ML19323C777

Amend 15 to License DPR-6,revising Tech Specs Re Max Average Planar Linear Heat Generation Rate

ML19323C777
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	10/17/1977
From:	Stello V Office of Nuclear Reactor Regulation
To:
Shared Package
ML19323C736	List: ML19323C735 ML19323C744 ML19323C746 ML19323C747 ML19323C748 ML19323C749 ML19323C750 ML19323C758 ML19323C763 ML19323C764 ML19323C766 ML19323C767 ML19323C770 ML19323C772 ML19323C777 ML19323C780 ML19323C781 ML19323C784 ML19323C786
References
FOIA-80-162 NUDOCS 8005190173
Download: ML19323C777 (16)

Text

--

.n

.....,4 O

o 8oos190/n UNITED STATES Ve p

Vjyv

-4 NUCLEAR REGULATORY COMMISSION

,7, kFM,j W/,SHINGTON, D. C. 20555

'5 a

,g v,/

....+

C0HSUMERS POWER COMPANY DOCKET N0. 50-155 BIG ROCK POINT PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 15 License No. DPR-6 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The applications for amendment by the Consumers Power Cnmpany (the licensee) dated December 17,1976 (as supplemented by letters dated February 9 and August 17, 1977) and April 15, 1977 (as e ipplemented by letters dated April 21, August 12 and 24, and September 26, 1977) comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The request for exemption from ECCS failure criterion of 10 CFR 50.46, Appendix K, Paragraph I.D.1 dated September 15, 1977 (as supplemented by letter dated October 12,1977) is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest.

C.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; D.

There is reasonable assurance (i) that the activities authorized by this amendment can. be conducted without endangering the health and safety of the public, and (ii) that such activities will be concucted in compliance with the Commission's regulations; E.

The issuance of this amendment will not be inimical to the ccmmon defense and security or to the health and safety of the public; and

nw y

..,. ~.. -

O O

• e

,r

. F.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraphs 2.C(2) and (3) of Facility License No.

DPR-6 are hereby amended to read as follows:

(2 ) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.15, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Exemption from 10 CFR %50.46, Appendix K, Paragraph I.D.1 Pursuant to 10 CFR 150.12 the licensee is granted an exemption from the ECCS failure criterion of 10 CFR 550.46, Appendix K, Paragraph I.D.1 as applied to a Loss-of-Coolant Accident followed by concurrent single failure in the redundant core spray system for the 1978 operating fuel cycle.

j j

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUC R

ATOR'f COMMISSION hctor e>10, irector Division of 0 rating Reactors Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 17, 1977 L

O O

ATTACHMENT TO LICENSE AMENDMENT NO. 15 FACILITY OPERATING LICENSE NO. DPR-6 i

DOCKET NO. 50-155 The Technical Specifications attached to Facility Operating License No.

DPR-6 are changed as follows:

1.

Add the following column to Table 5.1 for Reload G-3:

11 x 11 3

0.577 Zr-2 113 0.034 0

0.449 4

91.5 1

70 3

Helium >95%

2.

Add new Figure 5-8 attached.

3.

Replace the following revised tables in Section 5.2.1(b):

Tab'e 1 and Table 2 4.

Add new Figure 1 and Figure 2 following Table 2.

5.

Replace Table 8.2 with the revised Table 8.2.

6.

Replace Section 11.3.1.4 with the following revised pages:

11-1 through 11-5 7.

Delete Page 11-6 of Section 11.3.1.4.

8.

In Paragraph 4.1.2(b) revise the first sentence to read:

"A minimum of one reactor recirculating loop or its equivalent shall be used during all reactor power operations."

9.

Replace pages 11-14 and 11-16 of Section 11.3.3.4 with attached revised pages.

I

s., -

k 0

10lIiiiiillI1lili I

,I g8ap j.

g,. -

)

• I d

I

g n ili i'!l t,

i i i

- i i

< i :.

l l

l i.

I>

l!

!1 ir i

e M>

S t

i wr:. hti I

h iii.h

/[: p!l g!

- l t; W} l:ii,s >i;ll.:

i i b.

si m

y-i!!

l t!b :s s Ig,=='

tl n i

./

i m

d[',1 yt l

'liihi

,[mgti a

i

.s

!pi:M..e!. m

=

Ji.O.,, t i

u.

i ':vo 2 :

t l r

t4ai!

a I tth sus re.< : j;}M,i.,'; !,I4 I

n c

E i

!!;!lilli'itpiIT._l,1 o

=

-a A

c

.,,.;..-..;,a-a,

[jF0ll!11il8MII t )

g

,[

p, e.-

.......i,,-..iai..i..,nni.

j-l.-

i i.i, m,

l

l.l p,.! !,!., Il i l! I !

,: lc i

i l i- ;_

i i

i I

-.r 1

t

pm=t=i@

l s

t j

i I

3 I

l I;l t

u Idd S P't 3 - E W 3'8l m W '"

i 1

y i

?!

s, G

l l l l1 s

l c

(y.s.. =m=..

.:= c. :. ~ -h~.m j.

(

l e

y y

M 1-v,,t, _, _, _, _,... _,a.

~ n..cm

._~~b-t, lg co j3 :: Ex. 4't X MS==m W->; 2c%9Wn.g

, d,b l

l i<

a 1

1ls g

- -- a

_m as

-. O' x _' '

-d -ec it w,

3 %e i

',I l

5.

g

2..x.._.y ' '. _:-:-m..a:^=' 'C.-'@D& =&

} ii!

9 I

~ IT' T

~

~

d t

=

4 c':.'.x g.

4 s

- ~ -

l

$. ':N l-i s :Ip!

i W W W WiMistm:

l AF>:'G ES M.^siic7: ' E:

,i I:h g,..m m P-y f, U, fr -

I x

--yv.,--

3.u;

~ n y:-

,e c

! if,,

6.Me-9,9e,s_-e-eee)

,g Il l l I

i3 O, e

--- m 2-e a

z

, N.u_. r:10,.....,

I e

L

_J l

l llll' l

I l i

i s

i n

e!

'l i

I I i

y r

'l l

i

'J e

-ll lL

=

5 l

d i

b A ml-I(*

i

.1 i,

.,.'. m..-mm...

N-n m _ -'

,4.@ l '

U, N;

r_ -.-:.

l u p:M4hE &'d4Nk bm.-e j$-Mg J i

lij j

[.

i I

i M 'd ;pI E Q QT- "' p :p=<

4 WI:@='h!ME i

6

$s w+ % m m.-n--_wvo 4->xc-O

)

.i g,l p,,

x gy),'O,;g3;.k,,';j',,,,T&

ni

- :+ --0.. -- t C - 0 0 e -qaH q

,p.

p

.l

,ijk.4,,id'i:.l& f EMSAh_B{ll

F:'CNM4M

)

E 1.@M<j+f@$;4k69 p ib 1 'a:

.i y eL

[ W.-

w,.*v&MO, n u.: s.

..,,y, n.,_, N,

1 1_

- - m.,,

y,,

i.

u.

Y.Q.. N.'.'ifg.i<17 L

I

{tk

{

s e/:. N.ym ru

.L s

.?.

I g

J e

t

~

j n.,-..,

t Gn

-S-xL

-)

M,

x

)

).

{',

? %..'

'~

n.

_m c _

__ _s1

/

' :b h -a O

O N>s p(.t.,-

p t

(U4 FG '.

TABt.E 1 y!j ' Q<

l'

'/j.[-l 1

Reloads:

Reload Reload

^ 'f Nodified T & J-2' Reload C G-10 C-3

!:.iu

~

• I*N Hinic:u:n Critics 1 Heat Flux Ratio hi (

at Normal Operating Conditions

• 3.00 3.00 3.00 3.00 t;J

-.s

?

Minimum Bundle Dry Out Time **

Tigure 1.

Figure 2 Figure 2 Figure 2 1:> ;

D$;. l,:A, ;

Maximu:n yest Flux at overpower, Beu/h-ft-500,000 395,000 f.07,000 392,900

+8' a:Q rf -

i

/t J

'.,. Yo sdj MaximumgteadyStateHeatT1ux, k

Btu /h-ft 410,000 324,000 333,600 322.100 3

u.,1

,y g

Maximum Average Planar Linear Heat f

i Generation Race Steady State, w

kW/ft***

)

r Table 2 Table 2 Table 2 T,able 2

(

.*1 l

Stability Criterion: Maxicaza Measured zero-to-Peak Flux A=plitude.

..g Percent of Average Operating Flux 20 20 20 20

~

S

• /q d

Maximum Steady State Power Level, N

MWe 240 240 240 240 4

,. g Maxi =um *.*alue of Average Core Power

)

g Density $240 We, kW/L 46 46 46 46 74

(

3 l

Nominal Reactor Pressure During

.1 Steady State Power operacitm, psis 1335 1335 1335 1335

• g Mini =u:n Recirculation Flow Race. Lb/h 6 x 106 6 x 106 6x106 6x106 e.'

n Rate-of-Change-of-Reactor Power During

g. '

Power Operation:

t

, M i Control rod withdrawal during power operation shall be such that the average rate-of-d; change-of-reactor power is less than 50 MVg per minute when power is less than 120W,

glj less than TO We per =inute when power is between 120 ands 200 W, and 10 W g

g e per 3

i; minute when power is between 200 and 240 Wg.

h

}

p

• The bundle Minisurs Critical Heat Flux Ratio (MCHFR), based on the Exxon Nuclear ry.yj Corporation Senthesized Hench-Levy Correlation, n:ust be. abovy. this -ralue.

Il ~ Ni

?-

%~ ~k I

~

• • The actual dryout tice for each bundle (based on the General ZIectric Dry Out Q

k Correlation for Nonjet P:.:=p Boilina Vater Reactors, NE::E-20566) should be 4

g abuve the dryo'at eine shown in Figure 1 or 2, as approprince.

..ef ]j

• • • To be feturnined hy 11ne ar ext rapolat ion f rom Table 2.

l n -

Nhg l

UkD hend ent *:o. IS, 15 l

e4 4

I w

O O

v

,/

TABLE 2 MAPLHGR (kW/Ft) LIMITS Planar Average Exposure Reload Reload

(!Gd/STM)

Modified F

,F,J-2

. Reload G Reload G-1U Reload G-3 6.h53 6.h91 6.554 0

200 95 9.h alk 6 750 6.758 6.807 216 6.887 6.888 h3T 6.973 hh3 6.960 884 6.978 885 T.033 893 6 929 1,758 6.970 1,769 6.98h 1,T73 6.885 3,h9h 6.913 3,509 3,5h5 6 983 5,000 99 97 6.838 6,939 6.865 6,9To 6 9T8 T,085 10,000 99 97 6.8h7 10,h22 6.882 10,h82 T.019 10,690 6.867 13,938 6.goh 1h,019 7 069 1h,355 15,000 9.8 9.6 20,000 8.7 8.6 6.905 21,022 6.958 21,19h T.171 21,843 25,000 8.h 8.3 27,778 6.8h3 6.903 28,035 29,08h T.161 6.703 3h,013 6 923 35,157 6 958 35,322 Amendr.ent No. Id, 15

O O

2.5 2.0 mo Z

O M

m 1

S d

3 1.5 g

O u

O m

d Q ~= MAPLHGR LIMIT X NUMBER OF ACTIVE FUEL RODS 1.0

/

N 700 800 900 TOTAL BUNDLE PEAK LINEAR HEAT GENERATION RATE. Q(KW/ FT)

Figure 1 FUNCTION TCQ) FOR ENC 11X11 FUEL Amendment Nn.15

O O

1 2.5 8

~

2.0 8z l

O d

m l

6 s

l Y

i 1.5 g

O a

O i

0 Q = MAPLHGR LIMIT X M

g NUMBER OF ACTIVE FUEL RODS 1.0

/

N 600 700 800 TOTAL BUNDLE PEAK LINEAR HEAT GENERATION RATE, O ( KW/ FT)

Figure 2 FUNCTION TCQ) FOR GE 9 X 9 FUEL POWER VOID RELATION

A TABLE 8.2 Centercelt EEI UO -

U '#~

PuO2 mediate Advanced NFS-DA Minimus Core Burnout Ratio at Overpower 1.5*'

1.5" 1.5" 15 Transient Mininun Burnout Ratio in Event of Loss of Recirculation From Rated Power 1.5 15 l'. 5 15

',aximus Heat 5' lux et Overpower, h02,000 Stu/h-Ft2 500,000 Maximum Steady State Heat Flux, Btu /h-Ft2 410,000 500,000 500,000 329,000 Mani=um Avercge Planar Linear Heat Generation Rate, Steady State, kW/Ft Stability Criterion: Maximus Measured

ero-to-Peak Flux A=plitude, Percent 20 20 of Average Operating Flux Maxi =u= Steady State Power Level,IG 2h0 2h0 t

Meninal Reactor Pressure During

'1,335 Steady State ?cuer Operation, psig 1,335 Minimus Recirculation Flow Rate, Lb/h (Except During Pu=p Trip Tests or 6

Tatural Circultion Tests as Outlined in 6

6 x 10 6 x 10 Sec 8)

Nu ber of Bundles:

1 3

Pellet UO2 1

2 Pover UO2 Rate-of-Change-of-Reactor Power During

?crer Operation:

Control rod withdrawal during power operation shall be such that the average rate-of-chstge-of-reactor pcVer is less than 50 G per =inute when pcVer is g

less than 120 G:, less than 20 W: per =inute when power is between 120 and 200 :G, and 10 !G; per =inute when pcVer is between P.00 and 2h0 :G;.

• 3ared upon c *-* *' bat flux correlation, AFD 5286.

• • o icnger used in reactor.

Surveillance Requirement 1.imiting Conditions for Operation 11.4.1.4 EMERGENCY CORE COOLING SYSTEM I I. 't. l. 4 MERGEJy:Y_ CORE COOI_ING SYSTEM Applicability _:

dpplicability_:

Applies to the operating status of the emergency Applies to periodic testing requirements for the emergency core cooling systems.

core cooling system, Objective:

OJtj ective:

()

To assure the capability of the emergency To verify operability of the emergency core core cooling system to cool reactor fuel in cooling systems.

the event of a Loss'of Coolant Accident.

Specification:

S _e.ci_f (ca t ion :

P A.

Each month the following shall be performed:

A.

The two core spray systems (original and redundant) shall be operable whenever the Verify the operability of MO-7051, -7061, -7070,

-7071 and -7066 by remote manual actuation.

plant is in a power operation condition.

The original core spray system shall also Leak testing of the core spray heat exchanger.

be operable during refueling operations.

n.

ine core spray tecisculatluu systc Chall be Antnmatic netuation of both fire pumps.

operable whenever the plant is in a power Verify that valve MO-7069 is locked or sealed operation condition.

in open position.

C.

The core spray recirculation heat exchanger

(}{i shall not be taken out of service during Verify that the fire system transformer deluge valve is shut and its upstream isolation valve power operation for periods exceeding four (4) hours.

The heat exchanger shall be is locked or sealed in the shut position.

considered inoperable and out.of service if tube bundle leakage exceeds 0.2 gpm.

Verify that the hose required for backup cooling water to the core spray recirculation heat exchanger is installed on a designated rack in the screen house.

I' Verify operability of the condensate fill valve to the condenser hotwell.

I i

Amendment No.16,15 o c

a

!.imiting Conditions for Operation Surveillance Requirement l.4 EHERGENCY CORE COOLING SYSTEM (Contd) 11.4.1.4 EMERGENCY CORE COOLING SYSTEM (Contd)

B.

At each major refueling outage, the II.

Both fire pumps (electric and diesel) and following shall be performed:

the piping system to the core spray system tie-ins shall be operable whenever the plant is in a power operation condition and refueling.

Calibration of core spray system actuation and prcesure and flow instrumentatfoa.

E.

If Specifications A, B, C, and D are not met, a normal orderly shutdown shall be initiated Verify that the two core spray system con-tainment isolation check valves are not wit hin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the reactor shall be shut down as described in Section 1.2.5(a) stuck shut.

k) within twelve (12) hours and shut down as described in Section 1.2.5(a) and (b) within Calibration of fire cystem basket strainer the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. No work shall be differential pressure switches.

performed on the reactor or its connected systems when irradiated fuel is in the reactoi-Operability check of the core spray vessel which could result in lowering the recirculation system through the test reactor water level below elevaticn 610'5".

flow tank flow path, h F.

Until such time as the spray effectiveness of Verify manual and automatic actuation or the core sprav system valves MO-7051, prhaary core apray nozzles have been proven:

-7061, -7070 and -7071 with water flow n rmally blocked.

(1) two condensate pumps must be operating l

uuttur, powet operat*m. exccpt du-iag

"' ~'"""' ~~~~~" ~~ *** In,,

.....,c.._..._,

..._...-_,,,n startup and whem' power is <50%,

Verify that the hose used for backup

'(2) the condensate storage tank level shall cooling water to the core spray recircu-be 1 51 ouring power operaLivu, dud lation heat exchanner is operable and free

)l 6

of obvious defects.

(3) the condensate fill valve to the con-Perform a leak check and flow check of the denser hotwell is uperable.

backup cooling water hose when connected between the screen house fire water if (1), (2) and (3) cannot be maintained, a connection and the core spray recirculation normal orderly shutdown shall be initiated with-

]

heat exchanger.

in one (1) hour and the reactor shall be shut j

down as described in Section 1.2.5(a) within Instruments shall be checked, tested, and C.

twelve (12) hours and shut down as described in calibrated at least as frequently as

.Section 1.2.5(a) and (b) within the following listed in Table 11.4.1.4(a).

2a hours.

C.

l.atrument set points shall be as specified in 11-2 Amendment No. 16, 15

'+k'DDJ.b

TADT.FG 11.3.1.hn AND J1.h.3.hn i

Inut,rument.nt,f on That, ! nit,intes Ccre Spray J 1. ~i.1.bn I.in:l t ine, Con.Il t.lona for Opernt lon 11.h.1.hn Surveillance Requirement.a l

Trip System I.imi t.ing Condit, tons for Instrumant Ins t.rur.sen t.

Unrameter 1.o;;1e Set, Polnt.

_ Operability.

Tr. p Test.

Calibrat.lon Cr.en '.* ore Spray Vniven L 10'5" Elev

• .v :

1'i.

M.-

P...... 2 "

7

• I..n Quu te: /

~e Utster.

0.:e s. a 4 or

.c...a......,_

1 s,

... c.:

.o ans Lewc.

.2 4

W.lyca in Scricc Operation: (t) f Tria.ur.; freuuure One of Two for 1200 Tole, Power C.c.ut.lon Quar tel1.,-

Ikeh L.2,: r i

few (a)

P.ach of Tea and Ecfueling ncf.2eling i

Vulves in Geries Operationr (b) 2 e

Notes for Tablec 11.3. L.l a and 11.h.1.ha c sten,

( c.)

Initir'.lon of valyc cpcration requir.: both low reactcr v.tcr 1: vel coincident with low primarf-i piessure.

(b)

• h:. prio.ary core : pray cyctcc :h.Il t: cycilable for :: during refueling c;cration. Thc ' redundant ere-I

.; ruy uyst.em shall be inoperable during refueling operations with t,he valves blocked or otherwise defeat.ed h

(wh!]e t he pipind t;ection from the valves to the reactor head is dismantled).

11-3 Amendment No.16,15

!! nag:

Tlie core spray syst.em consists of two autom'at.ically actuated independent double capacity piping headers cap-chle of cooling reactor fuel for a range of Loss of Coolant Accidents.

Either syst.em by itself is capable of providing adequate evoling for postulated large breaks in all locations. When adequate depressurization rates are at hieved in the postulaLcd small-break situation, either core sprey system provides adequate cool-inn.

i'or the i n re,eni vuoniisie pipe brenk, a flev rate of apprvximately 1400 cpm in required nfter about 20

• eennan.

Euch cure upray ayulent hau 1(,0% coolia.6 capacit.y from each upray hender and each pump set.

Thus, specifying core cooling if the core spray syt hem is l

outh systems to be fully operat.ional vill assurt.yto a high degreaj h

requsred.

In addit.lun, the primary em spray ic required te 5 operable du-ia; mfueling operationa to provide fuel coolitig in the unlikely event, of an inadvertent draining of t.he reactor vessel.

'fhe core upray ayulems receive their s.ot.cr supply frcm the plant, fire system. The plant fire nystem supply from 1. uke Iiichirpui via L o redundani,1,000 gpm fire put:ps, one electric and ona diesel driven. These pumpn ju w

uturL nulumatically on decaying fire oyot.cm preusure.

If a passive failure of underground fire main piping chould occur during t,he long-t.erm cooling phase, the capability exists to bypass the affected portion of pip-Ing ut.tiizing a fire home i,u ensume t.lic s.unt,it..;ation of ler.g-tcr=.".,,CS cccling.

'rne core spray recirculation syst em is,provided to prevent excessive water buildup in the containment, sphere und i.o peu.ide fu. lu..g *.u r, ;, cst-acci *cr.t csclin;;. The cyc' v-consi=+ = ne tun pu=pn (lano gpm ench) and a neal, exchanges. Tue pusinpa t. uke a suctisn from the R.ccr levels of containmart end diacharge to the core spray ars ude r=. Tue ayulum is act.ust.ed ;; tar.ually t.^'cn the vnter level in the containment eisen to elevation 587.'eet.

'ilie '337-fcot elevation vill be achieved between 6 to 21s honrn operntion of one core spray and.one containment spray synt.em.

A Lest.1.ank and appropriai.e val ving la provided in the core sprtr/ rceirculation system so the pump suct. ion g

evudi LIuuu u.id *,;.e fici. charact :ristics cf thc.cyste enn M parindleeUy *eatad.

One core spray recliculat. ion' pump has adequate capacity to provide fuel cooling at anytime following a Loss of Coolant, Accident.

Cont.inuous conLaisunent spray operation is not required during the post-accident recircuin-1.lon phnse if only one recirculation pump is available.

4 a ' '

11-4

linses :

(Contd)

The i.perntle utatuu of the various systerns and components is to be demonstrated by periodic tests. Some of t.hece t.eut a will be performed while the reactor is operat,ind in the power range.

If a component is found to he inopernble, i t. Vill be persulble in most, enues 1.o effect repairs and restore the ayatem to full opernbility within u relut,1vely chort time.

For n ningic component to be inoperable does not negat.e the ability of the cyuti.m to perform it.= function, but il reducen the redundancy provided in t.he reactor design and thereby limitn the uhility to Lulerate additionnl equipment failuren.

If it develops that (n) the inoperable component is not, reps. ire.! withl;; the specified ull.wable time period; or (b) a second component in the same or related cyc-tem 1u found to be luoperchle, the reactor will initially be removed from service which will provide for a reduct.ica of the decay heat from the fuel and consequential reduct,lon of cooling requirements after a postu-Inled Losa of Cooinut Accident.

If t.he malfuncti.;. i. enw. L be rapidly corrected, the reset.or win be cooled t.o t.he chutdoun evndi Llon using normal cooldown procedureu.

In t hin condit.Jon, retenue of rioston productu or dnmnne of the fuel elements in not. considered pounibic.

She plant operating procedures require immediate action to effect repairs of an inoperabic component and, t.h e re fo ri., in mont, caucu, repairu will be complet.ed in less then t.he specified allownble repair timco. The limiting timca to repair are intended to:

(1) Assure that opernbility of the component will be restored pewipt.ly nini yet.,,(.?) Allow sufficient, t.ime to effect repairs using safe and proper procedures.

The lenhage rate limit for the core spray recirculation system hect exchanger has been established to asnore detection of any degradation of the integrity of the heat exchanger.

By Coimaission Memoranduiin and Order dated May 26, 1976, Consumers Power Company was granted a plant life exemp-I tion from the single failure criterion requirements of 10 CFR Part 50, 50.46 and Appendix K, Paragraph I.D.1 for the specific case of a Loss of Coolant Accident (LOCA) caused by a break in either core spray line. This exemption 'I was based on conditions specified in the Memorandum and Order and supporting NRC staff documents with which I

Consumers Power Company has complied.

b 3

Consumers Fowet Company has requested an exemption for Cycle 15 operation from the single failure criterion The NRC staff has granted the exemption for one cycle of operation i

Paragraph I.D.1, Appendix K to 10 CFR 50.46.

pending completion of tests of the original ring spray nozzles.

88-5 Amendment No. 16, 15

^

~~

Gurveillance Hequirement.

1,initing Lorutitions for Operation 11 3.3.ls COilTAlffMEtiT SPRAY SYSTE4 11.4.3.h COflTAIN14Ef1T SPRAY SYSTD4 Applicability:

Applicability:

Applies to the operating status of the con-Applies to the testing of the containment tainment, spray syntem.

spray system.

Objective:

Objective:

To assure the capability of the containment To verify the operability of the containment spray system to reduce containment pressure spray system.

In the event of a Loss of Coolant Accident.

Specification:

g Specification:

A.

Once each operating cycle, the followin6 A.

Durin6 Power operation each of the two shall be perfonned:

containment spray systems shall be 1.

Automatic actuation of the contain-operable, except that the power supply ment spray valve M0-706h (with water breaket.(52-2B45) must be locked open to preclude inadvertent operation of MO-7068.

flow manually blocked).

2.

Calibration of flow instrumentation.

a B.

If Specification A is not met, normal orderly shutdown shall be B.

At 1 cast once every six (6) months, initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the except for periods of continuous reactor shall be shut down as des-shutdown when the following shall bo. performed prior to startup:

cribed in Section 1.2.5(a) within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and shutdown as described in Section 1.2.5(a) f. (b) within the Verify operability of power-operated following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, valves required for proper system h

actuation.

C.

Ope'rability of the fire water supply Surveillance of fire water supply and and recirculation systems is governed C.

by Specification 11 3.1.h.

recirculation systems is governed by Speci fication.I 1.4.1. 4.

Instrument channels shall be tested and D.

calibrated as listed in Table 11.h.3.h(a)

E.

Each month verify that power supply breaker 52-2B45 for MO-7068 is locked open.

11-14 Amendment No.16,15

Elit'fd following a Loss.of Coolant The contairu.1cnt spray syste:r, are provided to reduce pressure in the containment They are initially supplied from the fire water system and Inter by the core spray recirculation Accident. They are not required to be in service at renetor coolant temperatures of 212*F or below because the nyctem.

renultruit tous of Coolant Accielent prer.sure is not sufficient to pressurize the containment.

The specified Operation of only one system is sufficient to provide the required containment spray flov.

flow of approximately LOO gpm is sufficient to remove post-accident ccre energy releases including a substan-tini chemical reaction involving hydrogen generatich to below design values.

I If a component is The operable status of these systems and. components is demonstrated by periodic tests.

found to be inoperable, it will be possible in most cases to effect repairs and restore the system to full If a single system becomes inoperable, a redundant system has opernbility within a relatively short time.

been provided with the ability to perform the spray function, but it reduces the redundancy provided by Q

pinnt design and limits the ability to tolc ate additional equipment failures.

Initiation of the containment spray system assures that containment design pressure vill not be exceeded due It has been conservatively calcu-to hydrogen generation assuming the core spray systems do not function.

lated that the energy release following a complete core meltdown (assuming no containmer.t. spray systems or core spray systems operate) would bring the containment pressure to approximately the de ign value (27 psig)

Subsequent LOCA analysis sy: stem modifications about 15 minutes after the postulated accident had occurred.

2 generation such that it is no longer significant and calculations show that and regulations have limited H Thus, the containment sprays are not required to prevent containment design pressures from being exceeded.

allow the operator adequate time to evaluate and block actuation, if system ope The backup containment spray valve (which may be normally operated upon failure of the primary containment spra valve) is disabled to preclude a single failure or inadvertent opening during a LOCA.

References:

1.

FHSR, Section 3 17, 1966.

Additional infor1 nation in support of Proposed Technical Specification Change No 8 dated March 2.

Safety Evaluation by the Research and Power Reactor Safety Branch, Division of Reactor Licensing, Consumer 3

Power Company, Proposed Change No 8 dated April lls,1966.

11-16

• Amendment No. 19, 15

-