 2323-5-0785, Rev.7 22616. Rev. 1 C85-206294 2323-5-0786, Rev.9 22728 C85-200339, Rev.1 2323-51-500, Rev.5 22737 C85-200355 Rev.1 2323-51-506 Rev.5 22836 2323-52-505, Rev.5 22878 (Sheets 1-7) 2323-52-508, Rev.2 22772 /

2323-52-506, Rev.3 .

No violations or deviations were identified.

8. Concrete Batch Plant Inspection. Unit I and 2 The NRC inspector inspected the concrete production facilities for the following specific characteristics for the following areas: (1) material

- - storage and handling of cement, aggregate, water and admixture, (2) batching equipment scales, weighing systems, admixture dispenser, and recorders, (3) central mixer (not applicable because it had been dismantled),

(4) ticketing system, and (5) delivery system.

The current batching is a manual operation since almost all concrete has been placed. The central sixer was dismantled ard removed from site two or three years ago when concrete placement was virtually completed.

Presently, the backup batch plant (which was a backup system for the central mixer) is in operation to complete the remaining concrete placements. This batch plant is in good condition and complied with the subject checklist except for one area.

The NRC inspector inspected the inside of one of three trucks used for mixing concrete (that is, the batch plant dispenses the correct weight of

-- materials as required by the specific design six numbers and the truck then mixes the batch to be placed.) The blades inside the truck are

. subject to wear and should be checked at a reasonable frequency. The Brown & Root (B&R) representative responsible for checking the blades in accordance with BAR Procedure 35-1195-CCP-10. Revision 6 dated ,

December 4,1978, was asked for evidence that the blades had been checked for wear on a quarterly basis as required by procedures and it was found that there was no record of such checks dating back to 1977 when they were initially checked.

In the FSAR Yolume V, Section 3.8.1.2.3, the applicant counits to ACI 304-73. In ACI 304, the maintenance of mixer blades is required.

Pro:edure CCP-10, paragraph 3.10 ' Truck Mixing," is silent on blade wear I but Section 3.11 infers that the blades should be checked for both central and truck mixing. The inspection of both central and truck mixing blades

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was not documented, although the BAR representative stated that the mixing blades were periodically inspected and laboratory testing wouid have probably indicated if there was a problem with the mixing blades.

Strengtn and uniformity tests have consistently been within the acceptable range indicating that concrete production was acceptable even though mixing blade inspection was not documented.

Otherwise, the condition of the inside of the truck was satisfactory as the drum and charging / discharging were clean. The water gage and drum counter were in good condition. ...

This failure to follow procedures is a violation of 10 CFR 50 Appendix B. I Criterion Y. Subsequent to the identification of this violation, the i blades were checked for wear and blade wear was presently within allowable Ilmits (445/8507-04; 446/8505-02). )

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,, No other violations or deviations were identified. !

3 9. Calibration Laboratory for Batch Plant Unit 1 and 2

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The NRC inspector obtained batch plant stale numbers from tags which indicated that the scales had been calibrated and were within the calibration frequency. Cement (MTE 779), Water (KTE 766), admixture scale (MTE 764), and aggregate (MTE 780) were reviewed. The scales had been periodically calibrated since the batch plant was activated. The records were adequate except as follows:

a. Scales NTE 766 records do not differentiate between the required accuracy of the scale and the digital readout. ,

b. Scales MTE 779 and 780 records show various accuracy rarges for the same scale; i.e., MTE 779 (SN749687) records the following: report dated January 1976 gives 15; report dated July 1976 gives 15 while the report dated October 1976 gives +/- 0.21.

the tbove items are unre-The solvedcalibration appeared pending further review oftothe bea 11 proper, cant however,s actions regarding the correction of these records (445/8507 ; 446/8505-03).

c. Records for scales MTE 779 records contained 88R memo IM-1108 dated July 16,1975, which described a nonconforming condition. This condi-tion affected the water and cement scales causing h 24-48 pound deviation (7.000 pound scale) during the calibration test. The meno stated that the condition was corrected and the scales were then calibrated; however, no deficienc report was written as required by B&R Procedure CP-QCP-1.3, { Tool and Equi nt Calibration and Tool Control' dated l 1

July 14,1975, and CP-QAP- 5.1, Tield Control of Nonconfonsing I Items," dated July 14, 1975. As a result there is no evidence that

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corrective action included an evaluation to determine ,1ficancrete production was adversely affected. ,,

This failure to assure that a nonconforming condition was evaluated is a violation of Criterion XV of 10 CFR Part 50, Appendix 8, (445/8507-06; 446/8505-04).

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10. Concrete Laboratory Testing Units 1. and 2 TUGC0 Procedure Q1-QP-11.1-1, Revision 6; was compared with ASME i Section !!!, Division 2, Subsections $222, 5223 and 5224 to assure that each ASTM testing requirement was incorporated into the procedure.

The NRC inspector inspected the testing laboratory equipment and found the test area and equipment were in good condition and each piece of equipment was tagged with a calibration sticker which showed it to be within the required cat (bration frequency. Test personnel were knowledgeable of test -

requirements and e:ivipment.

The NRC inspector witnessed field tests performed by laboratory personnel as follows:

Air Content (5) Slump (in.)

Date Truck No. Mix No. Ticket No.

1em N 64013 Reg 8.2-10.3 MA 70 wx 6/3/85 RT-41 925 57 Mea 8.7-9.1 NA 64014 Reg 5.0-7.0 5 max 70 max 6/3/85 RT-35 128 Mea 6.6 6.25* 57

<1 nit 5at slump was high; however, after additional truck rotations the

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sluep was four.d acceptable.

The fo11cwing laboratory equipment w&S checked and found to be within calibration: Torney Compression Tester, MTE 3031; Temperature Recorder MTE 3013 and 3014; Unit Volume Scale, MTE 1053; Pressure Meters NTE 3000B, 3002 and 3004; Steves MTE 1286,1239,1272,1274,1136A,1156,1094,1093, 1095,1178,1179,1300 and 1180; Aggregate scales, MET 1058 and 1067; and 2" grout sold MTE 1111.

The following test records for placement number 201-5805-034 were reviewed: (1) concrete placement inspection, (2) concrete placement summary and, (3) unit weight of fresh concrete.

No violations or deviations were identified.

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11. Inspection of Level C and D Storage Unit 1 and 2

,j The NRC inspector inspected all laydown areas where piping, electrical conduit, cable, and structural reinforcing steel were stored. These materials were neatly stored outside on cribbing in well drained areas which allowed air circulation and avoided trapping water. This met the Level "D" storage requiremeu.s of ANSI M45.2.2. ,.

The electrical warehouse contained miscellaneous electrical hardware.

This building was required to be fire resi' stanti weathertight, and well ventilated in order to meet Level "C" storage requirements. This warehouse was well kept and met all requirements except for a lock storage area located upstairs at the rear of this building (electrical temination tool room). Two minor problems were identified and the warehouse personnel initiated action to correct them.

The first problem noted'was that a box of nuclear grade cement was marked

shelf life out of date" but it had no hold tag. The box was subsequently o

tagged inaccordance with 70600 nonconfomance Procedure CP-QAP-16.1

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Revision 24 (Nonconformance Report (NCR) E85-200453) after being identified by the NRC. During discussions with the warehouseman, the NRC determined that engineering told the warehouseman to mark the material and lock it up, but did not tell him to apply an NCR or hold tag. Also, the NRC inspector noted a very small leak in the roof above the electrical temination tool room. This lesk was in an area that did not expose hardware to moisture.

The roof is currently being repaired.

The millwright werehouse storage area was inspected; however, only a small number of items or materials were stored in this area. The overall storage conditions in this area met or exceeded Level "C" storage requirements.

- - No violations or deviations were identified.

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12. Reactor Pressure Yessel and Internals Installation - Unit 2 This inspection was performed by an NRC inspector to verify final placement of the reactor pressure vessel (RPV) and internals by examining the completed installation and inspection records.

I a. Requirements for Placement of RPV Requirements for placement of the RPV to ensure proper fit-up of all other major NS$$ equipment are in Westinghouse Nuclear Services Division (WNSD) " Procedure for Setting of Major NSSS Cosponents".

Revision 2. dated February 13. 1979, and " General Reactor Vessel Setting Procedure" Revision 2. dated August 30. 1974. The NRC l

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i inspector reviewed the following drawings, which were Mkrenced in the RPV operation traveler, to verify implemt.ntation of WNSD recommendations: )

o WNSD drawing 1210E59 " Standard - Loop Plant RV Support Hardware Details and Assembly" ,

o WNSD drawing 1457F27 " Comanche Peak SES'RCS Equipment Supports -

Reactor Vessel Suppo-ts* -

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o CE drawing 10773-171-004 " General Arrangement Elevation" o CE drawing 10773-171-005 " General Arrangement Plan" Neither site prepared installation drawings nor specifications (which implementedtheWNSDrecommendedprocedurts)wereavailableandthe drawings exaaiined did not show certain specific installation _

3 criterion such as centering toleraness, levelness tolerances and clearance between support brackets and support shoes.

The inspector considers this matter unresolved. (446/8505-05)

b. Document Review The NRC inspector reviewed B&R Construction and Operation Traveler No. ME79-248-5500 which described the field instructions for installation of the Unit 2 RPV. Requirements recommended by WNSD procedures were implemented in the traveler. Worksheets attached to the traveler showed the RPV to be centered and leveled within the established tolerances. Traveler operation 19 required verification of a 0.020 to 0.005 inch clearance between the support bracket and

- - support shoe, after applying the shim plates. Change 5 subsequently changed the clearance to a 0.015 to .025 inch clearance. The installation data reflected in attachment 3B of the traveler indicated an as-built clearance of 0.012 to 0.026 inch which exeseds both the original and revised tolerances. This condition was accepted on the traveler based on Westinghouse concurrence, and there was no documented engineering evaluation onsite justifying the final tolerances. This matter is considered unresolved pending documentation validating the final installation tolerances. (446/8505-06)

The NRC inspector reviewed the following receiving records for the RPV hardware and found them to be in order:

o Report No.14322 for 54 ear.h closure studs, closure nuts, and closure washers o Report No. 09507 for vessel S/N 11713, Closure Head 11713 and 26 0-Rings 1

o Deviation notices and corrective action statementsi The NRC inspector reviewed the following completed travelers for '

internals installation and found them to be satisfactory:

o ME-84-4641-5500. " Assemble Upper Internals" o ME-84-4503-4000, " Install and Adfust Roto Locks" o ME-81-2145-5500. "Retorque UI Column Extension" o RI-80-385-5500. " Transport and Install Lower Internals" I

o ME-84-4617-5500. " Repair Lower Internals" o ME-84-4640-5500, ' Assemble Lower Internals *

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c. Visual Inspection

- At this time, visual inspection of the internals by the NRC inspector was not possible, and inspection was limited on the vessel placement to a walk-around beneath the vessel to inspect the azimuth markings I and for construction debris between the vessel and cavity. No problems were identified in this area.

d. Records of OA Audits or Surveillance The NRC inspector requested TUSCO QA audits or surveillance performed by TUGC0 of the Unit 2 RPV installation. TUGC0 did not make available any documentation of an audit or surveillance which evaluated specified placement criteria, placement procedures, hardware placement, or as-built records. This item is unresolved j

pending a more comprehensive review of these activities

. (446/8505-07).

No deviations were identified; however, two unretolved items were identified and are described in the above paragraphs. (ll.aandd)

13. Reactor Vessel Disorientation On February 20, 1979, the applicant reported to the NRC Resident Inspector that a design error had resulted in the reactor support structures being

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i placed in the wrong position on the reactor support pedestal such that the reactor would be out of position by 45 degrees. Initially. Unit 2 was to 1 be a mirror image of Unit 1 however, a design change was initiated to l pennit identical components for both units. The design change was implemented for the reactor vessel, but not for the pedestal support locations. The problem was not considered by the applicant to be

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reportable under provisions of 10 CFR Part 50.55(e) since the7 error could not have gone undetected.

l The deficiency was reported to the NRC Office of Inspection and Enforce-ment on February 22,1979 and during a March 27, 1979 meeting in Bethesda, Maryland, the applicant presented the proposed redes 19n and rework proce-dures for relocating the pedestal supports. No unresolved safety cancerns with the repair were identified at the meeting.',, ,

l During this inspection the NRC inspector reviewed various documentation relative to the disorientation prot:1em, including design changes and the construction traveler which implemented the repair.

The following documents were reviewed:

o NRC Inspection Reports 50-446/79-03; 50-446/79-07; 50-446/79-13

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.. o TUSI Conference Memo, dated March 1, 1979. H. C. Schmidt to S.

Burwell (NRC Licensing PM) o TUGC0 letter TXX-2980, dated April 30, 1979, to W. C. Seidle o NRC letter to TUGC0 dated May 29,1979 o DCA 3872 Revision 1. dated February 28, 1979,

Subject:

Rework of .

Structure for Placement of the RPV Support Shoes l

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o DCA 4122, dated March 22, 1979,

Subject:

Replacement of Rebar for RPV Supports o Construction Traveler CE79-018-5505, dated March 14,1979

Subject:

- Rework of Reactor No. 2 Cavity New RPV Support Locations

- o Grout Replacement Cards No. 007, 008, 009, 010, 014, and 015, various dates,

Subject:

Replacement of Grout around Rebar for Repair of RPV Support Shoes o Various Inspection Reports for Grout Properties and Application for RPV Support Shoes No violations or deviations were identified.

14. Reactor Coolant Pressure Boundary (RCPB) Systems .

The inspection was perfomed to verify: the applicants system for preparing, reviewing, and asintaining records for the RCPB piping and components; that selected records reflected compliance with NRC requirements and SAR conimitments for manufacture, test and installation of l

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' items; and as-built hardware was adequately marked and traceta@e to records. The following items were randomly selected and inspected: ,

a. Pressurizer Safety Valve - This item was inspected to the rw.itment stated in F5AR, Table 5.2-1 which includes ASME Secttn IJJ 'b71 Edition through Winter 1972 Addenda. Valve S/N N56964-00 W. which is installed in the B position, was inspected. Thedo11e=% records were reviewed: ,

o QA Receiving Inspection Report No. 21211 o Code Data Report Fom NV-1

c. ValveBodyCertifiedMaterialTestReports(CMTRs)

The valve was in place, however, installation had not been completed;

. therefore, the hardware installation inspection consisted of g verifying that the item was traceable to the records. ~

b. CVCS Spool Piece 301 - Requirements.for this item are stated in ASME, l 5ection III,1974 Edition through Summer 1974 Addenda, which is the comitment from the FSAR, Table 5.21. The item was field fabricated from bulk piping and purchased elbows and installed in the CVCS with field welds number 1 and 6 (ref. BRP-CS-2-RS-076). The following records were reviewed:

o BAR Code Data Report o Field Weld Data Card o NDE Reports

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o QA Receiving Reports for piping and elbows o CMTRs >

i The installed spool piece was inspected for weld quality and to '

verify that marking and traceability requirements had been met. The item had been marked with the spool piece number (3Q1) and the B&R drawing number which provided traceability to the material certifications.

c. Loop 3 RC Cold Leg - Requirements for this item are stated in ASME,Section III,1974 Edition through Summer 1974 Addenda, which is the comitment from the FSAR, Table 5.2-1. This pipin9 subassembly consists of a 27.5 inch cast pipe with a 22 degree elbow on the reactor end, a 10 inch 45 degree nozzle, a 3 inch nozzle, and three 2 ;

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1/2 inch thermowell installation bosses. The followirtshecords were I reviewed for the subassembly: ,

o QA Receiving Inspection Report No. 12389 o Westinghouse Quality Release (QRN 47523) o Code Data Report Form NPP-1 l

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o 27 1/2 inch line CMTR o 3 inch nozzle CMTR f o Field Weld Data Cards o NDE Reports (1) Sandusky Foundry and Machine Company test report for the cold -

- leg pipe certifies that material meets requirements of ASME

- Section II, 1974 editions through winter 1975. Southwest Fabrication and Welding Company code data report NPP-1 Form certified that the cold leg subassembly met requirements of ASME Section III, 1974 edition through winter 1975.

(2) The NRC inspector reviewed the proceduras and hydro test data applicable to Unit 1, since Unit 2 hydro had not been completed.

Requirements for the tests were presented in Procedures CP-QAP-12.2, " Inspection Procedure and Acceptance Criteria for ASME Pressure Testing" and CP-QAP-12.1, *ASME Section III Installation Verification, and M-5 Certification." Procedure CP-QAP-12.1 requires that a data package to be used in the test, be prepared with the test boundary and the additional following data shown:

o Base metal defects in which filler material has been added, and the depth of the base metal defect exceeds 3/8 inch or 10% of the actual thickness, whichever is less, o Untested vendor per 'n piping circumferential welds.

o Approximate location and material identification and description for permanent pressure boundary attachment with applicable support number referenced.

o Weld history, which shall reflect weld removal and/or weld repair.

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The completed hydro data package (PT-5501) for Unit;1, loop 3 cold leg was reviewed for compliance with the above requirements. Drawing No. BRP-RC-1-520-001 had been wsed to ,

annotate the test boundary. A handwritten statement on the !

drawing indicated: "No major base metal repairs could be ;

located" and *No hangers with weld attachments could be located." Welds performed by the pipe subassembly vendor, including the 22 degree circumferential weld shd the penetration fittings had not been identified. Tije following items are unresolved pending further review to determine:

o If the statement "no major base metal repairs" was based on :

a visual inspection or on a review of vendor and site inspection and repair records.

o if the shop circumferential weld attaching the 22 degree elbow to the pipe assembly was inspected during the test.

- o If welds for penetrations into pipe assembly were inspected since Procedure CP-QAP-12.1 does not require identification of such welds and they were not identified on the drawing.

The above issues will remain unresolved pending further i evaluation by the applicant (445/8507-07; 446/8505-09).

d. Personnel Q9alifications - personnef who had performed selected tasks were identified during inspection of installation records. Training and experience records for the personnel were reviewed to verify that employee qualificattor.s and maintenance of records were current and met requirements. Names or codes for five welders and two NDE examiners, who had performed tasks during installation of the items being inspected, were identified and their qualification records

- reviewed. There were no questions in this area of the inspection.

- No violations or deviations were identified.'

15. Special Plant Tours (Unit 1 and Unit 2) -

On May 23, 1985, the NRC inspector conducted a tour of selected areas of Unit 1 and Unit 2. The grou Technical Review Team (TRT) p consisted of one NRC inspector, two NRCrepresen TUEC representatives. The TUEC representatives tagged each area where a deficiency was alleged. With the alleger's consent, a tape recorder was also used to note locations and describe any alleged deficiencies. The allegers indicated that they had identified all deficiencies during the I

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tour and all other deficiencies that they had knowledge. ThehRCTRTis j analyzing this infonnation and will tecide what action, if ary.. .should be i taken.

During this tour the NRC inspector independently identified a questionable i practice in that the top of the the pipe chase at the north end of room 88 j in Unit 1, safeguards building had two large stickers whir,h stated that areas on the wall were reserved for pipe hangers GMH-51-1-58-038-006 and R1(f)1-087-X11. These stickers were dated 1980.._It was not evident whether hangers were missing or none were needed in these locations and the reserve tags were not removed. TUEC representatives were unable to answer the question immediately. This item is unresolved pending further review during a subsequent inspection. (445/8507-08).

No violations or deviations were identified.

"- 16. Routine plant Tours (Units 1 and 21 .

At various times during the inspection period NRC inspectors conducted

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general tours of the reactor building, fuel builcing, safeguards building, electrical and control building, and the turbine building. During the ,

tours, the NRC inspector observed housekeeping practices, preventive maintenance on installed equipment, ongoing construction work, and discussed various subjects with personnel engaged in work activities.

No violations or deviations were identified.

17. Review of Part 21 and 10 CFR 50.55(e) Construction Reoorts Status The NRC inspector reviewed all reports issued to date to assure that NRC and TUEC status logs were complete and up to date. A total of 183 reports have been submitted to date. This inspection priod one Part 21 report on Diesel Generator 011 Plugs and EquipmentHatchCoverandSA106 two 10 Piping CFR 50.5i(e))

(lightwall reports on the were submitted.

No violations or deviations were identified.

18. Exit Interviews TheNRC'inspectorsmetwithmembersoftheTUECstaff(denotedin paragraph 1) on May 10 and June 10,1985. The scope and findings of the inspection were discussed. The applicant acknowledged the findings.

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1 APPENDIX B U. S. NUCLEAR REGULATORY Col 9415510W i REGION IV ,[

NRC Inspection Report: 50-445/85-07 Permit: CPPR-126 )

50-446/85-05 CPPR-127 -

Docket: 50-445; 50-446 /

Applicant: Texas Utilities Electric Company (TUECI .

Skyway Tower .

400 North Olive Street I Lock Box 81 Dallas, Texas 75201 Facility Name: Comanche Peak. Steam Electric Station (CPSES)

- . Units 1 and 2 ,

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'a Inspection At: Glen Rose. Texas

- Inspection Conducted: April 1,1985, through June 21,1985 Inspectors: Je H. 5. Phillips, 5enior Resident

/0[hf~

Date Reactor Inspector Construction (pars. 1, 2, 3, 8, 9, 10, 11, 15, 16, 17, 18, and 19) d

& bk W J. E. Cumins, Senior Resident Reactor pati

/Wa./15 J Inspector Construction (April 1 - May 10,1985) ,

(pars. 1, 3, and 19) (

l D. E. Norman , Reactor Inspectort& sNrr Date (pars.1,12,13,14, and 19) bA' /W.tf23* l D. M. Hunnicutt. Section Chief Unte I Reactor Projects Branch 2 l (pars. 1, 4, 5, 6, 7, and 19) l l

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og Approved: bN /M.2/F5 U. M. Hunnicutt. 5ection Ghlef, Date Reactor project Section B '

Inspection Sumnary #

Inspection Conducted April 1.1985, throuch June *21.1985(Report 50-445/85-07) l Areas Inspected: Routine, announced and unannounced inspections of Unit 1 Which included plant tours and review of plant status, action on previous NRC inspection findings (violations / unresolved items), review of documentation for site dams, and review of 10 CFR part 21 and 10 CFR part 50.55(e) construction deficiency status, The inspection involved 77 inspectorehours psite by" four NRC inspectors.

- Q4] Q, Results: Within the areas inspected, tions were identified: fail- N ute to promptly correct an identified probles with RTE - Delta potential' Transfomer Tiltout Subassemblies, paragraph 3.a.; counertial non-shrink grout was used to grout the Unit i reactor coolant pump and steam generator supports in lieu of Class *E*. concrete, paragraph 3.b.; hydrogen recombiners out-of-specification voltage recorded on quality release document but QC receipt inspector accepted, paragraph 3.c; failure to provide objective evidence to show that central and truck mixer blades were inspected, paragraph 8; and failure to issue a deficiency report on cement scales that were out-of-calibra-tion, paragraph g.c.

Inspection Susenary Inspection Conducted April 1.1985, through June 21.1985(Report 4k6/85-05)

Areas Inspected: Routine, announced and unannounced inspections of Unit 2 which included plant tours and review of plant statuW. action on previous WRC inspection findings (violations / unresolved items), review of documentation for site dams, review of documentation for volds behind the stainigss steel cavity liner of reactor building, observation of NDE on liner plates, inspection of concrete batch plant, review of calibration laboratory records for batch plant, review of concrete laboratory testing, inspection of level C and D ,

j storage, review of reactor pressure vessel (RpV) and p ng records /coupleted I wort, and review of 10 CFR part 21 and 10 CFR part 50. el construction 1 deficiency status, and review of violation and unresolved items status. The inspection involved 335 inspector-hours onsite by four NRC is,spectors.

I l Results: Within the sixteen areas inspected five violations were identified:

failure to correct RTE-Delta transfomer probles, paragraph 3.a; failure to {

i provide objective evidence to show that concrete central and truck sixer blades !

were inspected, paragraph 8; failure to issue a deficiency report on cement I scales that were out-of-calibration, para criteria into specifications, procedures, graph Sc; failure and drawings, to translate paragraph Ita.; design and failure to maintain RPV installation tolerances / document devtations in a nonconformance report, paragraph 12b. ,

J 1

e 3-DETAILS

1. Persons Contacted Applicant Personnel f

M. McBay, Unit 2 Reactor Butiding Manager .

B. Ward, General Superintendent, Civil i D. Chandler, QA/QC Civil Ins stor l W. Cromeans, QA/QC, TUGC0 La) oratory / Civil Supervisor

fJ. Merritt, Assistant Project General Manager

fP. Halstead, Construction Site QA Manager '

' #C. Welch, QA Supervisor TUGC0 (Construction)

J. Walters, TUGC0 Mechanical Engineer K. Norman, TUGC0 Mechanical Engineer (

J. Hite, B&R Materials Engineer G. Purdy, BAR CPSES QA Manager

Denotes those present at May 10, 1985 exit interview.

fDenotes those present at June 10, 1985 exit interview.

The NRC inspectors also interviewed other applicant employees during this inspection period.

2. Plant Status Unit 1 At the time of this inspection, construction of Unit I was 99 percent !

complete. The fuel loading date for Unit 1 is pending the results of l ongoing NRC reviews.

1 Unit 2 At the time of this inspection, construction of Unit 2 was approximately 74 percent complete. Fuel loading is scheduled for approximately 18 months after Unit 1 fuel loading.

3. Applicant Action on Previous WRC Inspection Findings

a. (Closed) Unresolved Item 445/8440-02: Potential Problem with Potential Transformer T11 tout subassemblies.

i By letter dated June 15, 1983, Transamerica Delaval notified the applicant of an RTE - Delta 10 CFR Part 21 report to the NRC reporting a potential problem with the primary disconnect clips of

I

-4 the potential transformer tiltout assembly used in the emergency ,

diesel generator control panels at CPSES. The Transamerica Delaval i letter also provided instructions for correcting the problem.

However, the NRC inspector could not determine if the problem ~had been corrected at CPSES and made this an unresolved item. The applicant determined that the problem had not been corrected and subsequently performed the recommended corrective' action. The Unit 1 corrective action work activities were documented on startup work permits Z-2912 (train A) and 2-291Tl train B). The Unit 2 work activities are being tracked as master data base (M)B) item 3003-31.

The failure to promptly correct this identified problem is an apparent violation (445/8507-01; 446/8505-01).

b. (Closed) Unresolved Item 445/8416-03: Commercial Grout Used in Lieu of class T concrete The applicant determined that the use of non-shrink commercial grout in lieu of the Class "E" concrete specified on drawing 2323-51-0550 was acceptable. Design Change Authorization 21179 was issued to drawing 2323-51-0550 accepting the use of the commercial non-shrink grout. However, the failure to grout with Class "E" concrete as specified on the drawing at the time the work was accomplished is an apparent violation (445/8507-02).

c. (Closed) Unresolved Item 445/8416-04: Hydrogen Recombiners -

Out-of-Specification Voltage Recorded on Westinghouse Quality Release Document Quality Release N-41424 was revised changing the specified voltage from 10+-2V to 12+-2V which put the questionable voltage within specification limits. However, the failure of receipt inspection to verify that the QRN-41424 was filled out accurately as required by Proced9re QI-QAP7.2-8 is an apparent violation (445/8507-03).

d. (0 pen) Unresolved Item 445/8432-06; 446/8411-06; Lobbin Report Described Site surveillance Program Weaknesses During this reporting period the NRC inspector reviewed the status of this open ites several times and interviewed TUEC management and site surveillance personnel. The Lobbin report stated that the scope and objectives of the site surveillance program were unclear, lacking both purpose and direction.

There is no specific regulatory requirement to have a surveillance arogram; however, TUEC committed to have a surveillance program and gas established 1rocedures to implement such a program as a part of the 10 CFR Part 30 Appendix B, QA program. This extra effort is a strength; however, the NRC inspector also observed, as did the Lobbin l l

-- A

l l Report, that the surveillance program lacks both purpose and direction '

to be effective and complimentary to the audit and inspection programs. ;

L Since the TUEC audit group is not located on site, the TUEC surveil- :

i 1

lance program on site takes on added significance. I

)

This item was discussed with the TUEC site QC manager who described a reorganized site surveillance function and changes that have occurred. New procedures which describe this organization's duties and responsibilities are forthcoming. . ,

TUEC has elected to defer responding to the violations pertaining to the audit function in NRC Inspection Report 445/84-32; 446/84-11, but rather to have the Comanche Peak Response Team (CPRT) respond to this report and other QA matters. The surveillance issue is closely tied to the audit deficiencies in NRC Inspection Report No. 445/84-32; 446/84-11. This item will remain open pending the review and imple-mentation of the CPRT action plan. A special point of interest will be how audits and surveillance work together to evaluate the control of all safety-related activities on site to assure quality, i

especially the overview of quality control effectiveness.

4. Document Inspection of Site Dams The NRC inspector reviewed documents describing the inspection activities 3

performed on the Squaw Creek Dam (SCD) and the safe shutdown impoundment '

iSSI) for impounding cooling water for the two units at CPSES. The purpose of the SCD is to impound a cooling lake for CPSES. A secondary reservoir (SSI) is formed by a channel connecting the 500 impoundment to the SSI.

Three documented inspections have been performed since 1980. The inspections were:

a. Relevant data for SCD is contained in Phase ! Inspection, National Dam Safety Program, Squaw Creek Dam, Somervell County, Texas, Brazos River Basin, inspection by Texas Department of Water Resources.

Date of Inspection: June 10, 1980.

b. Inspection on August 25, 1982, by registered professional engineers from Mason-Johnston & Associates Inc., and Freese & Nichols, Inc.

c. Inspection on September 19, 1984, by a registered professional engineer from Mason-Johnston & Associates, Inc.

The inspection activities consisted of visual inspections by inspection teams that included accompanying Texas Utilities Service. Inc. (TUSI),

and Texas Utilities Generating Company (TUGCO) representatives.

Photographs were taken as a part ef the documentation. The data for the

l piezometer observations and the data for the surface reference monuments

~

were reviewed by applicant personnel and Mason-Johnston engineers. l No items of significance were observed or reported by these inspection teams. Slight erosion areas were observed and reported. A cracked area on the service spillway upstream right bridge seat was observed by the inspection teams and continued monitoring of this area was recommended by ,

Mason-Johnston and Associates. No signs of cracks, settlements, or horizontal movement at any location w' thin .the SCD or the SSI were reported. -

The NRC inspector reviewed the applicant's records and the Mason-Johnston inspection reports. These documents indicated that the SCD and SSI were structurally stable and that the applicant was performing inspection activities to maintain the structural integrity of these dams.

i The state of Texas requires periodic inspections of these dans (principally the SCD) due to inhabited dwellings downstream. The applicant has met these inspection requirements.

No violations or deviations were identified.

5. Voids Behind the Stainless Steel Cavity Liner in Unit 2 Reactor Building l The NRC inspector reviewed applicant records, including NCR C-5Z-01Z0Z; NCR C-1784, Rev.1; NCR C-1784, Rev. 2; NCR C-1766, Rev.1; NCR C 1791, Rev.' 1; NCR C-1824, Rev. 1; NCR C-1824. Rev. 2; Significant Deficiency Analysis Report (SDAR) - 26, dated December 12, 1979; DCA-20856; and Gibbs and Hill Specification 2323-55-18. The review of records and I documentation and discussions with various applicant personnel indicated the following:

l Structural concrete was placed in Unit 2 reactor building at elevation 819 feet 6-3/4 inches to 846 feet 6 inches on June 21, 1979. This concrete was placed adjacent to the stainless steel liner walls. The concrete forms for this pour were not removed until Detober 1979 due to subsequent concrete placements for the walls to elevation 860 feet 0 inches. When the fonas were removed, honeycombs and voids were observed by applicant personnel. The applicant's review of the extent of unconsolidated concrete resulted in the issuance of SDAR-26 on December 12, 1979. Investigations were begun and Meunow and Associates (M&A) of Charlotte, llorth i l

Carolina, were contracted to perform nondestructive testing on '

in-place concrete. M&A performed these tests on a two foot grid pattern on the compartment and liner sides of all four steam l generator (SG) compartment walls. The selected test locations did not include the locations where the voids were later found to be located; therefore, the voids were not detected during the M&A testing.

l 1

4' 4

In August 1982, preparations were made to pour the concrete annulus around the reactor vessel. When the expanded metal forework was

- removed from the reactor side of the compartment walls, voids were observed and NCR C-82-01202 was prepared. DCA 20856 was prepared as a procedure to repair the void area. DCA 20856 indicated that the voids were not extensive-(a surface area of about 28 square feet by 8 inches maximum depth) and that the repair procedure assured that' the total extent of voids had been identified. On'e half (0.5) of a cubic yard of concrete was used to c.omplete the repairs as indicated on grout pour card 261.

The applicant's review and evaluation of the gird pattern and a comparison of SG compartments 2 and 3 to 1 and 4 indicated that voids did not exist in SG compartments 2 and 3. The review of test girds extended down to elevation 834 feet, which is the floor elevation of the liner. The liner walls of SG compartments 1 and 4 were not tested at elevation 834 feet, but at elevation 836 feet which is above the area of the identified voids. No testing was done on the liner side of the area of the voids below elevation 836 feet. The program also included removal of 2 inch x 2 inch plugs from the stainless steel liner at locations where test ir.dications raised questions concerning the concrete. The inspections of the concrete by applicant personnel after the plugs were removed confirmed that there were no additional unconsolidated concrete areas (voids).

. The applicant removed stainless steel liner plates from three areas (one area about 1 foot by 1 1/2 feet and two areas about 3 feet by 1 foot, excavated or chipped to sound concrete, and cleaned the concrete surface area, One and one-quarter inch (11/4) diameter probe holes and grout access holes were drilled in the liner plates to determine the extent of and to assure full definition of the void area. Air access holes were drilled in the stainless steel liner -

plates to assure that grouting would be accomplished in accordance j with the procedure, j The procedure (DCA-20856) specifed that the grout was to be cured for 28 days or until the grout reached a compressive strength of 4000 psi. Repairs to the liner plates were specified in DCA-20856 and G4H Procedure 2323-55-18.

l of the I DCA-20856 required liner across weld seams,that under across no circumstances embedded weld plates,was or cuttinfnto leak ,

chase seal welds or drilling through the liner at leak chase channels, I embeds, or weld seams pemitted. Documentation review indicated that DCA-20856 was adhered to and that no cutting or drilling occurred in prohibited locations.

No violations or deviations were identified.

6. Nondestructive Testing Observations of Liner Plates in Fuel Transfer Canal i The NRC inspector observed portions of non-Q liquid penetrant examinations (PT) being performed on liner plate welds following re-installation of the liner plates in the areas of the fuel transfer canal removed for inspection and repair of the concrete. The inspector perfomed the PT on the welds as required by the repair package and the procedure (QI-QP-11.18-1, l

' Liquid Penetrant Examination"). Scattered weld porosity was identified by the inspection. The porosity was ground o'ut and a repeat PT was performed. The final inspection is scheduled to be performed by QC inspection personnel. The liner plate areas to be inspected by PT were identified in DCA 20856.

No violations or deviations were identified.

7. Cadwell Splice Observations and Records i

4. Calibration of Tensile Tester The NRC inspector observed the calibration of the Tinus-Olson Universal Testing Machine (Model Number 600-12 Identification Number M&TE-784) on April 2 and May 7,1985. The machine was calibrated just prior to performing tensile testing of cadweld splices and subsequent to completion of tensile testing each day that tensile testing was performed. The machine calibration date for April 2, 1985, prior to start of tensile testing was observed by the NRC inspector and recorded as follows:

Nominal load Calibration Reading Error Error Remarks (1bs) (1bs) (1bs) 5 ,

1 0 0 0 0 0 machine on !

4/2/85 !

100,000 99,750 +250 +0.25 !

200,000 199,600 +400 +0.2 300,00 299,450 +550 +0.18 350,000 350,300 -300 -0.08 400,000 401,200 -1200 -0.03 500,000 501,350 -1350 -0.27 l 600,000 602,450 -2450 -0.40

w i

\

9 The NRC inspector reviewed calibr& tion data for March 4, March 8, April 2, April 3, April 30 and May 7,1985. All calibration data set within the +/- 15 accuracy requirement specified by Calibration Procedure 35-1195-1El-37 Revision 3, dated March 11,1982. The reference standards were identified as follows:

! ID No. Manufacturer Calibration Due Date RS-75 SLH Electronics January 27, 1987 RS-75.3 BLH Electronics January 27, 1987 /

b. Observation of Cadweld Splice Tensile Testing (1) Qualification Tensile Testing

, On April 2,1985, the NRC inspector observed the following tensile testing of cadweld splices for cadwelder qualification:

EBD Q8, GBH Q1, GBH Q2 GBV Q1, SFD 04, SFD Q3, SFM IM, GAH Q? ,

GAV Q1, and GBV Q2.

Each of the above qualification cadweld splices was tensile tested to 400,000 pounds (100.000 psi) and met the requirements stated in the procedure.

(2) Prrtuction Tensile Testing The WRC inspector observed the tensile tester calibrations and the following production cadweld splices tensile testing on May 7, 1985: FXD 3P, FYD 4P, FTD 8P, FRD 87P, and FUD 6P.

Each of the above production cadweld splices was tested to 400,0C0 pounds (100,000 psiland met the requirements statd in

. the procedure.

(3) Installation of Production Cadweld Splices The NRC inspector observed installation of rebar and cadweld splices at frequent intervals (five or more observations per week during the weeks of April 8 and 15; May 6,13, 20, and 27; and June 3,1985). The re)ar installation for the Unit 2 closure was performed in the area identified as elevation 805 feet to elevation 875 feet and azimuth 300 degrees to 335 degrees. The installation activities observed included rebar spacing location of cadwelds, observation of selection and removal,for testing of cadweld splices for testing, and determination of location of rebars and cadwelds for the as-built drawings.

9

._--__---_-._.a

(4) Documentation Reviewed The NRC inspector reviewed the following documentation for the rebar placement and cadwelding for the Unit 2 containment :

(reactor building) closure area:

Drawings DCAs NCRS i

2323-S-0785, Rev.7 22616, Rev. 1 C85-200294 2323-S-0786, Rev.9 22728- -

C85-200339, Rev.1 2323-51-500, Rev.5 22737 C85-200355, Rev.1 2323-S1-506 Rev.5 22836 2323-52 505, Rev.5 22878 (Sheets 1-7) 2323-52-508. Rev.2 22772 2323-52-506, Rev.3 No violations or deviations were identified.

8. Concrete Batch Plant Inspection. Unit I and 2 The NRC inspector used a nationally recognized checklist to inspect the concrete production facilities. This list included the specific characteristics for the following areas: (1) saterial storage and handling of cement, aggregate, water and admixture, (2) batching equipment scales, weighing systems, admixture dispenser, and recorders, (3) central mixer (not ap ticketing system,'and (5)plicable delivery because system. it had been dismanteled), (4)

The current batching is a manual operation since almost all concrete has been placed. The central mixer was dismanteled and removed from site two

-or three years ago when concrete placement was virtually completed.

Presently, the backup batch plant (which was a backup system for the central mixer) is in operation to complete the remaining concrete placements. This batch plant is in good condition and complied with the subject checklist except for one area.

The NRC inspector inspected the inside of one of three trucks used for mixing concrete (that is, the batch plant dispenses the correct weight of materials as required by the specific design six numbers and the truck then mixes the batch to be placed.) The bhades inside the truck are <

subject to wear and should He checked at a reasonable frequency. The Brown & Root (84R) representative responsible for checking the blades in accordance with B&R Procedure 35-1195-CCP-10 Revision 5 dated December 4,1978, was asked for evidence that the blades had been checked ,

for wear on a quarterly basis and it was found that there was no record of I such checks dating back to 1977 when they were initially checked.

L Procedure CCP-10, paragra ph 3.10 " Truck Mixing *, is silent on blade wear

! but Section 3.11 infers t1at the blades should be checked for both l central and truck mixing. The inspection of both central and truck i

l

mixing blades was not documented, although the B&R representative stated that the mixing blades were periodically inspected and laboratory testitig would have probably indicated if there was a problem with the mixing blades.

Strength and uniformity tests have consistently been within the acceptable range indicating that concrete production was acceptable even though mixing blade inspection was not documented.

Otherwise,_the condition of the inside of the truck was satisfactory as i the drum and charging / discharging were clean.* The water gage and drum '

counter were in good condition.

This failure to follow procedures is a violation of 10 CFR 50, Appendix B, Criterion V. Subsequent to the identification of this violation, the blades were checked for wear and blade wear was presently within allowable limits (445/8507-04; 446/8505-02).

No other violations or deviations were identified.

9. Calibration Laboratory for 8atch Plant Unit 1 and 2 The NRC inspector obtained batch plant scale numbers from tags which indicated that the scales had been calibrated and were within the calibration frequency. Cement (MTE 779), Water (MTE 766), admixture scale (MTE 764), and aggregate (MTE 780) were reviewed. The scales had been periodically calibrated since the batch plant was activated. The records were adequate except as follows:

a. Scales NTE 766 records do not clearly differentiate between the ,

required accuracy of the scale and the digital readout.

b. Scales MTE 779 and 780 records show various accuracy ranges for the same scale; i.e., MTE 779 (SN749687) records the following: report l dated January 1976 gives 15; report dated July 1976 gives 15 while the report dated October 1976 gives +/- 0.25.

the above items are unre-The calibration solved pending appeared to beofproper, further review however,s actions regarding the the applicant correction of these records (445/8507-05; 446/8505-03).

c. Records for scales MTE 779 records contained B&R meno IM-1108 dated July 16,1975, which described a nonconforming condition. This condition affected the water and cement scales causing a 24-48 pound deviation during the calibration test. The meno stated that the condition was corrected and the scales were then calibrated; however, no deficiency report was written as required by SSR Procedure CP-QCP-1.3, " Tool and Equipment Calibration and Tool Control" dated July 14, 1975, and CP-QAP-15.1,

Field Control of Nonconforming Items," dated July 14, 1975. As a re: ult there is no evidence that corrective action included an evaluation to determine if concrete production was adversely affected.

e .'

l 1

This failure to assure that a nonconforming condition was evaluated is a violation of Criterion XV of 10 CFR Part 50, Appendix B, (445/8507-06;-446/8505-04).

10. Concrete Laboratory Testing Units 1. and 2 TUGC0 procedure QI-QP-11.1-1, Revision 6, was compar$1 with ASIE Section Ill, Division 2. Subsections 5222, 1223 and 5224 to assure that

)

each ASTN testing requirement was incorporated into the procedure.

The NRC inspector inspected the testing laboratory equipment and feitid ]

the test area and equipment were in good condition and each piece of equipment was tagged with a calibration sticker which showed it to lie I within the required calibration frequency. Test personnel were knowledge-able of test requirements and equipment.

The NRC inspector witnessed field tests performed by laboratory personnel !

as follows:

Date Truck No. Mix No. Ticket No. Air Content (5) Slump (in.) Temp (*F) 6/3/85 RT-41 925 64013 Req 8.2 10.3 NA 70 max Mea 8.7-9.1 NA 57 6/3/85RT-35 128 64014 Req 5.0-7.0 5 max 70 max Mea 6.6 6.25* 57

Truck was rejected by quality control but was later accepted when second slump reading came into required range.

The following laboratory equipment was checked and found to be within calibration: Forney Compression Tester, NTE 3031; Temperature Recorder NTE 3013 and 3014; Unit Volume Scale, NTE 1053; Pressure Meters NTE i 3000B, 3002 and 3004; Steves MTE 1286, 1239,1272,1274,1136A,1156, l 1094,1093,1095,1178,1179,1300 and 1180; Aggregate scales, IET 1058 l and 1067; and 2" grout sold NTE 1111.

The following test records for placement number 201-5805-034 were reviewed: (1) concrete placement inspection, (2) concrete placement summary and, (3) unit weight of fresh concrete.

No viciations or deviations were identified. l

11. Inspection of Level C and D Storane Unit I and 2 The NRC inspector inspected all laydown aMas where piping, electrical j conduit, cable, and structural reinfon:ing steel were stored. These i

materials were neatly stored outside on cribbing in well drained areas I

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which allowed air circulation and avoided trapping water. This met the Level "D" storage requirements of ANSI N45.2.2.

The electrical warehouse contained miscellaneous electrical hardware.

This building was required to be fire and tear resistant, weathertight, and !

well ventilated in order to meet Level "C" storage requirements. This warehouse was well kept and met all requirements except for a lock storage area l located upstairs at the rear of this buildinn (electri' cal termination q tool room). Two minor problems were identif' ed and the warehouse personnel initiated action to correct them. * -

The first problem noted was that a box of nuclear grade cement was marked

" shelf life out of date' but it had no hold tag. The box was subsequently tagged with Nonconformance Report (NCR) E85-200453 after being identified by the NRC. During discussions with the warehouseman, the NRC determined that engineering told the warehouseman to mark the material and lock it up, but did not tell him to apply an NCR or hold tag. TUEC should determine if engineering is aware of nonconforming material controls and provide training if this is other than an isolated instance. Also, the NRC inspector noted a very small leak in the roof above the electrical termination tool room. This leak was in an area that did not expose hardware to' moisture. The roof is currently being repaired. i The millwright warehouse storage area was inspected; however, only a i small number of items or materials were stored in this area. The overall storage conditions in this area met or exceeded Level 'C' storage requirements. '

I No violations or deviations were identified. I

12. . Reactor Pressure Vessel and Internals Installation - Unit 2 This inspection was perfonned by an NRC inspector to verify final placement of the reactor pressure vessel (RPV) and internals by examining the completed installation and inspection records.

a. Requirements for Placement of RPV Requirements for placement of the RPV to ensure proper fit-up of all other major NSSS equipment are in Westinghouse Nuclear Services Division ( WSD) " Procedure for Setting of Major NSSS Components",

Revision 2, dated February 13, 1979, and " General Reactor Vessel Setting Procedure

Revision 2, dated August 30, 1974. The NRC inspector reviewed the following drawings, stitch were referenced in the RPV operation traveler, to verify implementation of WSD j recommendations

)

o WSD drawing 1210E59 ' Standard - Loop Plant RV Support Hardware Details and Assembly" l o WSD drawing 1457F27

Comanche Peak SES RCS Equipment Supports

- Reactor Vessel Supports" i

i I

~-.- .

o CE drawing 10773-171-004 ' General Arrangement Elevation

i o CE drawing 10773-171-005 " General Arrangement Plan

Neither site prepared installation drawings nor specifications (which implemented the WNSD recommended procedures) were available and the drawings examined did not show certain specific installation criterion such as centering tolerances, . levelness tolerances and

. clearance between support brackets and support shoes. The lack of engineering documentation did not provide full control of the action and would allow changes to installation criteria important to i safety to be made without complying with established c%nge procedures.

l 4 gr This is & violation of 10 CFR 50, Appendix B, Criterion III (446/8505-05). .

b. Document Review &~&A*3-*?

The NRC inspector reviewed 88R Construction and Operation Traveler ,

No. ME79-248-5500 which described the field instructions for installation of the Unit 2 RPV. Requirements recommended by WNSD procedures were implemented in the traveler. Worksheets attached to 7 the traveler showed the RPV to be centered and leveled within the 5 established tolerances. Traveler operation 19 required verification i of a 0.020 to 0.005 inch clearance between the support bracket and l support shoe, after applying the shim plates. Change 5 subsequently changed the clearance to a 0.015 to .025 inch clearance. The installation data reflected in attachment 38 of the traveler indicated an as-butit clearance of 0.012 to 0.026 inch which exceeds both the original and revised tolerances. This conditipn was accented g on the traveler based on Westinghouse concurrence, andNere were neither nonconfomance reports nor documented engineering evaluations to determine if the condition was acceptable. This failure to document violation of nonconforming conditions 10 CFR 50, Appendix and engineering B, Criterion XV 446/8505(deviations 06) is a The NRC inspector reviewed the following receiving records for RPV hardware and found them to be in order:

o Report No.14322 for 54 each closure studs, closure nuts, and closure washers o Report No. 09507 for vessel S/N 11713, Closure Head 11713 and 26 0-Rings o Deviation notices and corrective action statements l

l

t The NRC inspector reviewed the following completed travelers for internals installation and found them to be satisfactory:

o ME-84-4641-5500 " Assemble Upper Internals

o ME-84-4503-4000, " Install and Adjust Roto Locks" o ME-81-2145-5500, "Retorque UI Column Extension' o RI-80-385-5500, " Transport and Ins' tall Lower Internals" o ME-84-4617-5500, " Repair Lower Internals" o ME-84-4640-5500 " Assemble Lower Internals *

c. Visual Inspection At this time, visual inspection of the internals by the NRC inspector was not possible, and inspection was limited on the vessel placensent to a walk-around beneath the vessel to inspect the azimuth markings and for construction debris between the vessel and cavity. No problems were identified in this area.

d. Records of QA Audits or Surveillance The NRC inspector requested TUGC0 QA audits or surveillance performed by TUGC0 of the Unit 2 RPV installation. TUGC0 did not make available any documentation of an audit or surveillance which evaluated speci-fled placement criteria, placement procedures, hardware placement, or as-built records. This ites is unresolved pending a more comprehen-sive review of these activities (446/8505-07).

No deviations were identified; however, two violations were identified and are described in the above paragraphs.

13. Reactor Vessel Disorientation On February 20, 1979, the applicant reported to the WRC Resident Inspector that a design error had resulted in the ruactor support structures being placed in the wrong position on the reactor support pedestal such that the reactor would t>e out of position by 45 degrees.

Initially. Unit 2 was to be a mirror image of Unit 1 however, a design change was initiated to pemit identical components for both units. The design change was implemented for the reactor vessel, but not for the pedestal support locations. The probles was not considered by the applicant to be reportable under provisions of 10 CFR Part 50.55(e) since the error could not have gone undetected.

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The deficiency was reported to the NRC Office of Inspection and Enforce-ment on February 22, 1979 and during a March 27, 1979 meeting in Bethesda, 1 Maryland, the applicant presented the proposed redesign and rework proce-dures for relocating the pedestal supports. No unresolved safety concerns with the repair were identified at the meeting.

During this inspection the NRC inspector reviewed varipus documentation relative to the disorientation probles, including desigt. changes and the construction traveler which implemented the'rmpair.

The following documents were reviewed:

o NRC Inspection Reports 50-446/79-03; 50-446/79-07; 50-446/79-13 o TUSI Conference Memo, dated March 1, 1979, H. C. Schmidt to S. Burwell (NRC Licensing PM) o TUGC0 letter TXX-2980, dated April 30, 1979, to W. C. Seidle o NRC letter to TUGC0 dated May 29, 1979 o DCA 3872, Revision 1 dated February 28, 1979,

Subject:

Rework of Structure for Placement of the RPV Support Shoes o DCA 4122, dated March 22, 1979,

Subject:

Replacement of Rebar for RPV Supports o Construction Traveler CE79-018-5505, dated March 14, 1979,

Subject:

Rework of Reactor No. 2 Cavity - New RPV Support t.ocations o Grout Replacement Cards No. 007, 008, 009, 010, 014, and 015, various dates,

Subject:

Replacement of Grout around Rebar for Repair of RPV Support Shoes o Various Inspection Reports for Grout Properties and Application for RPV Support Shoes No violations or deviations were identified. l

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14. Reactor Coolant Pressure Boundary (RCPB) Systems [ C' The inspection was performed to verify: the applicants system for preparing, reviewing, and maintaining records for the RCPB piping and 1 components; that selected records reflected compliance with NRC !

requirements and SAR commitments for snufacture, test and installation J l

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of items; and as-built hardware was adequately marked and traceable to l records. The following items were randomly selected and inspected: i

a. Pressurizer Safety Valve - This item was inspected to the coenitment k stated in F5AR, Table 5.2-1 tdhich includes ASME Section !!!,1971 Edition through Winter 1972 Addenda. Valve S/N N56964-00-007, which is installed in the 8 position was inspected. The following records were reviewed:

l c QA Receiving Inspection Report No. 21211 o Code Data Report Form NV-1 o Valve Body CMTR h1 The valve was in place, however, installation had not been completed; therefore, the hardware installation inspection consisted of verifying that the item was traceable to the records.

b. CYCS Spool Piece 301 - Requirements for this item are stated in ASME,Section III,1974 Edition through Summer 1974 Addenda, which is the commitment from the FSAR, Table 5.2-1. The ites was field 1 fabricated from bulk piping and purchased elbows and installed in the !

CVCS with field welds number 1 and 6 (ref. BRP-CS-2-RB-076). The following records were reviewed; o B&R Code Data Report o Field Weld Data Card I

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o NDE Reports o QA Receiving Reports for piping and elbows I

o Certified Material Test Reports (CMTR)

The installed spool piece was inspected for weld quality and to l verify that marking and traceability requirements had been met. The l item had been matted with the spool tecenumber(3Q1)andtheBAR !

drawing number which provided tracea 111ty to the material certifications.

c. Loo? 3 RC Cold Leg - Requirements for this item are stated in ASME, i

sec;fon III, IU4 Edition through Summer 1974 Addenda, which is the commitment from the FSAR, Table 5.2-1. This piping subassembly

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consists of a 27.5 inch cast pipe with a 22 degree elbow on the reactor end, a 10 inch 45 degree nozzle, a 3 inch nozzle, and three 21/2 inch thermowell installation bosses. The following records were reviewed for the subassembly:

o QA Receiving Inspection Report No. 12389 o Westinghouse Quality Release (QRN 47523)

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o Code Data Report Form NPP-1 o 27 1/2 inch ifne CMTR o 3 inch nozzle CMTR o Field Weld Data Cards o NDE Reports (1) Sandusky Foundary and Machine Company test report for the cold leg pipe certifies that material meets requirements of ASME Section !!, 1974 editions through winter 1975. Southwest Fabrication and Welding Company code data report NPP-1 Form certified that the cold leg subassentaly met requirements of ASME Section III, 1974 edition through winter 1975. The FSAR comitment is ASME Section 111, 1974 edition through summer 1974. This discrepancy is unresolved pending the applicant's evaluation to determine if material nonconformances exist (446/8505-08).

(2) The NRC inspector reviewed the procedures and hydro test data applicable to Unit 1, since Unit 2 hydro had not been completed.

Requirements for the tests were presented in Procedures l CP-QAP-12.2, ' Inspection Procedure and Acceptance Criteria for ASME Pressure Testing" and CP-QAP-12.1, *ASME Section 111 Installation, Verification, and N-5 Certification.' Procedure CP-QAP-12.1 requires that a data package to be used in the test, be prepared with the test boundary and the additional following data shown:

, o Baree metal defects in which filler material has been added, and the depth of the base metal defect exceeds 3/8 inch or 10% of the actual thickness, whichever is less.

c Untested vendor performed piping circumferential welds.

l o Approximate location and material identification and description for permanent pressure boundary attachment with applicable support number referenced.

' o Weld history, which shall reflect weld removal and/or weld repair.

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The completed hydro data package (PT-5501) for Unit 1, loop 3 cold leg was reviewed for compliance with the above requirements. Drawing No. BRP-RC-1-520 001 had been used to annotate the test boundary. A handwritten statement on the

. drawing" indicated: "No major base metal repairs could be l located and "No hangers with weld attachments could be j located." Welds performed by the pipe subass'embly vendor, l including the 22 degree circumferential weld and the i penetration fittings % d not been identified. The following 1 items are unresolved pending further review to determine: )

o If "no major base metal repairs" was based on a visual '

inspection or on a review of vendor and site inspection j e.nd repair records.

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o If the shop circumferential weld attaching the 22 degree l

. elbow to the pipe assembly was inspected during the test. i o If welds for penetrations into pipe assembly were inspected '

as procedure CP-0AP-12.1 does not require identification of such welds cnd they were not identified on the drawing.

The above issues will remain unresolved pending further evaluation by the applicant (445/8507-07; 446/8505-09).

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d. Personnel Qualifications - Personnel who had performed selected tasks were identified during inspection of installation records. Training, and experience records for the personnel were reviewed to verify that employee qualifications and maintenance of records were current and met requirements. Names or codes for five welders an.d tw NDE examiners, who had performed tasks durtag installation ni the items being inspected, were identified and their qualification re:ords j reviewed. There were no questions in this area of the inspection.

No violations or deviations were identified.

15. Special Plant Tours (Unit 1 and Unit 2) i On May 23, 1985, the NRC inspector conducted a tour of selv.ted areas of Unit 1 and Unit 2. The group consisted of one NRC inspector, two NAC !

Technical Review Team (TRT) representatives, two allegers, ar.d several )

TUEC representatives. The TUEC representatives tagged each area where a 1 deficiency was alleged. With the alleger's Sonsent, a tape recorder was '

also used to note locations and describe any alleged deficiencies. The i- allegers indicated that they had identified all deficiencies during the tour and all other deficiencies that they had knowledge. The NRC TRT is analyzing this information and will decide what action, if any, should be i taken. /

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During this tour. the NRC -inspector independently identified a questionable practice in that the top of the the. pipe chase at the north end of room 88 l in Unit 1, safeguards building had two large stickers which stated that areas on the wall were reserved for pipe hangers GHH-S1-1-58-038-006 and R1( ?)1-087-X11. These stickers were de,ted 1980. It was not evident whether hangers were missing or none were needed in these locations and the reserve tags were not removed. TUEC representatives were unable to answer the question innediately. This item.is unresolved pending further review during a subsequent inspection. -(445/8507-08).

No violations or deviations were identified.

16. Routine Plant Tours (Units 1 and 2)

At various times during the inspection period NRC inspectors conducted general tours of the reactor building, fuel butiding, safeguards building, electrical and control building, and the turbine building. 1 During the tours, the MRC inspector observed housekeeping practices, j preventive maintenance on installed equipment, ongoing construction work, and discussed various subjects with personnel engaged in work activities.

No violations or deviations were identified. .

17. Review of Part 21 and 10 CFR 50.C5(e) Construction Reports Status The NRC inspector reviewed all reports issued to date to assure that NRC and TUEC status logs were complete and up to date. A total of 183 reports have been submitted to date. This inspection period one Part 21 report on Diesel Generator 011 Plugs and two 10 CFR 50.55(e) reports on the Equipment Hatch Cover and SA106 Piping (light wall) were submitted.

No violations or deviations were identified.

18. Review of Violation and Unresolved Ites Status The NRC inspector reviewed all violations and unresolved items reported to date to assure that NRC and TUEC status logs were complete and up to I date. Two hundred nineteen itees 9ere reviewed. In addition, a trend analysis of NRC findings was perfomed to generally determine how many findings could be broadly classified under each criterion of 10 CFR, Part $0, Appendix B. The frequency of findings showed broad and general trends under the following criteria: II. QA Program; !!1. Design Control; V. Instructions, Procedures and Drawings; VII. Control of Purchased Material, Equipment and Services; IX. Control of Special Processes; X.

Inspection; XI. Test control; XI!!. Handling Storage and Shipping; XVII. )

QA Records; and XVIII Audits. The most significant trends were Criterion III, V, VII, IX, X, and XVIII. Also, a number of violations occurred with respect to 10 CFR 50.55(e) items.

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These findings mainly pertained to Unit 1 and related closely to trends identified by the NRC Technical Review Team TRT. These trends will be considered during followup on TRT findings. Also, Unit 2 inspection emphasis will consider these trends during future inspections. 4 No violations or deviations were identified.

19. Exit Interviews ,,

The NRC inspectors met with members of the TUE'C' staff (denoted in paragraph 1) on May 10 and June 10; 1985. The scope and findings of the inspection were discussed. The applicant acknowledged the findings.

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ML20238A390

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