ML19329D527

Requests Mod of Recently Revised Submission Date for Hurricane Data.Notifies of Reluctance to Accept Proprietary Matl as Basis for AEC Conclusions

ML19329D527
Person / Time
Site:	Crystal River
Issue date:	04/13/1973
From:	Deyoung R US ATOMIC ENERGY COMMISSION (AEC)
To:	Rodgers J FLORIDA POWER CORP.
Shared Package
ML19329D528	List: ML19329D527
References
NUDOCS 8003160127
Download: ML19329D527 (2)

Text

._

s Qi.,7.qm h '

,'~,". ?-

VQ{mM MK%?)Q1,'pty p'.w%k."'t:y*.: +. -. - >

.g y v,,sq w y-Q f;y &; g -

.w eeF 1(.,,:R; w*d JpyQ G M,7;b d'

Nf q$7'; m i J

., ', ~

.p-

~.. ', _'

%y & Q k $ $g J.

0 %Q.i.?.

g. n; 4 f L,'

y;,. %:> ::s...yQpf%

&yMu s

.;g: yQ

m;,%.,

v. -%

+. -..~,

i;

,, o

- ' :n. g

.,.'.g.,.

v:

g w

a Docket No. 50-302 t

J 5

Florida Power Corporaties

,4 ATIM:

Mr..J. T. Badgere 3.>

i

, ~,.f

..n, C:%.

Assistaat Yise Pr'esident &

'/,']

q g '.T h' Eseletr Project fasenger

l D' N

N F. O. Dez 14642

%dW l

l

> - 7 'h ec St. Petersburg, Florida 3h33.

^

Q 7)'I 7 c...s.n v

.x,,. _ e r y qt;,3

.. +

.+.. -

.s y p. f.-

E.'-

r; E.Q.);jjj&.k[2 Centlemen:

W -

m J,u..:.:

@.Q[.

_ Q:g y;

_, _ %p.

.Q f.';Qnf V,~ 1 -/

You are see of several appliaasts who are' utilising'the services of Demse ' fr%.

and Moore to establish hericase-related design requirements for maalear power plaats. At a meeting on Febenery 15, 1973, the Begulatory staff

. agreed to evaluate the Bemes,and Meere horricane surge esiculatismal_

model. To assist in this evalasties, the Regulatory samff agreed,te ferward to Demse and'Meere data es four hurricanes.' Thee. data'is'ess2 N,, 4 M p '

~

~

A idd tained in the open files of the U. 5. Army Ceestal N " ' ing Reseereh[ f g

Center (CERC).

,The Eagulatory staff agreed to forward these CERC data in

,;ever form they =-4"ated. in the filam., i.,91m.ee the aster 1AL 9er he.rrim,eme.N,M.m# '

~

9

~

..n a.

Camille was in digis4==d - fore, this has led, appare; ntly,ise sha:=i==aa-standtes

'f f-

'C/9 that the other meterial to be forwarded would be in digitized form.

+

.s,

e, e

1 We wish to infera you that as of March 30, 1973, all'ef the data en the 4 ;.,

.;f/

fear hurricanes has-been transmitted to. Dames and Mee.re.,.as. agreed, in.. t.he.s'.,c4y,'a sters in which it, eminded in the CERl files. With respect to the Demos med' V...C' Moore questiana,eenamined in the attme h==* to your Merah 28?1973 letter, 2%[

I

/ our answers have mise been provided to Bemes and Moose... We understand from ' C, '

~

your letter of Moreh! 24, J873 that you uill.. require until' July,7, '1973. to.wjf reduse' the data if thiedfiles 'and' to respond tofeeirMafeemmelialioque$t"d@I.

I't of. March 12, -1973

-3 bis, esimy la 'yesir==6a==4==. fame"thiusy 11y1973 we previamely specifiind teMeet umfortsmate' frem SE E 5 Maf 'as'de==

'[D

~

f review eeshedules". 'We urge 'yes 'te attempt to improve your%hd==4= dase. Mj Ql

., Y %.? "

'~ !

' J hY Q: ' 7. ;h N sS5 Fleese be assured,that the staff has me assertainty,as.-to its posiGua om. thel }

horriasse-related desisslrequir====*= fer..the Crystal River Dmit'3 Flaat( g12,'19,73 le

?

These requirements are set.forth in Eselosure 1 of eer March Your assumpties that calculatiem of hurricane eurge hydrographs by CIRC andl "

AEC implies that other hurriesse data has already been digitised is incorrect.

Calculation of historical hurricane surge levels by CERC has traditionally been dens by hand, a tae kique available to you and your censultants as sell

)

se to CEEC.

'.w,3;.)Y' F

?

,x.

n., q :

u

's --

^

~' Al OmCE >

~

'9".

p sumur >

ac a.

M N N b.'.,

%[. y$.hib '4 S

_.,..:..:....;.m.. :. :.'v* }$ 0 6

h. c

• hk $

, s'f[". :'.[w: w. s.n'lN d![y* ?'+

h..gfg. : '

,. 3Y. x_y y -w my m e onre >

,"6dioo,vmimurr. PRDf.riN' iOhrkdb 12iG-OY Md,pu' "a'M "~.8 h gs." " 2

,..ror= Atc-sis (R,ev: 9-53,) AE.cM' 0240i'..s m;3 m a:

s-.

G MD 4%N g y gg.%g.9.

w

~m

~

m%

g

,s 8 0 031601 R.7 f}-

., m.m., gy m; y,~ g m ec.wyv:.a s yw w.pa : m +*yc6 : y+(: :

W

. 7. sg a yx

~

,. :. 4 r. &

).y

.., s

/ w

(.J %! Q t;?m :x../

. mGi;;;p *,$

."r w&N.Q M. ?

..%m yp s. '

ap;:% ;.,.3 %u

-Md:

%w w c )}p.aea,dn j W:

a

/w-a v K. 7*s. 9."% m.

!~a L % c: g

.,, 4 m.m...

]V.Q. c.Q. v.Mgn

^ ^

y ; 'T ?

' =

,: e

- ~

s l

Florida Power Corporation p

~ g j g gg

'].

[

]

vi >, q..

s j

I You should understand that although the staff has agreed to consider 1

the Dames and Moore hurricane surge calculational model, we are very reluctant to accept, as a basis for staff conclusione, proprietary material which anast be kept from the public record of an application.

We believe,that the publia interest,is.best served by having all of.

... K '

.thetechnies(basesfor;shestaff,~eemolusioneenanappliestiesonCO,b.h;~

r 6the weeerd and - aW 1=Ma

' Qbinomifaiemanifeffour re$s,fof publia Siew'.J Bemes and Meere has('also'l NMd.

4 m x w w:w guya m p. p m p m.,.,m;[,3fg!,@J & %, k Q.

ita M M shis'madeer mf ::m: Qfh'M

)

p N. & M

"^[;%f %y[yQ&%RW%&sidade$i,u$[Y.&G%%%V.W h. p M h,h. h h;%i M f.'" '. $

[' ^h4 y M $ $ Q d Q.peYoung Q,,' [CSM S M&#W:%

e

Q..

t./ ;WM:yc Q

AM

\\

a f dW..hthYh..o@M -LWhM. U.M.g.

i

[; a N'h ';h.4.w%f.D.h'M(jjihgk.$g M,, %(cQ

.W

%..ry:v;& &Y g

-4

- g

< ~ ;,. %

f r.

A &.w*qu p ~.9., y c x %WV

.8

'W v

g c

e r4 m 4.. s i

kIDWe,deeA844Et,, _,BSEh,,,;

'$se Pre ls'ourisek Bater Beemasse g'%g @@M, l

.y)

. O L t

. RC".M ' WS,& p;,il'; 91/.

D,

.y [

,g

~-

1 A e

..a,

s

e. s....

4

(...

m. ;yn<~.m n. me i

1. .,,A.,f f e*

• SR;,A...Brandimone p pg, g f,4,Q y g q f.;%,'. j,y $.[A. M @,m.s n,n %g

,wo.

F C WJY;0 M;.i. M g" "~ {C s.e g,p iMeeJoneadesaA

%M 9-n.$ MNs@h!*E6 % & ? & $4 '

fC:

'% W M

q

~

GNetF3Oeirabeg%

i hM M-('

aJJ kl mg.+s m

.w kg m4Q l

y'f

.w.064 M%g<:.w? n n m,n

f.. 1

'9.DiISTRIBUTION Z N, n :,,

y

/

M-@

WW m #./

He

..a-e

. s iAECiEDR N'ja &,o g. 4%w &,.

-- w p Q,: Q %a. m w e %.u-v 5..;;~.pw' w y.pg..

nw.

w v

r" i f y-wm3 y

y Local.:PDR. n,. :.n.

mw

.p

~ $m +;& $e.++a-M w ~ m, s% n,8 Q M.i N M M 5 ?,b ' l - T;.c.O* 9' Z T $s'h NpNI

+.a x

e

.e -.fr a

Docket, File > "EJW S# ' Y #5 W M"

' J. W.a.c h "d '! G M A% W.A'N<s*&

?

W ~ MUNN RP RdI D>t &

%.R N uC^4 p h,3 MM M.

RCDei.oung

.;x jg.g 43p{ggygg yg y gpg g g gQf-ggg %p'.g - f '

g :.3. p.

gys43 yq;p, s.

.-PWR;Branc%Chiefg ggf RWKlecker

.g.gg%'I Nd, y?M 4 ' p. % cs m...g; y

O ggg,3 q.

g;;-

gg

'BCBuckle'y mfw 5W

u. y g -,y. s aw 3 -qv p

W;M N ;M @. W O

OM g

EIGou bourne

o.q k M M'k(Nh; M

. kh Ih 7PWR -4 gh hh, Nk,J.%v%. m. n%)g Qu?y, k,p:w,W Ag'4. w.

a

" h, + nf a

.,,, M, n.

dr.ief S

u

., w w

gg+f&ff Gw %mwg,, c

.s u

y%M qwA e

A.fef h& f hff jfGlxf Q

40(,3g;y W: %[ d i g@M.% gg g;,5 Q.* Q MfyR h ^pQDQ*py; _ i OGC p l

W G #,i N y yT Q Q;;;,%,

' J.W.(- G.a==411

,J... y.

g$kbM %g.j @ h k

.a.e Q c

.fpg ye! g@yg;W f

gn m &'g J.M.ic.C

-4 M

a H YL*NW,**"Y N,;[f hh 4m}ha~gnw wh k f

. w'.$',b 4 +p W. r e

4 nu a

d Ye yg.4' f %O W Y }g s4.pgp y g7 g}, g g.m'*pg g'..,rg g @(.

W H.' ;I}en ton, g;p ggy. [,

^ ;

+

- P. p,.+,,., M; A

F Ko

. a i n: + 2 p&

,(

' y" '

n

- :,'a

. L

- _ y.,,N. -,

n

'x

_/ f a

m.h'

\\

x t g :<

\\..,

w-

~p e '.m n,z m~e o n - y. -...h. mh e, s%%~%pg:md;eJih N :1N%e% '. 'i %!

m L>' $. 3..

s..

u m. ?. e~.$b s a( a.&.:c.

.V h

c g-

.; % m. w e c%a -

a

-m 5

3 t

r

- ngu

. w.c g u, wp -.a m,f M

-y. 9 3 y IN 4 14 4: er 9 V w: A R W S* #

• I'#

m- - x ;-@ 3 p 7

F* '**

,--.L;:p 'fFWRgJ-*----*-

'~ya %A y Q$%

x y

'. g BD/T- ----

+

-, " -..... }c.c.g.,..

's.,

m AB

J4 -

'DLy

..OrrlCE,3 -

+.a-

- -. - - - -,-. y--

a p:

. n.y.A, y

,y%c<,

,y3

.,9

,,7.,

t 7

7

• gp q&(g:g s_

w

.; g h 2 sth

. bAia N -..

E10 -- Nb dN-ut h.MNh!M kNNMINNMMS khNbN N b

s--

blW5ANNMUf8$MOMW5WNbMN

n

-n

,. w v e m' ?. aw m^ m vn w e.m yr~rm w wwrw w sy p

f,

,/ h

.A r%

• Td -

.;w

?'.,.->c 8

W 7A.q h y W x % d-s

,..y

.~ m s

. a:isWs: bW

.m LWf

~ M'N.cp;fp %y'% ;;'u' Md(W :

2p% ; ?

y.+

g ktd:4N F; ' 61

' b h

's n.M p c p y%w%.eg...d.GM'.

4m. a sy. Q..ggg~.

sp w

?u&.: m :yOg%q:M_ Q+.f Q.am%w q hm:

gg ~. -

e

%m:a ;a% 11 %

h3

,... e - w.s

1 r

v wn<;

.w., w m.~.,' ~ :w,. x....,. ;m e,y$1, m a s.o, +.

0,y.

.. a :

e

.~

ta.:g.1 y,

., w y

.w m,

a 2

.gg

RCDeYoung PWR Branch Chiefs RWKlecker APR 3 1973 BCBuckley M* E N EIGoulbourne (2)

PWR-4 Rdg JHendrie g.

AKenneke

./ ! ?

..1 b jt,f m f <*s c.?^

(RWKlecker

' ' " 7 f... '

<g, o

w 4

g,

p s

s.,,,..

YM Y

A g man;p 1:a. n.es w4Nlygg;.% E?!.y^^g%p@u a,g g

m-n m,

ww.pp p

s h.; %

us.e s amme, h * %: n% e% M'(( w,K % a n

pw

. ' ~

fi

. F. O. Best, sp.mMQ,hY#;

y@ e y% a*.

.n.,, sema mm m.- me gw6 o, T *d.,w"".y.'.3,.cY

4..

.A.

1 p u+r ; l.,4. + 3.(,4,.2 e. L,i A, m

,. ~p.

4.f,a. 44,.a....,',,f. e. s.,..,.c c.

- %'.MM lrN lgl.ypg ;i1Y',j, h 3.xgggY?,,W?:Yl*W. 3g g g

+e..

?

Y.'

4.

2-

=

f

-e

• y-1.,'

a M4

- W t

'J R

gggggg$&'g(y,,,.g' YU., NW,-

(

• = <

a.

1 s

W iM

?

. y =2 1

on m. u.s. er..

nsumorv a air m.'nmal anseer menares W:n mapart sur the cuestal aher-mes:3 assamer esmeretius next, we 2 7'

tima mat we used aesttia

.1 e=sh-.k+- to aumenete sur e=1-H he spesins Smosumassam regstssif $s listed.2s esfemalasere. OM'$; 9_Q l

g

. ' w 4 Qi)w%. % Q. a:!f W N Qld.r dig, W Q & M

+g f' ;inna.

e

%[.-;

[^

g3MTM hidisun;u gem M

J

$,.)y. ~;%g y

y k:g,,

jp l'QlM,W,7 lk eg&

-w,U h W

.M?

W

.u

.c

),

7i n

a-

M.[

Dal W rempannena Q g M ' jpg ja~

i33

,yyy '

jegdest ac1 base;te bei me-amaraense sen.setsummut air;me se OUWWell s h M M UJ 11 W FWW$stll k th$3' IQ y q,y y.g,

ren svlar

Yuary 3

,g y,,, y

- w@@M%

e z.nismens mi me."

S isiet

^ ' 333mmey he ~

TW

.m' iestaut'.ar:duesy;jua)eur suspumes. '

a

<..:,.> nyM gappa.pc w:my a +;

i v,w+ 3:wy;.1 m;y-m..m e n e

.. u m n

r u

.c 4

!khA h*e

. h,

';hi b b '

• .D hM h

.m y y p,; g. n' y.-

' w'.m ':

3&

. m %w&y m.Q p % m% m q, q g y &

p.

' % $f f20 W

Q Q sig WWM V6 ymm

. gW ; 3$gw;mgwy@qMgJ;G,QcJMh g3 w3:gw M sem.

am.wg" maarM #i23W#@ZMNM y

t MGPM+44.w:g%ess ac me. ism p p.g4 ppg.gp W R M. M @4 :

WifiMM N

~

w-awejp, n.g q Q,gy. g.

.an

,3,jM4 ng v

g g vy.e

- g n p qu.

lan.1

W jM

+-

- ;e A

c+

m ag=

4 Meques't ftR* Add 1Hamm1 M g A A' p.

m Ne

/g,.

I'

'Resida Pouer Serveruktam

• _g af p

F ab a.

. -M!f M

<Q**

omdr >

' '_.pr..(.Q.l.ggf.%samaammm..Q. 'wM,y. S

' ffWR.i4 ~ ~

MM12-4 ~

D T

M i ~ PMl)_.

yj -naa ngq % <

y9:p 6 ~

3

'q.q;.g%.nW6;;;d MWW$iWWii-

, mm, :..... cm y.n.,;m wn,a..

senucpay.ygg 2# 4 2HM & &9M.BMW 'hs M S M WMX Wi$$ $ w ww?$g$ m$lkd551sdW1&a w$Ai$$pn y yp ha,.ncr.+ nomam

.r.

ww:

y EAh n

62h0Whn W

80 03160 / V

'S

~,

REQUEST FOR ADDITIONAL INFORMATION FLORIDA POWER CORPORATION CRYSTAL RIVER LNIT 3 DOCKET NO. 50-302

1.0 INTRODUCTION

AND

SUMMARY

1.6 Describe the quality assurance programs and quality control checks that are designed to assure the mechanical integrity 'f your fuel over its anticipated lifetime including any design review effort, review and audit of quality assurance measures and your planned inspections of the fuel upon delivery. Indicate how your fuel design and manufacture will minimize possible failures from clad hydriding and UO - clad interaction.

2 O

t

-s

. 2.0 SITE AND ENVIRONMENT 2.20 Expand the population data presented in Figure 2-6 of the FSAR to include the projected population for each of the next four decades.

Describe the basis used to estimate the projected population.

2.21 Estimate the peak transient population in the area around the site.

Give particular attention to population changes and distribution within the low population zane.

2.22 Modify Figure 2-23 of the FSAR to show each gaseous radioactive ef fluent discharge point and state the distance between each discharge point and the nearest point on the site exclusion area boundary.

In addition, specify the location of the fuel storage tanks and provide an evaluation of the effects of a fire and/or explosion of these tanks on safe operation of the nuclear facility.

2.23 Identify the western property boundary along Crystal Bay and the Gulf of Mexico and describe your legal rights to any appurtenant structures which extend into the Bay or the Gulf. Discuss the manner in which public access to the exclusion area via the water front boundary will be controlled.

2.24 Identify all public facilities such as schools, hospitals, prisons, and parks within 10 miles of the site. Give their location, distance from the site, and their temporary and permanent populations.

I 2.25 Describe the agricultural uses of the land surrounding the site with respect to type, acreage, and annual harvest of principal crops and food products.

2.26 Identify all dairying operations within 10 miles of the plant site, state their size, and show their location with respect to the site on a ma, of suitable scale. Indicate the nearest area which might, in the future, be used for dairying.

2.27 Describe the extent of sport and commercial fishing in the vicinity i

of the site and provide data on species and annual catches of fish and other edible marine biota taken by sport and commercial fishermen.

j l

. 2.28 Identify the public and private recreational facilities in the vicinity of the plant site, give their location, distance from the site, and estimate their usage (man-day / year) for various recreational 1ctivities.

2.29 Describe major transportation routes, highways, railroads, and waterways, which pass within five miles of the site and give their distance and. location from the plant site.

2.30 Identify any manufacturing plants, chemical plants and storage facilities, and mineral mining operations within five miles of the plant site. State their distance and location from the site and products which are produced. Describe any hazardous materials used or produced by these firms and evaluate the effect on the plant of an accident involving these materials. If explosives are used at the nearby dolomite mining operation, evaluate the effect of an explosion involving the maximum quantity of stored explosives.

2.31 Please provide statistics for the area on the intensity and frequency of fog, hail, and thunderstorms.

2.32 Safety Guide 23 and portions of 10 CFR referenced therein point out the need for onsite meteorological measurements in the assessment of the consequences of accidental or routine emissions to the atmosphere during the operating phase of a power reactor. Discuss your plans for a continuing onsite meteorological program.

2.33 Provide evidence as to whether the period of meteorological record (6/19/70 - 6/19/71) was a reasonably representative year in the area.

S

. 4.0 REACTOR COOLANT SYSTEM 4.19 Supplement your response to Request 4.3 by describing the acceptance program that will be implemented to determine the acceptable anplitudes of the vibration for confirming the structural integrity of the piping and pipe restraints.

4.20, Clarify your responses to Requests 4.20 and 4.21 by providing a more 4.21 detailed description of the dynamic analyses performed for Class 1 piping and associated supports which determine the resulting loadings as a result of a postulated pipe break, as follows :

(1) justify any departure from the attached criteria (Attachment I) which determine the locations of design basis breaks on which the dynamic analyses are based; state the postulated rupture orien-cation, such as a circumferential and/or longitudinal break (s),

for each postulated design basis break location.

(2) describe the forcing functione to be used for the pipe whip dynamic analyses, including direction, rise time, magnitude, duration and initial conditions that adequately represent the jet stream dynamics and the system pressure differences.

(3) provide typical diagrams of the mathematical models used for the dynamic analysis.

(4) provide a surmary of the analyses performed to demonstrate that motion of ruptured lines where unrestrained will not sever adjacent inpacted piping or pierce impacted areas of containment steel liner. Justify any departure in (2), (3) and (4) above from the criteria specified in Attachment II.

~

i 1

l

\\

A, Attachment I (Requests 4.20, 4.21)

Acceptable Criteria of Pipe Whip Protections I.

Protection against pipe whip is required for high energy fluid systems (or portions of systems) except where A.

either of the following piping systes conditions are met (a) the service temperature is less than 200*F, a

1 (b) the design pressure is 275 psig or less, or B.

the piping is physically separated (or isolated) from other piping or components by protective barriers, or restrained from whipping by plant design features, such as concrete encasement, or C.

following a single break, the unrestrained pipe movement of either and of the ruptured pipe in any possible direction about a plastic hinge formed at the nearest pipe whip restraint cannot impact any structure, system or component important to safety, or D.

the internal energy level associated with the whipping pipe

?

The internal fluid energy level associated with the pipe break reaction may take into account any line restrictions (e.g., flow limiter) between the pressure source and break location, and the effects of either single-ended or double-ended flow conditions, as applicable.

i The energy level in a whipping pipe may be considered as insufficient to rupture an impacted pipe of equal or greater nominal pipe size and equal or heavier wall thickness.

i l

~.

can be demonstrated to be insufficient to i= pair the safety function of any structure, system, or component to an un-acceptable level.

I:

Design basis piping break locations in the piping systems specified under I should conform with the following acceptable criteria:

A.

ASME Section III Code Class 1 piping breaks should be postulated to occur at the following locations in each piping run or branch run:

(a) the terminal ends, and (b) any in'termediate locations between terminal ends where the primary plus secondary stress intensities S, (circum-ferential or longitudica1) derived on an elastically cal <.ulated basis under the loadings associated with one half safe shutdown earthquake and operacional plant conditions' exceeds 2.0 S for ferritic steel, and 2Piping is a pressure retaining component consisting of straight or curved pipe and pipe fittings (e.g., elbows, tees, and reducers).

3A piping run interconnects components such as pressure vessels, pumps and rigidly fixed valves that may act to restrain pipe movement beyond that required for design thermal displacement. A branch run differs from a piping run only in that it originates et a piping intersection, as a branch of the main pipe run.

Operational plant conditions include normal reactor operation, upset conditions (e.g., anticipated operational occurrences) and testing conditions.

S is the design. stress intensity as specified in Section III of the ASME B$ilerandPressureVesselCoda,"NuclearPowerPlantComponents."

% 2.4 S for austenitic steel, and m

(c) any intermediate locations between terminal ends where the cumulative usage factor (U)0 derived from the piping fatigue analysis and based on all normal, upset and testing plant conditions exceeds 0.1, and (d) at intermediate locations in addition to those determined by (b) and (c) above, selected on a reasonable basis as necessary to provide protection. As a minimum, there should be two intermediate locations for each piping run or branch run.

B.

ASME Section III Code Class 2 and 3 piping breaks should be postulated to occur at the following locations in each piping run or branch run:

(a) the terminal ends, and (b) any intermediate locations between terminal ends where either the circumferential or longitudinal stresses derived on an elastically calculated basis under the loadings associated with seismic events and operational plant con-ditions exceed 0.8 (S +S) o.

the expansion stresses U is the cumulative usage factor as specified in Section III of the ASME Boiler and Pressure Vessel Code, " Nuclear Power Plant Components."

7 I

~

~

Sh*8A Class 2 and 3 components, respectively, of the ASME Code Section III Winter 1972 Addenda.

f

G exceed 0.8 S, and (c) intermediate locations in addition to these determined by (b) above, selected on reasonable basis as necessary to provide protection. As a minimum, there should be two intarmediate locations for each piping run or branch run.

III. Pipe break orientation at the break locations as specified under II should conform with the following acceptable crit.eria:

1ongitudinal' breaks in piping runs and branch runs, 4 inches A.

nominal pipe size and larger, and/or 0

B.

circumferential breaks in piping runs and branch runs exceeding 1 inch nominal pipe size.

8 is the allowable stress, range for expansion stress calculated by the S

rules of NC-3600 of the ASME Coda,Section III, or the USA Standard Code for Pressure Piping, ANSI 331.1.0-1967.

Longitudinal breaks are parallel to the pipe axis and oriented at any point around the pipe circumference. The break area is equal to the effective cross-sectional flow area upstream of the break location.

Dynamic forces resulting from such breaks are assumed to cause lateral,

~

pipe movements in the direction normal to the pipe axis.

10Circumferential breaks are perpendicular to the pipe axis, and the break area is equivalent to the internal cross-sectional area of the ruptured Dynamic forces resulting from such breaks are assumed to separate pipe.

the piping axially, and cause whipping in any direction normal to the pipe axis.

D

. *g

-9_

Attachment II (Requests 4.20, 4.21)

PIPE WHIP DYNAMIC ANALYSIS l

Analytical basis should be provided to assure that pipe motion caused by the dynamic effects of the design basis breaks, either within or I

outside of the containment, will not impact or overload any structures, systems or components to the extent that the safety function is impaired or precluded. For those systems in which pipe whip effects are required

~

to be taken into account the dynamic analysis methods used should be compatible with the relative importance of a system to plant safety.

The following methods of dynamic analysis are acceptable provided the associated design criteria are met:

I.

Pipe Whip Dynamic Analysis Criteria An analysis of each pipe run or branch should be performed for a.

the longitudinal and circumferential postulated rupture at each design basis break location.

b.

The loading condition of a pipe run or branch prior to postulated rupture in terms of internal pressure temperature and stress state should be those conditions associated with normal operating

• condition of the system.

The dynamic system analysis method used should be compatible with c.

those permitted in Appendix F of ASME Section III - Nuclear Power Plant Components Code. The allowable design strain limit should not exceed 0.5 ultimate uniform strain of the materials of piping 4

9

^'

10 and restraints. Thus, the method of dynamic analysis used should be capable of determining the inelastic behavior of piping-restraint system response, d.

A 10% increareef S may be used in the analysis to account for y

strain rate effect.

e.

For a circumferential rupture, pipe whip dynamic analysis need i

caly be performed for that end (or ends) of the pipe or branch which is connected to energy source of sufficient capacity to develop jet stream.

j 1

f., Dynamic Analysis Methods used for calculating system response to the fluid inertia and wave thrust developed in a pipe following postulated rupture should adequately account for the effects of:

(1) mass inertia and stiffness properties of the piping system, (2) gaps provided between piping and restraint, (3) elastic and inelastic deformation of piping and restraint and (4) governing boundary conditions.

g.

Dynamic analysis methods and procedures should consist of:

(1) a representative mathematical model of the piping system and restraints, (2) the method of solution selected (3) solutions for the most severe response among the design basis breaks analyzed,

~

e 9

'N.

(4) solutions with demonstrable accuracy or justifiable conservatism.

h.

The extent of mathematical modeling and analysis needed should be governed by the method of analysis required by these criteria.

4 0

t e

6 g

o e

S e

O h

a

II. Acceptable Dynamic Analysis A.

Acceptable Models of Analysis:

(1) Lumped-Parameter Model Lumped mass points interconnected by springs take into account inertia and stiffness effects of the system and time histories of response are computed by numerical integration to account for gaps and inelastic effects.

Kinetic energy generated during the (2) Energy-balance model first movement of the pipe which impacts the restraint is converted into equivalent strain-energy. Maximum deformations are determined for various elastic-plastic combinations of the pipe and the restraint compatible with the level of Kinetic energy.

B.

Application of Models:

(1) Piping Systems - ASME Section III Code Class 1 Components For pipe breaks connected to a energy reservoir che %

a.

~

the forcing function remains at 70% or higher of the inicial pulse magnitude after 0.2 second of break occurrence either:

1.

Lumped-parameter model analysis is acceptable, or 2.

Energy-balance model analysis is acceptable provided an amplification factor of 1.5 is applied to the forcing function in order to account for the rebound 4

O 4

effects of the piping within the restraint.

b.

For pipe breaks associated with a energy reservoir where forcing function reduces to 70% or less of the initial pulse magnitude af ter 0.2 second of break occurrence, either:

1.

Lumped-parameter model is acceptable, or 2.

Energy-balance model is acceptable provided a factor of 1.2 is applied to the forcing function.

(2) Piping Systems - ASME Section III Code Class 2 & 3 Components Both models are acceptable and the same factors for forcing functions as stated in 1 for energy-balance models should be used.

III. Flow Thrust Force:

A.

The thrust force induced by jet flow at the design basis pipe break location should consider:

(1) the inizial pulse, (2) che thrust dip, and (3) the transient function defined by a time history.

B.g'The initial pulce should have a magnitude not less than T = KpA where p = system pressure prior to pipe break A = pipe break area, and K = thrust coefficient.

~

r

,, ~,--.

.o

N.

Acceptable K values should not be less than the following (a) 1.26 for saturated steam, water and staa n/ water mixture.

(b) 2.00 for subcooled water-nonflashing.

C.

A pulse rise time not exceeding one milisecond should be used for the initial pulse, unless longer crack propagation times or rupture opening times, can be substantiated by experimental data or analytical theory.

D.

The transient function, A (3) above, should be provided and justified. The shape of this transient function should be related to the capacity of the upstream energy reservoir, including source pressure, fluid enthalpy, and the capability of the reservoir to supply high energy flow stream to the break area for a prolonged interval. The shape of the transient function may be modified by considering the break area and the

.ystem flow conditions, the piping friction losses, the flow directional changes, and the application of flow limiting devices.

E.

If the transient function, A (3) above, is monotonically diminishing rapidly and its highest peak does not exceed the magnitude of the initial pulse, the pulse alone is sufficient to represent the flow thrust effect on the system. In such a case, the thrust should be considered to remain at a constant value as specified in B until the pipe completes its first impact wit' the restraint. This may be considered the forcing function uted in energy-balance model.

l

~

l

. 4.22 Response 4.22 is incomplete. Provide the stress or deformation limits and the design atandards or codes applicable to the pressurizer support.

4.27 The response to Request 4.27 states that a plastic analysis was used to design and analyze the pipe whip restraints installed to prevent a ruptured main steam er feedwater pipe from damaging a component of the reactor coolant pressure boundary. Provide a summary of the methods of the plastic analysis used including justification of the validity of such methods. In addition, provide the stress or strain limits applied to the results of the plastic analysis and justify the basis for their application.

4.28 Section 4.1.3.4 of the FSAR states that the pressurizer code safety valves comply with Article 9,Section III, of the ASME Boiler and Pressure Vessel Code. However, Artit.le 9 does not address itself to the structural integrity of valvr.s.

Provide the stress or deformation limits which will be ur,ed for safety valves and justify the basis for their application.

4.29 Section 4.2.2.8 of the FSAR stat es that Nds imposed on the safety and relief valves within the reactor cou ant pressure, boundary are limited to acceptable values. Define the term " acceptable" in terms of stress or deformation limits and justify the basis for their application.

4.30 Provide the dynamic testing procedures used in the design of Category 1 mechanical equipment (such as fans, pump drives, valve operators, heat exchanger tube bundles) to withstand seismic, accident and operational vibratory loading conditions,

including the methods and procedures employed which consider the frequency spectra and amplitudes calculated to exist at the

~'

equipment supports. Where tests or analyses do not include evaluation of the equipment in the operating mode, describe the bases for assuring that this equipment will function when.

subjected to seismic accident loadings and vibratory loadings.

4.31 Regarding the fracture toughness data obtained for all pressure-retaining ferrite materials of the reactor coolant pressure boundary, state the degree of compliance with the test methods and acceptance criteria of the recently revis~ed ASME Code Section III fracture toughness rules (Summer 1972 Addenda).

.m

_ 4.32 For the materials of the reactor vessel beltline (including welds),

provide tne initial upper snelf fracture energy levels, as determined by Charpy V-notch tests. Provide these levels in both directions, ir available in both directions.

4.34 Provide proposed operating pressure-temperature limitations during startup and shutdown and during hydrostatic testing of the reactor coolant system, using as a guide. Appendix G, " Protection Against Ncn-Ductile Failure," of the recently revised ASME ' Code Section III fracture toughness rules (Summer 1972 Addenda).

4.35 Describe the extent to which you have reviewed the design of affected systems and components to determine that annealing of the reactor vessel will be feasible should it be necessary because of radiation phtittlement after several years of operation.

State the maximum reactor vessel temperature that can be obtained using an in-place annealing procedure.

/~5

. 5.0 CONTAINMENT SYSTDiS AND SPECIAL STRUCIURES 5.14 Amend your response to Request 5.14 to provide sketches of the mathematical models employed including models of soil-structure interaction effects.

5.16 Amand your response to Request 5.16 to confirm the validity of the fixed base assumption by providing summary analytical results that indicate that the rocking and translational response is insignifi-cant.

Include a brief description of the mathematical model and damping values (rocking, vertical, translation, and torsion) that have been used to consider the soil structure interaction.

5.18 Amend your response to Request 5.18 to describe the measures taken to i

consider the effects on floor response spe:tra (e.g. peak width and period coordinates) of expected variations of structural properties, damping, soil properties and soil structure interaction.

5.26 The responses to Requests 5.26.1 and 5.26.6 are not acceptable.

Your response to Request 5.26.1 has not demonstrated theoretically that the response spectrum method gives conservative results compared to the time history method. Demonstrate, for all floor response spectra developed, that the response spectrum method used gives conservative results compared to the time history method unless the use of the time-history (or equivalent) method is planned to develop the floor response spectra. Clarify your response to Request 5.26.6 by stating whether the internal shears, moments and internal reactions were determined for each mode separately and then combined using the square root of the sum of the squares or by describing any other method that was used to obtain the combined total response.

5.33 Appendix A to 10 CFR Part 100 requires that suitable instrumentation '

be provided to determine, promptly, the seismic response of nuclear power plant features important to safety and to permit comparison of such responses with that used as the design basis. A single strong motion accelerograph installed on the top of the containment structure is not acceptable to meet the intent of Safety Guide 12, " instrumentation for Earthquakes." In addition, seismic instrumentation is also needed for other Category I structures and components.

5.33.1 Provide an appropriately modified seismic instrumentation program which includes the selection of the type, number, location and

<s i

utilization of strong motion accelerographs to record seismic events and data on the frequency, amplitude and phase relation-ship of the seismic response of the containment structure.

5.33.2 In addition, describe the seismic instrumentation system (such as peak recording accelerographs and peak deflection recorders) that you propose to install in selected Category I structures (other than containment structure) and on selected Category I components.

Specify the basis for selection of these structures and components, and for the location of instrumentation, as well as the extent to which this instrumentation will' serve, following a seismic event to verify the predicted response for these structures and components as derived from seismic analyses. Include the criteria and procedures that will be used to compare measured responses of structures and components with the resules of dynamic system analyses.

5.33.3 Describe the provisions that will be utilized to signal to the control room operator the value of the peak acceleration level experienced at the facility foundations within a few minutes after the earthquake. Include the basis for establishing pre-determined values for correlating the readout of the seismic instruments with the ground motion specified for this site, i

=

rs 6.0 ENGINEERED SAFEGUARDS 6.11 Provide an analysis of a double-ended break in the line that connects a core flooding tank (CFI) to the reactor vessel. The location of this break should be between the reactor vessel nozzle and the first check valve leading to the CFT. This analysis should be conducted using the evaluation model approved under the Interim Polic/ Statement, Ap endix A, Part 4.

6

, 's s

11.0 RADIOACTIVE WASTE AND RADIATION PROTECTION 11.12 Provide sufficient information to clarify the following discrepancies:

11.12.1 The annual quantity of radioisotopes released from the radioactive liquid waste system as estimated on page 11-Sa of the FSAR does not agree with the estimate given on page 11-9c and page 11-37c of the FSAR.

11.12.2 The writeup on page ll-4b of the FSAR states that laundry wastes will besenttothemgscellaneouswastestoragetank,iftP-activityis greater than 10 pC1/ml after dilution. Figure ll-1B of the FSAR shows this stream being sent to the neutralizer tank.

11.12.3 Figure V-12 in the Environmental Report shows the deborating demin-eralizer regenerants being sent to the neutralizer tank.

Figure ll-lb of the FSAR shows these wastes being sent to the miscellaneous waste storage tank.

11.13 On the basis of the design and operation of the proposed radwaste treat-ment system, provide estLmates of the maximum whole body dose to an individual and the maximuu organ dose to an individual that would be received by the general public at the site boundary as a result of the release of liquid and gaseous affluents.

11.14 In our opinion, releasg of untreated radioactive liquid waste with a concentration of 10 pC1/ml is not as low as practicable. Justify the release of untreated laundry wastes with concentrations of 9

10 pCi/ml as being as low as practicable.

11.15 Provide additional details on the reactor building purge exhaust duct (Rm-A1) and the auxiliary building and fuel handling area exhaust duct (Rm-A2) monitors, including, the distance of monitoring filter elements from the exhaust stream and the bases for this distance, the location of the sample intake in the exhaust stream and basis for this location.

11.16 Describe the capability in the plant of making periodic isotopic analyses comparable to those outlined in Safety Guide 21 of noble gases, iodines, particulates, and liquids released to environment.

11.17 For each logical grouping of instruments, describe the calibration procedures to be used.

. 11.18 Describe the operational characteristics of radiation monitors, including type of detector, sensitivity, range, met. nod of calibration, set-points (and their bases), the criteria used to determine necessity for and location of the monitors.

11.19 Provide an estimate of the yearly onsite man-rem exposures from the plant.

Compare the estimated doses with experience from relevant operating plants.

l 1

~

-m s

. 14.0 SAFETY ANALYSIS 14. 19 Provide assurance that a single failure of any feedwater isolation valve will not result in continued feedwater addition to the affected steam generator (s) during a steam line break accident or justify that auch a failure would not conpromise safe shutdown of the plant to an unacceptable degree.

-o b

y

,m-,y

-r,

- - -, -,,