Text

Proposed Tech Specs Re Reactor Protection & ESFs Trip Setpoints
	ML20236T268
Person / Time
Site:	Millstone
Issue date:	07/21/1998
From:	; NORTHEAST NUCLEAR ENERGY CO.
To:	;
Shared Package
ML20236T265	List: B17190, Application for Amend to License DPR-65,changing Tech Specs 2.1.1, Safety Limits - Reactor Core, 2.2.1, Limiting Safety Sys Settings - Reactor Trip Setpoints & 3.3.1.1, Instrumentation - Reactor Protective Instrumentation; ML20236T268;
References
	NUDOCS 9807280068
	Download: ML20236T268 (76)
	v • d • e

_ - ____-__ - -______ - - _ __ - _-_ _ _ _ _ - _ _ _ - _-_ - _ _ _ - _ _ _ -

Docket No. 50-336 B17190 -

Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints Marked Up Pages l

l l

July 1998 9907280068 980721 PDR ADOCK 05000336 P

PDR i

s

l INDEX

-NOVGmbGi 3, i335 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

(

~

SECTION 6, &ar,& &~'.L~~ Za./.h udse s... 3/Y 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.....................

3/4 7-1 Safety Val ves..................... '. 3/4 7-1 Auxiliary Feedwater Pumps...............". 3/4 7-4 Condensate Storage Tank................

3/4 7-6 Activity 3/4 7-7 Main Steam Line Isolation Valves 3/4 7-9 ain Feedwater Isolation Components (MF'ICs)......

3/4 7-9a 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 3/4 7-10 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM 3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM 3 /4 7-12 3/4.7.5 FLOOD LEVEL......................

3 /4 7-13 l

3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM.......

3/4 7-16 3/4.7.7 SEALED SOURCE CONTAMINATION..............

3/4 7-19 3/4.7.8 SNUBBERS 3 / 4 7 - 21

(

l 3/4.7.9 DEL ET ED........................

3 /4 7-3 3 3/4.7.10 DEL ETED........................

3 /4 7-33 3/4.7.11 ULTIMATE HEAT SINK 3 / 4 7 -3 4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 3/4 8-1 Operating.......................

3/4 8-1 Shutdown 3/4 8-5 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS...........

3/4 8-6 A.C. Distribution - Operating.............

3/4 8-6 A.C. Distribution - Shutdown 3/4 8-7 D.C. Distribution - Operating.............

3/4 8-8 D.C. Distribution - Shutdown 3 /4 8-10 D.C. Distribution (Turbine Battery) - Operating....

3/4 8-11

(

MILLSTONE - UNIT 2 VIII AmendmentNo.77,M,H,1H)ff ms in,in,in.

L--

February 2, 1976 No CH44GE I

Fof %"Rd/n6%

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS @L P 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and maxi.

mum cold les coolant temperature shall not exceed the limits shown on Figure 2.1 '.

~

APPLICABILITY: MODES 1 and 2.

ACTION:

Whenever the point defined by the combination of maximum cold leg temper-ature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

i i

REACTOR COOLANT SYSTEM PRESSURE

)

2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1. 2, 3, 4 and 5.

ACTION:

4 MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor. Cool' ant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

MODES 3. 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia.

l reduce the Reactor Coolant Systen pressure to within its limi.t within 5 minutes.

)

MILLSTONE - UNIT 2' 2-1

g 5

5 3

8 o

2 4

6 0

g 0

0 0

g 0

8

'v F

~

I G

UR vOLR FFLOF TTUV. ',

AFEE LOauPO HHNE.

E IL t SA URs ER EECS 2

vTSC CI yA ES 2

0 EH TT TNsA PR CMRE sEHO USCl E

OI TL.

1 M AR AT l

O O-RN A A N TR 1

O IU N N C EI I F I

T L

M NL N SP O

eE N M A OA OU TO EI ND PM I W Se R

S R

T NN L

E 0

EFT N YC E

E U

A T

N O

RV SL A

4 A

I EO L

AA E

A, C

0 M SO A

C T

Y N R A L

TL I S C

L A

I U SS O

AL OR P

NE E

a R

N Cl RO ETu s

G A M.

P l

g T,

C S UT OW SS E A

O A AE RA LPS A.

B I EU S L

R FTD E

0 E

I N

OT C

MCM U E

I I E R.

6 T

T 0

YNO E

j TFD E O,

ii SI M.

P E

E E T E R

R HDO N A

M E

T.

T PI B Il.

1 A

F O

L E

ONE N

A R

M C 0 T[

A m

1 8

R 0

G N

N O

I f

3 S

F.

A A

F I

E E

1 l

D 0

5 0

g Y

T H

L E

\\

l R

,'n g

H M

il., *.;.:.;

.l

l.

.l el ^,

{

I A

T L

Q. pE <- m,,, 53 P

1 h

\\

=

O 2

, m9 < " m,f <->gN W

0 i* g

\\:.,,

(.

M. * :

, q F

E gf'o d' -

.Qecgh O

R 3$=

i 4.

13t>g

,,8' U

o, o=

'.j R

b d%,

. i

'7 2

N.

.i..

t ;; '

~

R S

\\

i_

E A

1 y,

.R g :

y' ig'

'~

s

4~

h I

s C

4 1

0 0

p c-i

_ R 5

'n..

's t

1.

T y t

C

_ O

_ O r

-g=

T," e" A

. ='

'p

~r U L

's N

r

'i

_ A 1

C N

6 Q

\\

C T.

0 A. i f.

n E

P i

P I

3?,

.j, e -~

'k..

UM A

. l B

6 P

9v

.*W E

(,

'I

~f,;.. -

.e L

k',

A+..

S

\\

i P

.1"U,

. O s

.f ;

Y

.- 51

~

P

~

}'

O 1

i P

8 E

E 0

i.

A

,Y -

R

' i-d',

R i

\\

~A l

.. n '_r A

~

,f[

+

l T

O I

\\~.-

2 N

N G

v w.

~.

. '.. f.

2

i 0

0 m$zaDz[

"'m kcEECes z N *N *g %i t

2 3

Qm Nmm -

August 1, 1975 fl/o C04AiGE

}

FoR JAthKAl#hrw SAFETY LIMITS AND LIMIilNG SAFETY SYSTEM SETTINGS OmV 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.21.

APPLICABILITY: AS SHOWN FOR EACH CHANNEL IN T BLE 3.3-1.

ACTION:

With a reactor protective instrumentation setpoint less conservative than

. the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.

1 b

KILLSTONE - UNIT 2 2-3

R n

t E%a n

h s

W7 c

oe a

O a

tn l

P4Rf og e

i

'.1 E o. on dl W

ci e

AsO%R t

ttd gl MRf P 7. E ra sin e

uma W or S

EoL6O te v

ji E

H A0P cp e

dl1 U e TmM1 ao Lr a L

l uR L e o

e2 A b mE 5A rs rt th V a iH M

p a g tr i ea nt2 c4 E

lt' hTfR fm

.i i i a ae ods.

oE ou m s s i Wn pee) i L i mD H

p p p p s e ter4 S p EmT 5 r

S A p 5aTu g4 0 4 #p g ecu(

T N A 7.hRiE AmD sxg 3

0 2 f m

ei2 I

O 0h 2 4

M L t 9t xT 9t 8 2

P a p F -

i

- +. 4 e

it I

L o ifaA t

rof 2 L

A N swomR 1w 1 ss1 2s Tno2 T

N r

l e

0 h

P t

T t

ir g

o E

eo S

Rf tn n

)

Eo ni h

4 f

P W

L at c

of(

o1 I

OtA a

t o

.T l a R

PnM or e

2 R

T iR oe ds-r 6. O e

LoE cp ee2 P

s 2. E N

A pH o

t n s

1 O

NtT r

l si2 e4R I

Re os e

ul l

2 T EsD t p v

j d nS 7

e dtn eoT A

H E

cm 2 T e

TmT au 9

Lr aia T

l uA ep 3

hiI o

m ttN N

N b emR r

rt ti1 aI E E I

a vi ea nl -

M O c onf f

L 4

g t r i scL B

2 P

I i i i

i a ae oe A R bio oh m s i Wn ph2 ginfG T T T

l am t

p p s e tt E

p

%i r

S ncI S

p p

%g e s

% a 6. R N

6.h4E

7. w 5

i eT A

9 sde I

t pA 0

P 1 0

m er aSR 1 w 3 8

6a peu I

t 9t1W 9o 8

+. 4 3-e icg E

r E

R o V

O l

e rP i

T N sw5P 2f 1 s1 2 2s TeF op t

rxi I

peO TC E

E sdR T

peO O

mrC RP s

h h uu i(se p

g g -

ah i i R

m 4 et O

u H H e m

T P

r i

C u

h r

t ef s

r it A

t w

p e e s e

wad E

p n

R T

o m

r r e t

y re i

a l

u u u r a

t w

i I

r l

F P

s s P W

N T h o owf i

U.

g o t

t e e r r) n fl c s s s

oi l

r C

n n

r r o o5 e fe i

L o H a

a P P t) t( D t

p A t r

l l

N c l

o o

ow r t r( rw r an a5 a nt s O a e t

o oo e n e eo e I

e v cg C

CL z e n nL w ol u l ae T R e an i m e e

o) ool C

l ei r) r-r n G) G - P3 Coa N

l Rt o1 o u i 2

(

cv U a r a

t( td s a m( ml l

r F u e rr c

ce s t a

ae ah o rm n w ue aw ae e n ew ev cg t ou a o op eo ep r o t o te oi ct m M P FO RL RS P C SL SL LH aci ean Reirm n

gee 1

2 3

4 5 6 7 8

9 ihh sTT e

D

ab

,hOk'gqe.

[

g gg F* s ? w5 y? $

,E gybyM?gw

v c

A d

r R

i M

n e

E t

R a

n H

a E

e T

m H

g o

T s

t n

t e

m di3 e

u n

i a

em-h a

e t

e ti2 w

h n

sl e

w i

4 u

2 d

b a

4 M'

je e

d t

S dhs v

l e

r h

E o

l v

e c

U ate.

m a

o c

a S

r

Lo e

h m

n e

Age tdu) r s

e u

n e g4 V

f ei(

r m

1 ocF g

y s

r o

E

'px 4 i l

s y

o r

L tef -

s l

a l

s f

B e

o2 p a

p l

s A

st c

y a

e e

N os2 0 i

b c

c n

O pne 0

t i

o o

L i

nd 5 a

t r

L rof n m

d a

p S

A TtI a 2 o

e m

s T

t t

o d

r I

u r

t n

e M

a e

u a

t I

s a

t e

n L_

l i

b i

e a

m T

b n

s y

o n

N}

l lti t

I l

l l

i a

O o

a l

l t

r P

n

)

h u

a a

t T

4 f

h E

of(

s l

s u

o t o e

c w

S 4

N s

r s

l t

P ds -

O s

a s

a a

ee2 I

t n T

p a

c f

I bgE s

p o

R y

T si2 A

A y

ul T

b s

t N

j d O

o C

t u

1 O

dt n N

I aia Re n

p V

2 T T

ti3 L

W <g 1 R

e t

E m

u O 9, a Aj e

o A

N nl -

B 2 T 2

A P

r I

i N

qMA u

d 0

oe g

T E E s

e P

ph2 i

L L M T

tt s

e L

a r

Ac, B U E

e s

p e

e M

s M m

e A R S

sde R*

El :

T T R

n

~,H S

er 0

E o

P peu 0

H' s

N T <^, :*,

i e

I I

icg 5

T d

t R

rxi u

c E

T TeF 2

D4 0

D u

E48; l

V E

T 7gT c

a I

n T

A r

A e

C R

i R

h E

t f

t T

)

f W

n O

1 o

o.

n i

(

R RP

%R

%E o

o e

5 E d

5W p

e t

R r s c

W e

1O t

wO s

P e

O a p i

a s

w s

T s m w

oP ap u

s u

'u o l

oL t

C T e P aL eL p

lA p

A c

I r

r b A yG eM a

i E

N P t d -

M bG bR r

P U

n y

d R

7 E

t s

w a He e E y

dH l

L o l

- r s H l e eT e

e A L o

u sT l v s

h n

N / o es a

ae sD t

n O n C ns p D u t.

aE a

I i

ie E

n; pT f

h yT a

yA o

c T g r br b

C r o rP A

m-bR N a t u

e e

ef n.

u R

s r

U N cg T) b f bt bo s

o F

an 3

l ei f(

o e

f y

y -

y%

c a Rt o

m a

d a%

ad a5 n

f r rr si m5 me m1 a

o e ue su v

w h

op ol p2 po p2 o

h T FO LF m

i-l c

i i

rs re r5 l

a L

Ti Tr T4 Ca E

0 I

)

1 I

1 2

3 4

5

(

g3 ?Cr ~

M.

5 r h o g E *** [

n e

2S

FGbruary 2.1976 6

ff./0 C/Ul/vCE l~d.$

TA/RI//nM.tw Oniq.

i 1

l l

1.0

.. _......13.,13,.<

.4__

s._

1 1.._

t

_ _. 1 s-

... _.._4 t.-

... i. t..

. 1'#.

i

1...

1

72..s

t..m.t.

,. h..

~. _.....T.._.

. _.. S.

...a g

.... 3.

4._=.3:

1

_~_.

J...!.....

4

-.:r.I:

..t

_.e s.

.1.

0.6

=

N....

SLOPE.=..1.0

_.I

.e e

e.e.

.. =.

2

~. _ -

g,

=_

2 l

+

j 4

=_.

0.2

..-. f..

1

.m

. p.c

,.s.

,-1.

=

~

0 0.2 OA 0.s 0.8 1.0 FRACTION OF RATED THERMAL POWER FIGURE 2.21 Local Power Density - High Trip Setpoint Part 1 (Fraction of RATED THERMAL. POWER Versus OR I 2

Mll.t. STONE - UNIT 2 26

(

a November 10, 1992 l

A/o CHAN6E

/~02.TAth!Pt+7%eu cozy N

1 6

0.ti i.

!ith I

i' j

i

.i j

-jJ 8

8 g_. g. -

l' i.l -

' l' I

dNACCEPTA8'LE OPERATION T:

.l -

j-i.

f....

j i

REGION

. a..

I

i l 0.4 l'

i (0.4. 0.65) j' l 7_

u.

g.

i I. :

i:

eN i

.i

.l.

i s

til:

i:i

ll:

l" j-N !-

l i

T l."

. l ;-

."l.'- Jl it N

i-i t

1

.j..

i

\\

(.2.1.0);

02 a

i 6

4 i

_i _

3..

4 o

s..

i 1,_ _ !

j.

.i l

l-l-

i i-i

(

l i_-

i..... i. _,

.2...

2.,......:..:.

j ACCEPTABLE OPERATION l i

Y o

REGION t

g i

g l

q..

l

.L..

e i

....!....8._.

(0.0, 1.25) i t

4 i

i

)

^!

i j

l;.

i i

-0'2 i

t i:

{-

j j

f ( 0.2,1.0) i.

j.

j, y r-

"] j Me I i

3

.e 1

0.4

[;.

ll,,,.

.j,.,

(-0.4, 0.65) i

.[..

l-l l-j.

l UNACCEPTABLE OPERATION i

REGION 4

s-l i-i i

i. g. a:

t i

l' I

i I

I :

-0.6 0 0.2 0.4 0.6 0.8 1.0 1.2 1.4 OR2 FIGURE 2.2-2 Local Power Density - High Trip Setpoint Part 2 (OR 2 Versus Y )

g i

MILI. STONE - UNIT 2 2-7 Amendment No. 38, 52, 164

v.

r+.

0 4.

g i :

g

=

j..

. ~

ip j:'g.

5

[

g.% - l

!g*

9. :

3 4

.Y A

0 Ig Is.

0 r

.0

- +

I 2

i.

.+-

/**

)

8 g

g 3.

b't.

h.;'

3.:

5:

1 t

. 7

-I I,

s 4.i 3'

t

(

l

-.u.,

0

,I 8

g M p--l

\\,N.0

.= -

.'u x i 2

. g,..

2 sr e

V

. 7 -

g 4

1 Y

+

(

N ;

1ha 1

B ;'

+= l D

t t

1 r

O.

."1

-=

..% =,

a 0

x P

q

.s t

1 s

u i

5 n

o 2

p g

b..

t 2

n

=

i" 2

l

[8 e

P i"

2 S

I i

P

-:s.

i"

?!._

.N ip 3-T,

.t.

i u.

0

.- H i-7-

i r

[i.

T 2

2

. g g

e r

Y K.

s u

s i

D e

N t

. s 1

r A

0.

P I

g.

i g.

l 2

i. -

l.i.

w

. j,i l i.

o L

/

n R

%g

y-.

2 a

g y

p--

ig g:,

r O

g 9

0 M

A

K 2

l 0 :

1 l.!,

a

=

1 m

a

^

1 r

B N

+

D

_8 N3 h

e O

.,-q.*

g T

Y 3

3 i- '

4 0

E:

'.g.

i65 l

l ig 2

1 1

3 g

=.

.0

.s 2

R ;~

E H

u L

{gt

=

W ;; ' :, t

] ;;,. i g

i

= : : --,

E y

R I

i~:l

,_g 4

U

. :i A

c 0

G

g-I N-

=

l.h 'l F

. : L_ ;.

I 5

6 5

4 2

1 0.0-1 1

1 1

E, r. $O2m c2.-

N mh g

l' l

May 12, 1979 i

hu C&6d Ra.nuR&ngr.2w oney i.

1.2 Nb8" A x OR 1

g

/

Masso

/*

4n t/

f qpTRIP. 2215 x ODNB + 14.28 x Tin -8240 7!~

1.0

l;[

1.00

..j

_,,. _ f

-- n - +-

..y 0.885 mg -

/

.; 1.00 l

p ^

/

.:0.83 "-

O.8

/:

/: ;

i:0.7ii /.:-

n i./ : ;-

!!O.6j[.j %.

...... L :_

n!...

7-I-

l l-d i!

):(

- {- -

M..ii.,

3-e-

_2..

QR 0's

- ' :": jn

-l:: :

i

.i-

j ;

j::-

A

ii[li:

S!!I

'i!

!Y :

I.!';

illii

'i f"

ip

4

/i

.l 2

[

'I m[.

.i

I 0.45 4

- fi..-

-r-l

' )

.. :.):'

E;:

2 0.4

M.,

i:l~i i'

d() ~

'=

t..

i fil

?

I'

.=

ii.15 niii

. # is :i:"

it-

p

/;

l 0.2 i:if ilf; "l. '

l.

! i.

s........::

m '-

..l;m - uj:;-

l.:

p 3;m- :--...

i.

__1/

ii!!;

.:it:

'i:

/1...

i::

iil :

i j:-

J.,

0 0.2 0.4 0.6 0.8 1.0 1.2 FR ACTION OF RATED THERMAL POWER f

FIGURE 2.2-4 Thermal Margin / Low Pressure Trip Setpoint (Part 2 Fraction of an RATED THERMAL POWER Versus OR )

3 J

MILLSTONE. UNIT 2 2-9 i

9

-Hard 20, im 2.1 SAFETY LIMITS 1

BASES 2.1.1 REACTOR CORE i

The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 21 kw/ft. Centerline fuel melting will not occur for this peak linear heat rate. Overheating of the. fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result 1n excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coeffic'ieTiG DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the XNB correlation. The XNB DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.17. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

1 The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the minimum DNBR is no less than 1.17.

The limits in Figure 2.1-1 were calculated for reactor coolant inlet temperatures less than or equal to 580*F. The dashed line at 580*F coolant inlet temperatures is not a safety limit; however, operation above 580*F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POWER levels higher than-14M of RATED THERMAL POWER is prohibited by the high power level trip setpoint s ecified in Table 2.2-1.

The area of safe operation is below and to the left of these lines.

//U.%

i r

MILLSTONE - UNIT 2 B 2-1 Amendment No. 7,52,91,///

1

=

j October 6, 1930

/UO CffAWGE 1

Fc)$ IA/F#m+r.tw QAJL Y e

l l

i l

(

i l

THIS PAGE INTENTIONALLY LEFT BLANK.

i

~.;.-..

J I

f l

l

(.

MILLSTO'NE - Ohit 2 B 2-2

' Amendment No. 52. 61 i

.v.

a l

t L

march 20, 1003 SAFETY LIMIT BASES The conditions for the Thermal Margin Safety Limit curves in figure 2.1-1 to be valid are shown on the figure.

The reactor protective system in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated combinati conditions for reactor coolant system temperature, pressure, an.on of transient level that would result in a DNBR'of less than 1.17 and preclude the existenced thermal p of flow instabilities.

2.1.2 REACTOR COOLANT SYSTEM P SURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section 111 of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure.

The Reactor Coolant System piping, valves and fittings, are designed to ANSI B31.7, Class I which permits a maximum transient pressure of 110% (2750 psia) of component design pressure.

The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

~

The entire Reactor Coolant System is hydrotested at 3125 psia to demon-strate integrity prior to initial operation.

l 1

MILLSTONE - UNIT 2 B 2-3 Amendment No.7/5[,M,[//

~

hw c, 33c

.2.2 LIMITING SAFETY SYSTEM SETTINGS

~

BASES

{

t.

1.

l 2.2.1 REACTOR TRIP SET POINTS

~

The Reactor Trip Setpoints specified in Table 2.2-1 are'the value,s at which the Reactor Trips are set for each parameter.

The Trip Yalues.have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding th~eir'safrty~1fmiti.~ ~0peratlorf with a Trip e

Setpoint less conservative than its setpoint but within its specified Allowable Yalue is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed to occur for each trip p

used in the accident analyses.

L Manual Reactor Trip The Manual Reaci.or Trip is a redundant' channel to the automatic protective instrumentation channels and provides manual reactor trip F

capability.

t

{

' Power level-Hioh The Power Level-High t. rip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer

[

Pressure-High or Thermal Margin / Low Pressure trip.

The Power Level-High trip setpoint is operator adjustable and can be set no higher than 9.6% above the indicated THERMAL POWER level.

Operator j

action is required to increase the trip setpoint as THERMAL POWER is increased.

The trip setpoint is automatically decreased as THERMAL POWER decreases.

The trip setpoint has a maximum value of 106.6% of RATED THERMAL POWER and a minimum setpoint of 14.6% of RATED THERMAL POWER.

Adding,to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximuni actual steady-state THERMAL POWER level at sthich a trip would be actuated is +1tf of RATED THERMAL POWER, which is the value used in the accident analyses.

Ill,(, %

Reactor Coolant Flow-Low The Reactor Coolant Flow ' Low trip provides core protection to prevent Df;3 in the event of a sudden significant decrease in reactor coolant flow.

Provisions have been made in the reactor protective system to permit MILLSTONE - UNIT 2 B 2-4 Amendment No.

t

.(,

c 'U '

4

a!2re 20

'w-LIMITING SAFETY SYSTEM SETTINGS ffo % 4 o h <. h w M A BASES Reactor Coolant flow-tow (Continued)j Q4 &p@ o4@h/,6as n.f-s a w o n r

gewah operation of the reactor a[t reduced power if one or two reactor coola are taken out of service.

The low-flow trip setpointf and. Allowable Valucg 4e= the veris;.i reecter ceelent yo y cc Lir.: tic s have been derived in con eration of instrument errors and response times of equipment involved to Aaintain the DNBR~ above 1.17 under normal operation and expected tran For reat-operation ith only two three reacto coolant pumps erat uig, the Re tor Coolant ow-Low trip etpoints, the wer Level-High rip j

seti nts, and t hermal Marg /Luw Pressure ip setpoints automatical-i t

ly hanged whe he pump cond ion selector s tch is manuali set to the estred two-or three-pump sition. Chang g these trip s points durin two

{

and three ump operation even' the min um value of DN from going low I.17 d ing normal oper ional transien and anticipat transients y en only two p<- three reactor olant pumps ar operatino.

l!

Pressurizer Pressure-Hiah

/

safety valves and main steam line safety valves, provid system protection against overpressurization in the event of loss of load without reactor trip.

setting (2500 psia) of the pressurizer code safThis trip's setpoint is 100 psi be j

operation with the power-operated relief valves valves and its concurrent tion of the pressurizer code safety valves.

voids the undesirable opera-f Containment Pressure-Hiah fagiumaleh C

j is initiated concurrently with a safety injection.The Containmen is identical to the safety injection setpoint.

The setpont for this trip b

Steam Generator Pressure-tow

--fyh The Steam Generator Pressure-Low tr provides protection against an cooldown of the reactor coolant. excessive rate of heat extraction from the The setting < f 580 p ia-is sufficiently below the full-load operating point so as not to interfere with normal opera-tion, but still high enough to provide the required protection in the event of i

excessively high stea.m flow.

Tht: Octting.::: u;ad with en uncert:inty f;ctor

+& 22 p;i ir the occide t enelyses. --

n

})

MitLSTONE - UNIT 2 B 2-5 Amendment No. 52, 61, 139

March 20, 1989 A)O C04+CE LIMITING SAFETY SYSTEM SETTINGS

[og.7a/Ageng 7,t o+

i BASES

('

Steam Generator Water Level - Low The Steam Generator Water Level-Low Trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design l

pressure of the reactor coolant system will not be exceeded. The specified srt Mnt provides allowance that there will be sufficierit water inve'ntory 1.n t. -..- - tM

eam generators at the time of trip to provide a margin of more than. --.~.-._

m'. twas before auxiliary feedwater is required.

Local Power Density-Hiah The local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring,- is provided to ensure that the peak local power density in the fuel which corresponds to fuel centerline melting will not occur as a conse-quence of axial power maldistributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2.

The AXIAL SHAPE INDEX is cilculated from the upper and lower e'x-core neutron detector channels. The calculated setpoints are generated as a function of THERMAL POWER level.

The trip is automatically bypassed below 15 percent power.

The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints. In addition, CEA gr.oup sequencing in accordance with the Specifications 3.1.3.5 and 3.1.3,6 is. assumed.

Finally, the maximum insertion of CEA banks which can

[

occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

Thermal Marcin/ Low Pressure The Thermal Margin / Low Pressure trip is provided to prevent operation when the DNBR is less than 1.17.

4 s

i I

i P

l MILLSTONE - UNIT 2 8 2-6 Amendment No. 28,41,52,61,139 l

E___.__________

~

-rd. m c,

'., 10 m l

l l

j LIMITING SAFETY SYSTEM SETTINGS t

BASES Thermal Marcin/ Low Pressure (Continued) l The trip is initiated whenever the reactor coolant system pressure signal drops below either 1850 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of AT power or l

l neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximus AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function.

In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.

i Finally, the maximum insertion of CEA banks which can occur du~ ring any anticipated operational occurrence prior to a Power Level-High trip is assumed.

Thermal Margin / Low Pressure trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and processing error. A safety margin is provided which includes: an allowance of 5% of RATED THERMAL POWER to compensate for potential power measurement error; an allowance offF to

'MI-compensate for potential temperature measurement uncertainty; and a urther allowance of 74 psi to compensate for pressure measurement error, trip system l

)

processing error, and time delay associated with providing effective termina-tion of the occurrence that exhibits the most rapid decrease in margin to the safety limit.

The 74 psi allowance is made up of a 5 psi bias, a 19 psi pressure measurement allowance and a 50 psi tima delay allowance.

loss of TurMne A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL POWER.

This trip provides turbine protection, reduces l

the severity of the ensuring transient and helps avoid the lifting of the main l

steam line safety valves during the ensuing transient, thus extending the service life of these ='=#

No credit was taken in the accident analyses for operation of this trip.

Its functional capability at the specified trip setting is required to e'nhance the overall reliability of the Reactor Protec-tion System.

l/alves I

)

Amendment No. 5, JJ, J # h MILLSTONE - UNIT 2 B 2-7 0048 l

March 20, 1989

,AJ O C //4A/GE i

LIMITING SAFETY SYSTEM SETTINGS gg rapogmgmy BASES.

O*6I Undersneed - Reactor Coolant Pumos The Underspeed - Reactor Coolant Pumps trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant pump speed (with resulting decrease in flow) on all four reactor coolant pumps. The trip setpoint ensures that a reactor trip will be generated, considering instrument errors ~1nJ Tesp~ont~e'tfuiffi in ' sufficient time to allow

~

the DNBR to be m&intained above 1.17 following a 4 pump loss of flow event'

[

l i

-~-

~

i HILLSTONE - UNIT 2 B 2-8 Amendment No. ;2,61,139

June 10,1996

,yd C /Mit/6E 3/4.3 INSTRUMENTATION g~og 7ev/-o/?/n Troy 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION O tt/ 4 LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and

~

bypasses of Table 3.3-1 shall be OPERABLE.

~

l APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLlNCEREQUIREMENTS 4.3.1.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE

) during the at power CHANNEL FUNCTIONAL TEST of channels affected by

~

operation.

The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip l

function shall be demonstrated to be within its limit at least once per 18 months.

Neutron detectors are exempt from response time testing. Each test l

shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels

column of Table 3.3-1.

l

)

MILLSTONE - UNIT 2 3/4 3-1 Amendment No. 77.198 om i

l l

L__________

i

go Sgmm p^

Rh t

f ORm O

I T

C 1

2 2

3 A

~

E

)

L d

BS d

(

)

Ar n

3 c

)c Cn a

(

Ly I

(

2 2

t P

P A

1 1

1, 1

SE MLL ll ED MNA NR I

NAE l

I iP MCO 2

3 3

3 NO I

T A

TN E

S M

LP U

EI R

NR

)

1 T

NT f

)

S A

(

a b

c a

c 3

N iO

(

l I

CT 1

2 2

'2

(

2 2

3 E

E V

L I

B T

S A

C

.L T

E OE T

NN O

N R

LA P

Ai l

TC R

O O

TF T

O 2

4 4

4

!1 ! >,

C A

w E

R o

L h

h e

c w

g g

h r

i o

i e

i f

g u

l i

L l

l f

s u

r H

s a

u e

rw s

r r

do

, w e

e s

e P

yL p

o r

r e

t y

l f

h l

u u

r a

t w

I r

g F

s s

P W

o f

T i

s s

s L

ee H

t e

e r

r n

/

nr r

n r

r o

o e

n iu o

a P

P t

t w D

i bs t

l a

ao g

rs T

c l

o r

t r

rL r

r ue I

a e

o e

n e

e e

a Tr N

e v

C z

e n

n-w M

P U

R e

t m

e e

o f

L r

r n

G Gl P

l od L

l o

u i

e a

t A

a r

t s

a m

mv l

m su N

u e

c s

t a

ae a

r sl O

n w

a e

n e

eL c

e oF I

a o

e r

o t

t o

h L

T M

P R

P C

S S

L T

CNU 0

F 1

2 3

4 5

6 7

8 9

1 3"rp,gA h5

m 3

'"s" k=

e U"

Ph th%

N yp O

I T

4 2

CA n

E L

BS 5

)

AE e

CD

(

IO 4

2

^

LM P

P 3

1 A

SE MLL UEB MNA I NR 2

3

..N A E I HP MCO

, ii' N

O I

S T

LP

)

c A

EI a

)

T NR

(

d N

NT 0

2 e

E A

u M

HO n

U CT i

R vt T

n S

o N

C I

S

(

.L E

OE 1

V NN I

N 3

T LA 4

4 C

AH 3

E TC T

O E

O TF L

R O

B P

AT R

OTC A

n E

o R

r tu en N w o cd it r

mu o

hh t

tS c

i a

r -

es a

Rp gr m

oo u

Lt P

T i

d I

en et N

go en U

nM pa a

sl L

R x ro A

u eo N

el dC O

dF n

I i

U T

W C

y N

U 1

2 F

1 1

x FmUM i c3 y

. ?'w i$

F ?

.~

~

ttaren zo, W8r IABLE 3.3-1 (Continued) 4 f/% g,,3, TABLE NOTATION N 7 " ' Y '*

o.,

drive system capable of CEA withdrawal.*With the protective system t (a)

Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be l

automatically removed when THERMAL POWER is 2 5% of RATED THERMAL L

(b) Trip may be manually bypassed h b; 7a0 pt: d r.

11 CEAs are fully inserted; bypass shall be automatically removedgen 700 pie.

(c)

Trip may be~ bypassed below 15% of RATED THERMAL POWER; bypass sha automatically removed when THERMAL POWER is 115% of RATED THERMAL 6 sam POWER.

9 ~~- N (d)

Trip does not need to be operable if all the control rod drive mec anismsgene j

are de-energized or if the RCS boron concentration is greater than or equal to the refueling concentration of Specification.3.9.1,.

(e)

Trip may be bypassed during testing pursuant to Special Test Exception l

3.10.3.

(f)

AT Power input to trip may be bypassed below 5% of i

THERMAL POWER.

(

t ACTION STATEMENTS ACTION 1 -

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STA within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and/or open the protective system trip breakers.

ACTION 2 -

With the numb'er of OPERABLE channels one less than the. Tota Number of Channels. and with the THERMAL POWER level:

15% of RATED THERMAL POWER, immediately place the inoperable a.

l channel in the bypassed condition; restore-the inoperable l

channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

L b.

15% of RATED THERMAL POWER, operatioh may continue with the inoperable channel in the bypassed condition, provided the following conditions are satisfied:

I 9

(

MILLSTONE - UNIT 2 3/4 3-4 Amendment No. J, E 7/, J S kf

O August 1,1975 TABLE 3.3-1 (Continued)

A/o C//dME s

)

Fott TwhRist strzon ACTION STATEMENTS gj 1.

All functional units receiving an input from the-bypassed channel are also placed in the bypassed condition.

2.

The Minim e Channels OPERABLE requirement is met; however, one additional channel may be-removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for I

surveillance testing per Specification 4.3.1.1 provided one of the inoperable channeh is placed in the tripped condition.

ACTI.ON 3 With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:

a.

< 5% of RATED THERMAL POWER, imediately place the Tnoperable channel in the bypassed condition, restore the inoperable channel.to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

)

b.

> 5% of RATED THERMAL POWER, power operation may continue.

ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, imediately verify compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, and at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter.

s e

MILLSTONE - UNIT 2 3/4 3-5

)

June 10,1996 i

/OO Cth9tv6E

(

non-ave.y i

==

0 THIS PAGE INTENTIALLY LEFT BLANK

@O MILLSTONE - UNIT 2 3/4 3-6 Amendment No. pp,198 0240 u___-__---__.-_-------__._--------------------

)3 gb 4' I

l ChoT i

tE iC iND l

WAJ 3

LP NLI I

2 2

2 II R

Ef A.

A.

SVE ERR N

11 1

1 1

N DU OS M

L A

S LN l N

/

T EOT

)

N NI S 1

l E

NTE

(

E l

U II M

ACT R

CU S

MM H

H H

S i

F l

t Q

E i

i'R Q

E

)

C N

3 N

O

(

A LI MQ L

ET L

NA NR I

))

E AB A.

24 A.

V l I

((

i R

CL N

DD R

R R

N U

A 1

S C

v, 3

N O

4 I

T E

A L

T L

B N

EK A

E NC T

M NE l

A.

U Al R

l C

N SS S

S S

N T

lC SN I

E V

w I

o T

L C

h h

e E

w g

g h

r c

T o

i i

e g

u i

i i

O L

l l

i s

l i

R r

P u

l s

u e

aw s

r r

ro R

w e

e s

e P

dL O

p o

r r

e t

y ly T

i h

l u

u r

a t

w I

C r

gr F

s s

P W

o i

A T

l e

s s

s L

- e E

R r

Il w t

e e

r r

n

/

er o

n r

r o

o e

n nu o

P a

P P

t t w D

i is t

r l

a ao g

bs c

l re o

r t

r rL r

r re T

a eaw o

e n

e e

e a

ur

~

I e

veo C

z e

n n -

w M

TP N

R el P i

m e

e o

U l c r

r n

G Gl P

l fd l

uT o

u i

e a

oi L

a rNA t

s a

m mv m

u l

A u

e c

s t

a ae a

r sl N

n w

a e

n e

el c

e sF O

a o..

e r

o t

t o

h o

M P ab R

P C

S S

L T

L I

T

- Q C

N U

1 2

3 4

5 6

7 8

9 0

F 1

2IG$5 e

g[

{[

gha?. mPO*

l

,ll;lilIii'

,1 i

~

?x G3 o

m

)4 l

d M

i CE n

l C a

li ND d

WAL 5

n LR a

NLI II U 4

2 2

2 EQ SVE ERR 3

1 1

1 0S a

DU H

)

1

(

L U

A

/

S LN S

T EOT

)

N NI S 1

d E

NTE

(

n M

ACT U

a E

iN

/

l R

CU S

M M

M I

F U

Q ER E

C N

N O

A LI L

ET

)

L NA

' d I

NR e

E AB A.

A.

u V

l l i

n R

Cl N

R N

N i

U A

t S

C s

n

^

o N

(C O

I T

A 1

T L

3 N

EK E

NC 4

M NE U

Ai A.

A.

l E

R l C S

S H

N l

L T

C B

S A

N T

I E

V n

I o

T r

C t

E u

T e

m O

N e

R t

P c

s i

y s

R m

r S

r O

h o

e T

t t

n k

C i

c o

a A

r a

i e

E a

es t

r R

gr Rp c

B oo m

e

'l t 2 u t

p t

P o

i T

en d

r r

I go et P

T N

nM en U

a pa rc r

Rx sl oi o

L u

ro t g t

A el eo co c

N dF dC aL a

n e

e O

i W

U R

R I

T C

N U

1 2

3 4

F 1

1 1

1 E ua 3 '"

EZ" Mb s

(sgae E M N

D C

.l llll L

September 2, 1976

.s g o C //y1 N 6 E

}

_ TABLE 4.3-1 (Continued) g yygg7.>

TABLE NOTATION N

With reactor trip breaker closed.

.(1)'-

If not performed;in previous 7 days.

-(2).-

Heat balance only, above 155 of RATED THERMAL POWER; "I-

~

8 adjust " Nuclear Power Calibrate" potentiometers to make nuclear power signals agree with calorimetric calculation.

During PHYSICS TESTS, these daily calibrations of nuclear 14 power and AT power may be suspended provided these calibra-tions are performed upon reaching'each major' test power plateau and prior to proceeding to the next major test power pTateau.

(3)

' Ab~ovi'15% of RATED THERMAL POWER, recalibrates the excore

'8 detectors which monitor the AXIAL SHAPE INDEX by using the incore detectors or restrict THERMAL POWER during subsequent operations to < 90% of the maximum allowed THERMAL. POWER level with the existing Reactor Coolant Pump combination.

nl4)

Above 15% of RATED THERMAL POWER, adjust " AT Pwr Calibrate" potentiometers to null " Nuclear Pwr - AT Pwr".

During j

~

PHYSCIS TESTS, these daily calibrations of nuclear power and AT power may be suspended provided these calibrations are 14 performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.

l.

1

)

MILLSTONE - UNIT 2 3/4 3-9 i

i June 10,1996 INSTRUMENTATION A /o C //W 4 s I

ft)G h P 4 m + Ze w 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION g ty LINITING CONDITION FOR OPERATION l

i 3.3.2.1 The engineered safety feature actuation system instrumentation.

channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their l

trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4.

l l

APPLICABILITY: As shown in Table 3.3-3.

ACTION:

With an engineered safety feature actuation system instru-a.

mentation channel trip setpoint-less conservative than the value shown in the Allowable Values column of Table 3.3-4, either adjust the trip setpoint to be consistent with the value specified in the Trip Setpoint column of Table 3.3-4 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or declare the chantiel inoperable and take the ACTION shown in Table 3.3-3.

b.

With an engineered safety feature actuation system instru-I mentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature acutation system ins'; amen-tation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operstrois' "

during the modes and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affects by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.

i NILLSTONE - UNIT 2 3/4 3-10 Amendment No.198 0247

l

, " +

March 1, 1979 Nd'C0/f;OGh INSTRUMENTATION MMMN

~

0A/ W

~ SURVEILLANCE REQUIREMENTS (Continued)

~

l j

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of*each ESF l

function'shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at.least once every N times 18. months where N is the total number of redundant channels in a specific ESF function as shown in the " Total.No. of.Chenpd s". Column of Table 3.3-3.

4.3.2.1.4 The trip value shall be'such that the containment purge effluent shall not result in calculated concentr4tions of radioactivity offsite in excess of 10 CFR Part 20, Appendix B. Table II.6 Forghe purposes of calculating this trip value, a x/Q = 5.8 x 10' sec/m shall be used,when the syst

. and a X/Q = 7.5 x 10-gm is a}igned to purge through the building vent sec/m shall be used when the system is aligned to purge through the Unit 1 stack, the gaseous and aprticulate (Half Lives greater than 8 days) radioactivity shall be asusmed to be Xe-133 and Cs-137,5respectively. However, the setpoints shall be no greater than 5 x 10 cpm.

s h.

tf MILLSTONE - UNIT 2 3/4 3-11

8

)

a 9

(

a 1

E 4

3 4

4 4

3

(

L BS 2

AE 3

3

)

3 3

3

)

2 CD e

IO

(

(e y

LM 2

2 2

R 2

2 2

r P

a P

u A

1 1

1 r

b e

I F

N O

,o I

T SE c

T UEB A

MLL N

MNA pw E

I NR M

NAE U

I HP R

MCO 2

3 3

2 3

2 2

3 3

T N

S I

2 M

E T

S S

LP Y

EI S

NR 3

NT

)

N A

b 3

O HO

(

I CT 1

2 2

1 2

1 1

2 2

3 TA E

U L

T B

C A

A T

.S E

OL R

NE U

N T

LN A

AA E

TH F

OC T

2 4

4 2

' 4 2

2 4

4 Y

F T

O

' E F

)

A S

)

I s

e e

) s e

C e

e D

) n r

r S n r

(

r r

E S o u

u Ao u

ul( u

s R

At s

s St s

Np p

s s

E I t s

s Ct s

Oi i

s s

E Su e

e

( u e

I r r

e e

N (B

r r

B rh TT T

r r

I P

P Y

P g A(

(

P P'

G Np Ap i

L N

Oi t

r Ri tH OS S

t r

E

.I r

n e

P r n

SA)

A) n e

TT e

z ST e -

II s I s e

z C(

m i

(

m Cn S n m

,i E

n r

T nh T

o o

n r

T Jl ih u

Nl ig Nl t lt ih u

I Na ag sw Ea ai E at at ag sw N

I u ti so Mu tH Muu uu ti so U

.n nH eL Nn n

N nB nB nH eL Y a o

r I a o

I a a

o r

L TM C

P AM C

AM M

C P

A E

T T

N F

N N

O A.

O.

O I

Sa b

c Ca b

Ca b

c d

TC NU 9

F 1

2 3

3pMgm 8 c$ m w1 w',.~

~

,s 0

N O

I 2

2 I

T I

2 2

1 1

C A

L c

a 4

E 4

)

4 4

)

BS

(

AE 3

3 3

3 CD IO LM 2

2 2

2 N

P O

P I

A 1

1 1

1 TATN EM i*

UR SE T

MLL S

UEB;'

N MNA I

INR 2

3 3

2 2

3 3

2 NAE M

IHP E

MCO

)

T d

S e

Y u

S n

S i

N LP t

O EI n

I NR o

T NT A

A I

2 2

I 1

2 2

1

.v C

(

U iO l

T CT 3

C A

3 E

3 RU E

T L

A S

B E

L A

F

.E T

ON Y

NN T

A 2

4 4

2 2

4 4

2 E

Ll i

F AC A

T S

OF TO D

E

)

R s

N E

n O

E o

I N

t T

I t

e A

e e

G u

r R

r r

N B

u T

p u

u E

s L

i p

s s

p p

s I

r i

s s

i i

e r

F T

r e

e

)

r r

r ow

(

T r

r S

T T

P t o G

(

P P

A

(

aL N

S R

t r

I A

S t

r PS S

E I

n e-D F

A n

e M(

A N

S e

n L

B I

e z

U R

SN S)

I M

m ee I

E)

S) m i

L n

Gr U

s s

n r

O s

l i

u B

l n l n i

u TI l n T

M a

a ms ao ao a

s NT ao I

AN u

th as E

ut ut th s

EA ut N

EO n

ng ee R

nt nt ng ew ML nt U

TI a

oi t r U) au au oi ro NU au ST M

Cl SP SS MB MB CH Pt I C MB i

L A

OA AR A

NL LF TI N

I O CB NC OE NE O

AS I

MI a

b c

E(

a b

c d

CR a

V T

C N

aU F

4 S

6 3Ey="'EMN y

kEEE g~ ~

8" l

NO eO

&m I

33 2

2 TC m hh {,

A cT E

LBS AE 3

3 CD IO 6

2 2

LM N

P O

P 5

I A

I 1

TATNEM SE s

s U

MLL u

'u R

UEB b

B T

NNA

/

11 NR 3

3 I

SN NAE IHP I

MCO M

E

)

d T

e S

u Y

n S

S s

s i

u u

N LP tn O

EI B

B I

NR

/

o 11 2

2 C

T NT A

A

(

U HO T

CT 3

C A

3 E

3 R

E U

S L

T L

B A

.E s

s E

ON u

u A

NN b

B F

T A

/

Y LH 22 4

4 T

AC E

T F

OF A

TO

, !Il.

S D

E R

r E

o E

t N

n i

s s

I o

rn u-u-

G i

oo Br Br N

t tM e

e E

a i

yd -

i ne cn cn d

ot nU) nU) a Ma e( s e( s E

R l

g y

G su rea rea t

uc e gl e gl RN UO n

oi mae mae e

et R

Etre Etro PI T m

sr E

l n

l w

T TA n

aa W

Woeo voet NL ihGP N

EO ag O

kvg kvg I

ral ral ti P

6ete 6ete U

MS nH I

F 1dl v o

1. d l v l

AE C

O

.noe noe l

L 4Uvl 4Uv1 A

N TV S

O NL S

OA O

]

T CV a

L a

b C

NU F

7 8

3 Emigq"

,s* a*

{3 9a. OD9y2 g.

i

~

-June 10, iS96 -

i

~

4 N

q l

e w

l n

n m

C N

N

=

4 g

~

N

~ S$

E gwm g.

-h 3

5 EEk R

I zuo m

m l

E 7

m l

t;

==

m U -E dS c

==

8

+

zw a

s l

E 5S N

N n

U gm U

m i

li!$

?

U 4l

[

c 155

=

b N

o m

i eUd5 s

=

w a

83 g

DE N

04 w

as N

u U

k

(

=

s-u3 p

E 5

E EI j

g y

i 03

%N E

x av p

E 4

i h@

D J'

w a

o Millstone Unit No. 2 3/4 3-15 Amendment No. 77, 77, 777. ///

oma

Nanuary 17 1.000 TABLE 3.3-3 (Continuedi

(

IABLE NOTATION (a) Trip function may be bypassed when pressurizer pressure is < +Me psta; bypass shall be automatically removed when pressurizer pressure is 14MO-psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

y<,, fAm (c) Trip function may be bypassed belw 000 p:.e bypass shall be genei. k p nssu n er 4 Nfrd automatically removed ert-er : bete 000 psia. g_,w., g,,,4, (d) Deleted C ""

D" l

(e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

ACTION 2 -

Vith the number of OPERABLE channels one less than the Total

('

Neber of Channels and with the pressurizer pressure:

i

< if59 psia; imediately place the inoperable channel in a.

the bypassed condition; restore the inoperable channel to

/95o OPERABLE status prior to increasing the pressurizer ressure above,4MG psia.

b.

A iMG psia, operation may continue with the inoperable -

channel in the bypassed condition, provided the following conditions are satisfied:

1.

All functional units receiving an input from the bypassed channel are also placed in the bypassed condition.

2.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided one of the inoperable channels is placed in the tripped i

condition.

C MILLSTONE - UNIT 2 3/4 3-16 AmendmentNo.JJJ,J7J,k 0210

-Januari G 1956 -

-g TABLE 3.3-3 (Continued)

ACTION 3 -

With less than the minimum channels OPERABLE the containment l

purge valves are to be maintained closed.

ACTION 4 -

. With the number of OPERABLE channels one less than the Total l

Number of Channels and with the pressurizer pressure:

l l

a.

psta: immediately place:the inoperable channel in l

the bypassed condition; restore the inoperable channel t~o I

/TSo OPERABLE status prior to increasing the pressurizer pressure above-1950 psia.

47 M psia, eration may continue with the inoperable channel in the bypassed conditio~n, provided the following condition is satisfied:

1.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided RQIfi of the inoperable channels are placed in the bypassed condition.

)

M!!1 STONE - UNIT 2 3/43-17 AmendmentNo.JJp,177./k 0310

vip vl e

e e

l a

l l

l a

b f

b g

b b

i b

a a

g s

a i

a a

s a

g i c

i p

c s

c c

p c

i s

E i i i

t p

i s

LS l

p 5 l

l l

5 p

p 7 F

l BE p

p 1

p p

AU p

0 2

p 1

p p

t 2

p 9

r WL A

2 0

A A

A f

0 A

f e

OA r

5 0

5

9 LV t

5 1

t 1

t t

5 1

t 5

4 L

o o

o S

A N

5 1 N

i N

N s i N

1 1 EU L

AV P

I RT N

O I

T e

e e

A T

l l

l T

N b

b b

N I

a a

a g

a a

g a

a g

E O

c i

c c

c i

1 c

i a

i N

P i

g s

l s

t i

s s

t s

i U

T l

i p

i l

p p

i p

s R

E p

s p

p p

T S

p p

0 p

8 p

p 5

4 p

5 S

A 0

6 4

A A

3 0

A 6

d 0

N 4,

N P

o [+ 4 I

I t

t 9

t t

t 4

5 R

o o

M T

N 1

N i

N N

2 N

i 1

E 4

T S

3 Y

S 3

jGy N

E O

L I

B T

A A

/

2 Y g

I U

"J T

/

g C

%9 g

/

A

/

E RU h

T g

i AE H

w s

F o

i h

L h

g h

h Y

g w

i

)

g w

g T

o H

s s

i i

o i

E H

L

)

n n

H L

H e

F S o o

r A

At t

u S

)

I t t

N) s s

e e

) s e

Cu u

e e

Os e

s D

)

n r

r S n r

(B B

r r

I n

r e

E 9 o u

u Ao u

u u

T o u

r R

A t s

s St s

Np p

s s

At s

P E

I t

s s

Ct s

Oi i

s s

L t s

E Su e

e

( u e

r r

e e

Ou e

r I

N (B

r r

B r

TA (T T

r r

S B r

o I

P P

Y P

(

P P

I P

t G

Np Ap L

p a

N Oi t

r Ri t

OS S

t r

E i

t r

E I

r n

e Pr n

SA A

n e

Nr n

e T

e z

S T e

I I

e z

IT e

n C (T I

L(

m e

m E

T(

m C

S m

i i

n T

n r

n G

T n

r J

l i

u N

N u

M I

l i

l l

i l

i N

Na a

s Ea a

Ea a

a s

Aa a

m U

u t

s Mu t

Mu u

t s

Eu t

a I

n n

e Nn n

N n n

n e

T n n

e L

Ya o

r I

a o

a a

o r

Sa o

t I

T M C

P A M C

A M M

C P

M C

S A

E T

T N

N F

N N

I O

A O

A O.

I S a b

c C a b

Ca b

c d

Ma b

c T

C NU F

1 2

3 4

5 Pag

iS R^,.E f[

xy mD.

_,V n

g 2

v i

2 i

o mh mh b

t3 t3 a

ei ei e

tw4 t w4 es e

e ll a

l e den den bb i

bh aags co co cct p acn eni eni cno uat nat E

iis iit lda ida LS llWp5 l

t BE pp arc arc p0 o AU

,r p1b

(,

voi voi pp WL cf cf AA A

OA tt$t

?Y f

eci eci hac hac S

LV t

n 6

t e

E L

oo-U A

N N 1 >,_

l '"

9 4TS

<T S o

a np

.np L

f AV P

I RT 4

4 N

d d

O en 1

en 1

I a

t w4 t w4 e

T v

2 2

i i

A mh mh o

T I

t3 t3 b

N ef ei E

ee e

e e

M r

ll l s den den bb b e R

r a a ga ahm eni eni co co E-

)

T

@'sB ccif d

S cco uat uat ti s s i nt lda lda e

N E

il p p l it arc arc u

I A

pp p

o voi vof n

V pp 9

p3b cf cf i

M AA e

A eci eci t

E P

M

^Y h ac h ac n

T I

t t

n t

e t

e o

S R

o C

Y T

N i>_

oCa np np

<TS

<TS N'[t

(

S 4

N v

O 3

I T

3 AU 4

w M

)S E

T o

N

)

L C

S L

B A

Y A

A j

A F

T E

R N

B S

k O

R E

)

h

(

n I

h U

(

s) gw

) a T

g

)

T nsio N

sT A

i s

A N

onHL O

n L

H ry E

O t o I

oe O

ad ia F

I tt T

t g S

H T

ut A

t a I

Y A

Buee L

ur n

(8 T

R Brr U

Bo S

o yn E

T p

uu C

t E

i F

L ipss R

pS V

t t a A

I rf ss I

i L

a vt ih S

F Tree C

rr A

i y

(Trr E

Te V

d t

t r i

D G

(PP R

(t a

i E

N S

ce R

a E

R v

At AStr P

SW G

i I

E D

FAne M

A R

t t

eea E

L BI ez U

Rg U

n c

t r N

I ESnt S

Sn P

e A

ag I

U ut r G

B lll u T

ll T

n s

us i

m l

N T

aaas N

ae N

E I

E uut s E

uu E

a o

iv l

u ce N

R nnne f

U U

aaor f

nf M

t e

ti ae N

n s

rL l

S MMCP l

MR I

o a

a L

O A

A C

G P

A L

T T

N C

N N

O N

O O

I E

abcd C

ab C

a TCN U

F 5

6 7

y 3{$7 ez4 n

$[

gykM%

h 2pe +

t:

em S

ai d

G t

n h

ho td t c i n ie wo ws e

c l

se s0 b

t s t

a l

l2y c

2 E

o1 o

a i

S LS v

v1l i

E BE 0

e p

L l

U AU 7

3 0.d p

A A c0 L

WL 7 1y 6

A OA 8

a 68e V

LV 2 0.l 3

m t

4 2

L e

ni o

P A

12d 1at N

2 I

RT NO I

e T

m A

ai d

T t

n N

h ho E

td t c M

i n i e U

wo ws e

R T

c l

T N

se s0 b

)

S I

t s t

a d

N O

l l2y c

e I

P o1 o

a i

u T

v v

l l

n 1

M E

0 e

p i

E S

2 0 0.d p

t 0

A c

T 1

y n

S P

9 a

78e 2

p o

Y I

3 m

t

+

2 0.l C

S R

e ni o

(

T 12d 1at N

2 2

N 4

O I

3 TA 3

UT E

C g

L B

A 4

AT g

E R

U e

e T

ge go A

an aw E

t o tt F

l l

ol ol Y

ve ve T

rv rv w

E ee ee o

F dl dl L

A n

n S

U -

D s) s)

l E

us us e

R By By v

4 E

a a

e u

E yl yl L

7 O

&< w N

ce ce I

nr nr R

r G

e e

E o

~

N ge ge T

t E

E rg rg A a

v ea ea ea W

r G-mt mt D

e R

El El E

n Gw E

o o

E e

T W

vv vv F

G

., t O

kr kr l

w un I

P N

e e

Y a

m O

mO U

6d 6d R

u a

F si 1 n 1 n A

n e

dw O

.U

.U a

t I

L A

4(

L M

S Eo S

I N

7C S

O X

O U

L a

b A

a b

S[

a I

T C

NU F

8 9

c

/

x ** %8m g o* *

,s* YU g$&i g' t.

~

zMu

~

gO

June 10,1996

_k 0 CHanGf fc2 2nneurxn dMy 1

l l

1 This Page Intentionally Deleted

)

l l

).

MILLSTONE - UNIT 2 3/4 3 21 Amendment No. #7,198

]

l d

June 10.1996 h0 C//8N64 f

FCA INfoemsm

\\

l OklY 1

This Page Intentionally Deleted I

Millstone Unit No. 2 3/4 3 Amendment No. J. 7 JJ, 77 177 198 In

4 June 10,1996 l

l AJO C/M NSE

)

foR TkfoR/ttdrzau l

ONLY This Page Intentionally Deleted 1

l 1

1 1)

Millstone Unit No. 2 3/4 3 22a Amendment No. 1. 7 71.

91 197.

179 198

il llIjqljiIl l

^

sADD ^

bO n%

k hh yh)

H ot&4 CE IC HND WAE 333 33 333 333 333 LR NLI IIU A222

.'2 2 2 22 222 222 EQ A.

A. A.

A.

A. A.

S SVE T

ERR N11I N

11 NN111 N111 NN111 N

DU E

OS M

M ER I

U Q

E R

E L

C A

N LN A

EO L

NIT

)

-)

)

L NTS 1

1 1

I ACE

(

E HNT RMMM R

MM RRHMM RHMM RRMMM V

CU R

F U

i.,

S NO I

TA N

T O

N LI E

ET M

NA U

NR A.

A.

A. A.

A.

A. A.

A.

2 R

AB NRRN N

RN NNRRN NRRN NNRRN T

HI CL S

3 N

AC I

4 E

M E

L B

T A

S T

Y S

N O

I L

A.

A. A.

A.

A. A.

A.

T EK A

NC NSSN N

SN NNSSN NSSN NHSSN U

NE T

lAl l

)

C l C S

A C

w A

F E

o B

L R

h h

h E()s)h U

gw

)) gw g-gw T

ioc c

ssioc c

N n si o c i

A HLi nnHLi Hei OonHLi i

E g

g oo g

rg t o g

)

F

- - o o

S I

tt - - o

- uo tt - - o A

T

)

L

) -

L tt L

sL uuee N )s e s Aut L

Y see

)se C

O R

I Buee T

nrrn nr n

(BBrrn nren S

I T

Brrn E

)ouuo Aou o

uuo Touro L

S p

uuo F

t ssi t s i

ppssi tsPi ipssi A

A S

N A

I t sst t s t

iisst ts t

risst I

S ueea C

S ue a

Orreea l

F u e r.a Treea (Brru (Br u

TTrru o

I Brou C ( (T r r u T

s D

ppt P

t A((ppt Ptt ppt Y

I N

E Np c Ap c

c p

ac S

c R

E O itrA L

I rne R ithA rng O SStrA EitrA D AStrA AAne rne FAne I

P E

S N

L N

T (T e z c S (T e i c II ezc Tenc BI ezc I

C mii mHi CSmii mei I

I E

T n

t nrt L(nGt U ESmii I

nrt T

B nrt G

T li ua li - a lli ua li a

lli ua J

N aa m

N N

I M

Naasm E

E aaasm aamm A

E aaasm E

N ut so Mutho M

I uut so ut ao Ruut so E

U nnet n n gt nnnet nnet Unnnet N

T Yaoru aoiu N aaoru aot u aaoru IA MCHA MMCPA S

L T MCPA I

S A

MCSA O MMCPA A

E N

T T

N L

F O

N N

I C

b A

O..

O.....

A N.....

I S ab cd Cab c

C ab cd e Mabcd Eabcde TC NU F

1 2

3 4

5 x hgE. e5" '

t'. ** YO kR2E z. M5 Co i

s 38 ll l!ll ll[lIl

I

,I

;,i l!

lllli1\\

H l

3~

U C E C

IHN SS S

WA D EE T

E DD L

2 N

N R

OO L

E I

I M

SE U 33 MM 3

3 33 I

LL A.2 2 3,

E EVQ A.

22 LL 2

2 r

R DRE N. 1 1 2,

OUR N

AA t

I I1 1

1 U

MS

/

Q E

R EC N

A L

L A

L N

L IE EO V

NT I

R NC T

)

1 U

AN E R

MM MM M

M RMM n

S

(

S HUT f

N CF O

ITAT N

N O

)

LI d

E E T e

M N A u

n U

N R B

A.

i R

AI tn T

HL N

RN RR R

R NRN R

o S

C N

CA C

(

I 2

M 3

E T

s t

S E

Y L

L S

E B

A N

N K T

O N C E

A.

A I

N SN SS 5

S NSN S

I T

HI A

C C N

l U

O T

I C

TA A

L E

O R

S h

w n

I U

g e

e c

w T

s c

S g

g o

)

i E

H a

a Li n

ig t

t g

o w

A o e V

o o

E t

o l

l n

s s

L w

L L

o o

l g L t

F

)

ua A

o u v u v e

Be Be R

vn o

r r

Y S

Bo n

V r

i r

eo t

t o

o T

A d

E Li 8

i yd y

a t

pt E

i L

E R

iS t

r a

G d o n c ne c no T

ra t

4 i

ri

_P n (U n n (U w A

o u; g F

M (S Tr u

at o R

e o

e t

t t a

rim gel gel W

ac r

A

(

t UN c

U n

t o e rge r ge D

rA A

W S

Sa SO AW A

P e av e av E

a e

nMt R

D T

E ml e mt e nc S

w c I

T T Rgoi T

e a

t E

ei E

w E

I l

l l N

msl N

N A SnLa uu W

Eo-Eo-F Ga N

t t

R U

E L al m

M a ei O

v v) wr)s Y

am m E

G ioc v

n E

e l i l

E r

E L

MU u uko e

t ke s ke 6

u o

t s r P

dy dy R

nat N

A NC nf nt N

na a 6

a 6

a nl nl A

ae u m

G N

A R ae au oGP F

I I

I 1.U e 1 Ue t

MRTA A

C O

4 r

L MSA d

5 I

I N

O T C T

I S

E I

E T

NE N

S X

T m'

C OR O

O U

S e' C l

N C

ab c

C a

L a

b A

abc U

O F

6 7

8 9

/

h "<- @

'@Z~

wI gg +r y 4g r

o l'

llIll

4 April 9,198)

/$l0 C$$A'W

.)

TABLE 4.3-2 (Continued)

Q y,vgg TABLE IJOTATIE{

ONLP (1) The. coincident logic circuits shall be tested automatically or man'ually at least once per 31 days. The automatic test feature -

shal.1 be verified OPERABLE at l' east once per 31 days.

~

53 j

_,,, ; a l

l b

1 l

i m

)

I 14TTTRIONE WIT 2 3/4 3-25 Amendment Ib. 67

).

e

_-_U

4 INSERT NEW PAGE - Pace 3/4 7-9d PLANT SYSTEMS STEAM GENERATOR BLOWDOWN ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.8 Each steam generator blowdown isolation valve shall be operable.

APPLICABILITY MODES 1,2, and 3.

ACTION:

With one or more steam generator blowdown isolation valves inoperable, either:

a.

Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or b.

Isolate the affected steam generator blowdown line within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or c.

Be in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.8 Verify the closure time of each steam generator blowdown isolation valve is 510 seconds on an actual or simulated closure signal at least once per 18 months.

3/4 7-9d l

_a

i June 10,1996 3/4.3 INSTRUMENTATION M MM6 R R 3'u MM7% ca Y 3/4.3.1 AND 3/4.3.2 PRDTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and byptsses ensure that 1) the associated ESF action and/or reactor trip _will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint, 2

3) sufficient redundancy) the specified coincidence logic is maintained, is maintained to service for testing or maintenance, and 4)pemit a channel to be out of sufficient system functional capability 1s avat lable fcr protective and ESF purposes from diverse parameters.

The-OPERABILITY of these systems is required to provide the overall reliability, redundance and diversity assumed available in the facility design for the-protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The surveillance requirements specified for these systems ensur9 that the overall system functional capability is maintained comparable to the origir.a1 design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

The Reactor Protective and Engineered Safety Feature response times are contained in the Millstone Unit No. 2 Technical Requirements Manual. Changes to the Technical Requirements Manual require a 10CFR50.59 review as well as a review by the Plant Operations Review Committee.

The containment airborne radioactivity monitors (gaseous and particulate) are provided to initiate closure of the containment purge valves upon detection of high radioactivity levels in the containment.

Closure of these valves prevents excessive amounts of radioactivity from being released to the environs in the event of an accident.

i i

i

)

\\

MILLSTONE - UNIT 2 8 3/4 3-1 Amendment No. #7,15,198 l

0261 1

j

S 4

.10/7/94 I

INSTRUMENTATION

/>O C ///hvM s

i BASES

D W 3/4.3.1 AND 3/4.3.7 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION (Continued)

The maximum allowable trip value for these monitors corresponds to calculated concentrations at the site boundary which would not. exceed the concentrations listed in 10 CFR Part 20, Appendix B, Table II. Exposure fcr a year to the concentrations in 10 CFR Part 20, Appendix B.

Table corresponds to a total body dose to an individuni of 500 mrem which is well below the guidelines of 10 CFR Part 100 for an individual at any point on the exclusion area boundary for two hours.

Determination of the monitor's trip value in counts per minute, which is the a'clual instrument response, involves several factors including:

1)the atmosphe_ric dispersion (x/Q), 2) isotopic composition of the sample, 3) sample flow rate, 4) sample collectica efficiency, 5) counting efficiency,*and 6)'the background radiation level at the detector.

The x/Q of 5.8 x 10 sec/m is the highest annual average x/Q estimated for the site boundary (0.48 miles in the NE sector) for vent releases from the containment and 7.5 x 10* sec/m' is the highest annual average x/Q estimated for an off-site location (3 miles in the NNE sector) for releases from the Unit I. stack.

This calculation also assumes that the isotopic composition is xenon-133 for gaseous radioactivity and cesium-137 for particulate radioactivity (Half Lives greater than 8 days).

The' upper limit of 5 x 10' cpm is approximately 90 percent of full instrument scale.

SRAS tooic Modification Action Statement 4 of Table 3.3-3, which applies only to the SRAS logic, specifies that during surveillance testing the second inoperable channel must also be placed in the bypassed condition. For the SRAS logic, placing the second inoperable channei in the tripped condition (as in Action Statement 2) could result in the false generation of a SRAS signal due to an additional failure which causes a trip signal in either of the remaining channels at the onset of a LOCA.

The false generation of the SRAS signal leads to unacceptable consequences for LOCA mitigation.

With Action Statement 4, during the two-hour period when two channels are bypassed, no additional failure can result it, the false generation of the SRAS signal. However, an additional failure that prevents a trip of either of the two remaining channels may prevent the generation of a true SRAS signal while in this Action' Statement.

If no SRAS is generated at the appropriate time, operating proceduresinstruct the operator to ensure that the SRAS actuation occurs when the refueling water storage tank level decreases.

Due to the limited period of vulnerability, and the existence of operator requirements to manually initiate i

an SRAS if an automatic initiation does not occur, this risk is considered ;

i acceptable.

l I

MILLSTONE - UNIT 2 B 3/4 3-2 Amendment No. JU.179 C146 I

I

--10/7/0.; -

BASES (Continued)

YAM 6f r AD l

Sensor Cabinet Power SuDM Y Auctioneerin0 The auctioneering circuit of the ESFAS sensor cabinets ensures that two sensor cabinets do not de-energize upon loss of a D.C. bus, thereby resulting in the false generation of an SRAS. Power source VA-10 provides normal power to sensor cabinet A and' backup power to sensor cabinet D.-

VA-40 'provides normal power to sensor cabinet D and backup power to cabinet A.

Power sources VA-20 and VA-30 and sensor. cabinets B and C are stellarly arranged.

If the normal or backup power source ~ for an ESFAS Sensor Cabinet is lost, two sensor cabinets would be supplied from the same power source, but would still be operating with no subsequent trip signals present. However, any additional failure associated with this power source would result in the loss of the two sensor cabinets, consequently generating a false SRAS.

The 48-hour Action Statement ensures that the probability of a Action Statement and an additional failure of the remaining power source, while in this Action Statement is sufficiently small.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual

[

i channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

The spent fuel storage area monitors provide a signal to direct the ventilation exhaust from the spent fuel storage area through a filter train when the dose rate exceeds the setpoint.

The filter train is provided to reduce the particulate and iodine radioactivity released to the atmosphere.

Should an accident involving spent fuel occur, the 100 mR/hr7ctuation setpoint would be sufficient to limit any consequences at the exclusion area boundary to those evaluated in the NRC ~;afety Evaluation, Section 15 (May1974).

l

(

MILLSTONE - UNIT 2 B 3/4 3-2a AmendmentNo.U/.(([

C146

INSERT A - Paae B 3/4 3-2a Steam Generator Blowdown Isolation Automatic isolation of steam generator blowdown will occur on low steam generator water level. An auxiliary feedwater actuation signal will also be generated at this steam generator water level. Isolation of steam generator blowdown will conserve steam generator water inventory following a loss of main feedwater, i

I J

i 1

- Muy 17,1995 BASES 3/4.7.1.4 ACTIVITY (Continued) of 10 CFR Part 100 limits in the event of a steam line rupture.

The dose calculations for an assumed steam line rupture include the effects of a coincident 1.0 GPM primary te secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical These values are consistent with the assumptions used in the accident power.

analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture.

This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cocidown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 MAIN FEEDWATER ISOLATION COMP 0NENTS (MFICs)

~ ')

Feedwater isolation response time ensures a rapid isolation of feed flow

/

to the steam generators via the feedwater regulating valves, feedwater bypass valves, and as backup, feed pump discharge valves.

The response time includes signal generation time and valve stroke.

Feed line block valves also receive a feedwater isolation signal since the steam line break accident analysis credits them in prevention of feed line volume flashing in some cases.

Since the block valves are not credited with isolation, they are not required to operate as fast as the isolation valves although equal response times for all valves are specified.

Feedwater pumps are assumed to trip immediately with an MSI signal.

Gag 7 4-t 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fr.acture toughness stress limits.

The limitations of 70*F and 200-psig are based on a steam generator RT,,,, of 50*F and are sufficient to prevent brittle fracture.

3/4.7.3 REACTOR BUILDING CLDSED COOLING WATER SYSTEM The OPERABILITY of the reactor building closed cooling water system ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions.

The redundant cooling capacity of this system, assuming a single failure, is consistent with

}

the assumptions used in the accident analyses.

M.u.LSTONE - UNIT 2 B 3/4 7-3 Amendment No. 77, mM IL i

0 C

INSERT B - Paoe B 3/4 7-3 3/4.7.1.8 STEAM GENERATOR BLOWDOWN ISOLATION VALVES The steam generator blowdown isolation valves will isolate steam generator blowdown on low steam generator water level. An auxiliary feedwater actuation signal will also be generated at this steam generator water level. Isolation of steam generator blowdown will conserve steam generator water inventory following a loss of main feedwater.

The steam generator blowdown isolation valves will also close

)

automatically upon receipt of a containment isolation signal or a high radiation signal (steam generator blowdown or condenser air ejector discharge).

l

,~

Docket No. 50-336 B17190 Millstone Nuclear Power Station, Unit No. 2

. Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints j

l Retyped Pages 1

i

(

l l

l 1

i July 1998

INDEX LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION Pf_GE 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.....................

3/4 7-1 S a fety Val ves.....................

3/4 7-1 Auxiliary Feedwater Pumps...............

3/4 7-4 Condensate Storage Tank................

3/4 7-6 Activity.......................

3/4 7-7 Main Steam Line Isolation Valves 3/4 7-9

' Main Feedwater Isolation Components (MF.ICs)......

3/4 7-9a Steam Generator Blowdown Isolation Valves.......

3/4 7-9d l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION 3/4 7-10 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM.....

3/4 7-11 3/4.7.4 SERVICE WATER SYSTEM.................

3/4 7 - 12 3/4.7.5 F LOOD L E V E L......................

3/4 7-13 3/4.7.6 CONTROL ROOM EMERGENCY VENTILATION SYSTEM.......

3/4 7-16 3/4.7.7 SEALED SOURCE CONTAMINATION..............

3/4 7-19 3/4.7.8 SNUBBERS 3 /4 7 - 21 I

3/4.7.9 DELETED........................

3/4 7-33 3/4.7.10 DELETED........................

3/4 7-33 3/4.7.11 ULTIMATE HEAT SINK..................

3/ 4 7 - 3 4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES.....................

3/4 8-1 Operating.......................

3/4 8-1 Shutdown 3/4 8-5 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS...........

3/4 8-6 A.C. Distribution - Operating.............

3/4 8-6 A.C. Distribution - Shutdown.............

3/4 8-7 D.C. Distribution - Operating.............

3/4 8-8 D.C. Distribution - Shutdown.............

3/4 8 - 10 l

D.C. Distribution (Turbine Battery) - Operating....

3/4 8-11 i

MILLSTONE - UNIT 2 VIII Amendment No. 77,79,99,J99, own 199 191, 199 191, a

n 0L R F L OF TT U V-FEE OI PO H H NE-0 S A R ME R EEC S-2 TS C I

R T

E S 0

H T

I A P C MRE-E T O NS T R OI T L-H R S

E RN A I

MA T C O -

E I

F-I A N O R O N C

M N L-P U N L

O U T O-U I

N 5 E M TO A P MI W N

0 8 R E A N D

E E

A S 0 A N

L R V S I

4 T

N llc 0

E7 T T

Y C

A A E

C S-E I

N O

T L S I

R A

O E

L I

U S-P E

M B N R O L

N E A T

Y FA R

A G

A S Ml A

L L

O P

S S E-B OC A

L R

S L P I

I UT O

NC M L

I U

A-L E

E O

M M S-E E T I O

C RO 6

U T W

I E U I

4 R

R 0

SA E A

T FD R O

E-P F

R A T D F N S I R

F I E T M-E L C E C A

EO T R

U E R D O E-A O C

TN O E

Nl A

C T

O N B T-T C R T

Y P I I

I E

TE O

U R

O 0

A

- N O

N T

RT 8

T H

O 0 I

F O

CEF N

ORI R

S G

s OM A

U T

LA R

E A L D

1 E

T N

0 T

0 q

M T

\\\\

A.2 H

E PR1 R

3 3

UG M

!g s,

1 A

l l

l.

8 i,'

l l

' \\

MI i

N L

P

.I a

$g

\\

qu S

P 2 S

O 0

n

,,g m=

9g

(

9g

\\

,\\g

\\

A W

O 5 q

%==:*

$r F

E P E R

- 3g E

T

\\

R

_ g,,

\\

Y A

I 0

f#4

\\

I T

L 4

(

\\

I N M N

G

4 I

,/

T

4 "p

U o

N

,/

N A

1

\\

4

s#

C 6

^

0

\\

C E

\\

N P

T 3

A

\\

N BL E

T 3

8 1

8 O

\\\\

0 P

E

\\\\

R A

T

\\

I O

2 N

0 0

j$

y8 5t m N?

<!eg zd g

(

a lf d

R n

t h

s E%a n

c oe W7 a

a t n O

l e

i P4Rf og dl 1Eo on e

L W

ci ttd AsO%R t

l sin M

P 7.E ra e

uma Rf W or v

ji S

EoL6O te e

dl1 E

H A0P cp Lr a U e TmM1 ao o

e2 L

l uR L e rt th A b emE$A rs ea nt2 V a viH M

p a g t r i c

onTfR fm i i a ae ods E i bi oE ou m s s i Wn pee)

L l

AmD H

p p p p s e t er4 B

p E mT %

r p

%g ecu(

S A

p

%aTu 94 7 7

0. m sxg T

W A AmD 3

0 7.hRiE 0

7 ei2 O

0h 2 4

I 7

8a p

F -

N L t 9t xT 9t 8

2 5 6 3 e it 2

L o

ifaA I

i t

rof L

A N swomR 2w 2 1 5 2 2s Tno2 T

r N

e I

h 0

t P

t ir T

o eo E

g n

)

Rf t n h

4 f,

S E o ni c

of(

o1.

W L

at a

t o

.T P

OtA l a e

2 R

r 6. O I

P nM or ds -

R e

iR oe ee2 T

s 2. E P

LoE cp t n s

A pH o

N l

si2 e4R MtT r

e ul 1

O l

Re os v

j d

nS I

E sD t p e

dt n eoT 2 T H

E cm Lr aia hiI A

e TmT au o

m ttM 2 T T

l uA ep rt ti1 aI N

N b emR r

ea nl -

scL E E I

a vi 4

a g

t r i L M O

c onf f

i i a ae oe 2

i i

B U fG P i bi o oh m s s i Wn ph 2 gin A R T

l am%

t p

p p s e tt ncI T T E

p

%i r

p

%g e s

ieT S

S p

% a 6. R 7 w 7 2 1

sde t pA N

A 6.h4E 0

9 m

er aSR 4

1 I

P 1 w 3

3 9

9 a peu r

E I

t 9t1W 9 o 8 2 4 6 3 e icg e rP E

R o

i O

l t

rxi peO V

T N sw5P 2f 2 5 5 2 2s TeF op I

T E

C sdR E

peO T

mrC O

u u R

h h pse P

s g g ah p

i i 4et R

m H H e m

O u

r h

n T

P u

tei C

s r

it A

t w

p e e s e

wad E

p n

o m

r r e t

y re T i a

l u

u u r a

t w

R i

I r ~

l F

P s s P W

owf i

N T h o

s s s

l oi U

g o t

t e e r r) n fl c r i C n

n r r o o5 e fe L

o H a

a P P t) t( D t

p A

t r

l l

a5 a nt s N c l

o o

ow r t r( rw r an O

a e t

o oo e n e eo e l ae I

e v cg C

CL z e n nL w ol u T R e an i

m e e

o) ool C

L ei r) r-r n G) G -

P3 Coa N

l Rt o1 o

u i 2

(

cv U a r a

t(

td s a m( ml l

r F

u e rr c

ce s t a

ae ah o rm n w ue aw ae e n ew ev cg tou a o op eo ep r o t o t e oi ct m M P FO RL RS P C SL SL LH aci ean Reirm ngee ihh 1

2 3

4 5 6 7 8

9 sTT e

D..

ab zPGeE

$[

{

k3&g z.w? M.D' y %.

i uM* wy w4* 4 e5

A a

R e

M n

r E

t R

a e

o H

r E

t T

e n

H e

s T

s g

d 3

n n

e e

e i

n i

m tt2 h

e t

a si w

y n

e h

l um2 w

t i

l ji d

a u

s S

dl s e

f d

t E

a e

v e

r h

U er L

thu) m e

v e

c o

r o

c a

A nt g4 e

m n

e a

V i

i (

r e

u odF g

s r

m E

pe 4 i r

o y

A L

t ef -

s l

E y

o r

B eco2 p l

C l

s f

A sx a

l s

W es2 0 c

a e

e l

O p

e 0

i c

c n

l L

it nd 5 t

a i

o o

S L

roin a

t r

T A

Tnl a 1 m

d a

p I

o n

m s

M t

a o

d r

I u

a t

n e

L a

ai u

a t

s a

t T

i N

e sp l

i t

b p

e a

m I

o 0

b n

s O

n

)

P l

0 0 o

n 4

l 0 8 l

i a

T of(

E a

8 l

t r

t o h

2 a

a t

S 4

s h

l ds -

s s

u o

P ee2 I

s si c

w t n R

s i

s l

t si2 e

s a

T N

a ul O

p e r a

c f

j d

I y

ru p

o N

dt n 1

O T

b us y

aia I

A s s b

s t

m T

se t

u 2 T T

ti3 O

R e r n

p

- A N

nl -

rp R

e t

2 T N

E I

i 2

W p

E m

u N

O oe r

W e

o g

E O

E E P

ph2 i

L P

ro O

r L M T

tt s

B ot P

u d

B U E

e s

p A

L ar L

a r

t a s

e A R S

sde T

A re A

e e

T T er 0

S P

peu 0

M en M

m e

N R

ne R

n I

i cg 5

I R

rxi E

eg E

o s

T TeF 2

H g

H i

E e

T m

T t

V d

ma u

c I

D ae D

u T

E et E

lc a

C T

t s T

E n

A s

A e

)

T R

n R

h i

1 OR ne t

(

t f

eh f

P n

e o

h w o

n i

R r s c

w R

o o

%R d

%E O

u p

i p

s m l w 5 E de 5W e

T t

s u uo W

ev 1O t

C e

T e P aL wO so P

a A

s P

sm w

u I

r r

o E

N P t d -

l ae oL t

R L

pr lA p

U n

y e

c w a He b A y

eM a

i L

o l

- r M

by bR r

R l

E t

A L o

u d

s N / o es e E yl dH l

H l a eT e

e O

n C ns s T l c s

h n

I i

ie s

ai sD t

n T g r br a D ut aE a

C r o rP p

N a t u

E na pT f

h yT am yA o

c U M cg T) b A

mo bR F

an 3

l ei f(

R t

s r

e eu ef n.

u a Rt o

b f ba bo os o

m a

d r rr si o

i e f

y ye y%

tc e ue su a%

ab a5 an f

h op ol m5 m

m1 l a o

T FO LF l

uw p2 pl p2 co h

i ia i

ll c

rs rh rs al a

0 1

Ti Ts Ti C a E

1 1

)

1 2

3 4

5

(

hEI E " "

[

g[ge g D.

e 2.1 SAFETY LINITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate at or less than 21 kw/ft.

Centerline fuel melting will not occur for this peak linear heat rate.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB.through the XNB correlation. The XNB DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.17. This value' corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

i The' curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the minimum DNBR is no less than 1.17.

The limits in Figure 2.1-1 were calculated for reactor coolant inlet temperatures less than or equal to 580'F. The dashed line at 580'F. coolant inlet temperatures is not a safety limit; however, operation above 580'F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature.

Reactor operation atTHERMALPOWERlevelshigherthan111.6%ofRATEDTHERMALPOWERisprohibitedl by the high power level trip setpoint specified in Table 2.2-1.

The area of safe operation is below and to the left of these lines.

l-NILLSTONE - UNIT 2 B 2-1 Amendment No. 7, S #, J S 0377

SAFETY LIMIT BASES The conditions for the Thermal Margin Safety Limit curves in figure 2.1-1 to be valid are shown on the figure.

The reactor protective system in combination with the Limiting Conditions for Operation, is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and thermal power level that would result in a DNBR of less than 1.17 and preclude the existence of flow instabilities.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE l

The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110% (2750 psia) of design pressure.

The Reactor Coolant System piping, valves and fittings, are designed to ANSI B31.7, Class I which permits a maximum transient pressure of 110% (2750 psia) of component design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 psia to demon-strate integrity prior to initial operation.

1 l

l l

MILLSTONE - UNIT 2 B 2-3 Amendment No. 7, 77 JJ. J#,

0377

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SET POINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Values have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a Trip Setpoint less conservative than its setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed to occur for each trip used in the accident analyses.

Manual Reactor Trio The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Level-Hiah The Power Level-High trip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin / Low Pressure trip.

The Power Level-High trip setpoint is operator adjustable and can be set no higher than 9.6% above the indicated THERMAL POWER level. Operator action is required to increase the trip ~setpoint as THERMAL POWER is increased. The trip setpoint is automatically decreased as THERMAL POWER l

decreases. The trip setpoint has a maximum value of 106.6% of RATED THERMAL

)

POWER and a minimum setpoint of 14.6% of RATED THERMAL POWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual steady-state THERMAL POWER level at which a trip would be actuated is 111.6% of RATED THERMAL POWER, which is l

the value used in the accident analyses.

Reactor Coolant Flow-low The Reactor Coolant Flow-Low trip provides core protection to prevent DNB in the event of a sudden significant decrease in reactor coolant i

flow. Provisions have been made in the reactor protection system to permit MILLSTONE - UNIT 2 8 2-4 Amendment No. #,

0377

LIMITING SAFETY SYSTEN SETTING 1 BASES Reactor Coolant Flow-Low (Continued) operation of the reactor at reduced power if one or two reactor coolant pumps are taken out of service. However, power operation with fewer than four reactor coolant pumps operating has not been analyzed and is prohibited.

The low-flow trip setpoint and Allowable Value have been derived in consideration of instrument errors and response times of equipment involved to maintain the DNBR above 1.17 under normal operation and expected transients.

Pressurizer Pressure-Hiah The pressurizer Pressure-High trip, backed up by the pressurizer code safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is approximately 100 psi below the l

nominal lift setting (2500 psia) of the pressurizer code safety valves and its concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.

Containment Pressure-Hiah The Containment Pressure-High trip provides assurance that a reactor trip is initiated concurrently with a safety injection. The setpont for this trip is identical to the safety injection setpoint.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and subsequent cooldown of the reactor coolant. The trip setting is sufficiently below the full-load operating point so as not to interfere with normal operation, but l

still high enough to provide the required protection in the event of excessively high steam flow.

i NILLSTONE - UNIT 2 B 2-5 Amendment No. JJ JJ J7),

0377

n 4

LIMITING SAFETY SYSTEN SETTINGS BASES ll

' Thermal Marain/ Low Pressure (Continued)

The trip is~ ' initiated whenever the reactor coolant system pressure signal drops below either 1850 psia or a computed value as described below, whichever

'is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimus'value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous-operation are assumed in the generation of this trip function.

In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed.

Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

Thermal Margin / Low Pressure-trip setpoints are derived from the core safety limits through application of appropriate allowances for equipment response time measurement uncertainties and proces:ing error. A safety margin

'is provided which includes: an allowance of 5% of RATED THERMAL POWER to compensate for _ potential _ power measurement error; an allowance of 2.25'F to l

compensate for-potential temperature measurement uncertainty; and a further allowance of. 74 psi. to compensate for pressure measurement error, trip system processing error, and time delay associated with providing effective termina-tion of the occurrence that exhibits the most rapid decrease in margin to the j

safety limit. The 74 psi allowance is made up of a 5 psi bias, a 19 psi pressure measurement allowance and a 50 psi time delay allowance.

1 i

toss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15% of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuring transient and helps avoid the lifting of the main j

steam line safety valves during the ensuing transient, thus extending the i

service' life of these valves.

No credit was taken in the accident analyses for l operation of this trip.

Its functional capability at the specified. trip

- setting is required to enhance the'overall reliability of the Reactor Protec-tion System.

[

l MILLSTONE - UNIT 2 8 2-7 Amendment No. 77, 77 J77, 177, 0377.

L.

F TABLE 3.3-1 (Continued)

TABLE NOTATION

With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 15% of RATED THERMAL POWER.

(b) Trip may be manually bypassed when steam generator pressure is < 800 psia and all CEAs are fully inserted; bypass shall be automatically removed when steam generator pressure is 1 800 psia.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 215% of RATED THERMAL POWER.

(d) Trip does not need to be operable if all the control rod drive mechanisms are de-energized or if the RCS boron concentration is greater than or equal to the refueling concentration of Specification 3.9.1.

(e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

(f) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 15% of RATED THERMAL POWER.

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and/or open the protective system trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels and with the THERMAL POWER level:

a.

5 5% of RATED THERMAL POWER, immediately place the inoperable channel in the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.

b.

1 5% of RATED THERMAL POWER, operation may continue with the inoperable channel in the bypassed condition, provided the following conditions are satisfied:

l I

MILLSTONE - UNIT 2 3/43-4 Amendment No. 7, 77, 77, #7, J#

0378

NO I

1 2

2 TCA E

L BS 3

3 3

S AE CD IO 2

2 2

LM N

P O

P 1

1 1

I A

TATN M

SE E

U MLL R

UEB p

T MNA m

S INR u

N NAE p

IHP

/

I MCO 1

3 3

M E

)

d T

e S

u YS n

S i

t N

LP p

O EI m

n I

NR o

u T

NT A

A C

p

/

(

U HO 1

2 2

T CT 3

C A

3 E

3 R

E U

.S T

OL L

B A

NE E

N A

F LN p

T AA m

Y TH u

T OC p

E T

/

F F

1 4

4 A

O S

D E

RE E

N I

G N

N W

E O

D R

r W

r E

o O

o L

T t

t B

A a

a W

rw rw R

D eo O

eo E

nL E

e T

nL e

A T

F G -

G -

R l

E I

a ml ml Y

N NU R

u ae ae E

n ev ev A

G a

t e t e I

L M

SL A

M SL L

I N

A O

X E

U T

I T

A a

b S

a CN U

0 F

9 1

% 8. E ;; =*. m t

j wi w,G kaBa O* D, C. D.

?

!I

)

.s IABLE 3.3-3 (Continued)

TABLE NOTATION (a) Trip function may be bypassed when pressurizer pressure is < 1850 psia; bypass shall be automatically removed when pressurizer pressure is 21850 psia.

(b) An SIAS signal is first necessary to enable CSAS logic.

(c) Trip function may be bypassed when steam generator pressure is < 700 psia; bypass shall be automatically removed when steam generator pressure is 1 700 psia.

(d)

Deleted (e) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

ACTION 2 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the pressurizer pressure:

I a.

< 1850 psia; immediately place the inoperable channel in the bypassed condition; restore the inoperable channel to j

OPERABLE status prior to increasing the pressurizer l

pressure above 1850 psia.

I b.

21850 psia, operation may continue with the inoperable channel in the bypassed condition, provided the following conditions are satisfied 1.

All functional units receiving an input from the bypassed channel are also placed in the bypassed condition.

2.

The Minimum Channels OPERABLE requirement is met;

)

however, one additional channel may be removed from i

service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided one of the inoperable channels is placed in the tripped l

condition.

~

i i

NILLSTONE - UNIT 2 3/4 3-16 Amendment No. JJ), //p, Jpp, 0380

l.-

i TABLE 3.3-3 (Continued)

ACTION 3 -

With less than the minimum channels OPERABLE the containment purge valves are to be maintained closed.

ACTION 4 -

With the number of OPERABLE channels one less than the Total Number of Channels and with the pressurizer pressure:

a.

< 1850 psia: immediately place the inoperable channel in

[

the bypassed condition; restore the inoperable ~ channel to OPERABLE status prior to increasing the pressurizer pressure above 1850 psia.

l b.

21850 psia, operation may continue with the inoperable l

channel in the bypassed conditi.on, provided the following condition is satisfied:

1.

The Minimum Channels OPERABLE requirement is met; however, one additional channel may be removed from service for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1 provided BOTH of the inoperable channels are placed in the bypassed condition.

l NILLSTONE - UNIT 2 3/4 3-17 Amendment No. JJJ J77. JJJ.

0393

_ _ _ _ _ _ _ - _ _ _ _ ~

e e

e l

l l

b b

g b

b b

a g

a a

i a

a g

a a

g c

i i

c s

c c

i i

c i

a E

i s

s i

p i

i s

s i

LS l

p p

l l

l p

p l

p s

BE p

p 1

p p

AU p

7 4

p OA 0

0 A

1 p

p 7

4 p

7 WL A

A A

0 0

A 0

8 7

0 7

5 S

LV t

5 1

t 1

t t

5 1

t 5

5 E

L o

o o

U A

N 5

2 N

1 N

N s

2 N

5 2

LAV P

I RT NO I

TAT e

e e

N T

l l

l E

N b

b b

M I

a g

a a

g a

a g

a a

g U

O c

i i

c i

c c

i i

c i

a R

P i

s s

i s

i i

s s

i s

i T

T l

p p

l p

l l

p p

l p

s S

E p

p p

p N

S p

2 4

p 8

p p

2 4

p 2

I A

1 A

4 A

A 4

P 4

1 A

4 2

7 7

7 M

I t

4 1

t 9

t t

4 1

t 4

5 E

R o

o o

4 T

T N

1 1

N 5

N N

s 2

N 5

1 S

3 Y

S 3

E NO L

I BA T

T AUTCA E

RUTA h

E g

i F

H wo Y

T h

L E

h g

h h

F g

w i

)

g w

g i

o H

s s

i o

i A

H L

) n n

H L

H e

S So o

r A t t

u D

)

I E

)

t t

N) s s

e Cu u

e e

Os e

s s

e e

)

R

)

n r

r S n r

(B B

r r

I n

r e

E S o u

u Ao u

u u

To u

r E

At s

s S t s

Np p

s s

A t s

P N

I t

s s

C t s

Oi i

s s

L t s

I Su e

e

(

u e

I r

r e

e Ou e

r G

(B r

r B

r T T T

r r

S B r

o N

P P

Y P

A(

(

P P

I P

t E

Np Ap L

p a

Oi t

r R

i t

O S S

t r

E i

t r

I r n

e Pr n

S A A

n e

Nr n

e T

e z

S T e

I C (T I

I e

z IT e

n m

i i

L(

m e

E T(

m C

S m

n T

T n

r n

G J

l i

u N

N l

i l

l i

u M

I l

i N

Na a

s Ea a

Ea a

a s

Aa a

m U

u t

s Mu t

Mu u

t s

E u t

a I

n n

e Nn n

Nn n

n e

Tn n

e L

Ya o

r I

I a

o a

a o

r Sa o

t T M C

P A M C

A M M

C P

M C

S A

E N

T T

N F

N N

I O

O A

I Sa b

c Ca b

Ca b

c d

Ma b

c T

CNU F

1 2

3 4

z am e3[

{"

25, My 4 g8

4 4

d d

e 1

n n

2 2

i i

mh mh m

rt3 rt3 o

ei ei t

t w4 t w4 ee e

t e

e ll l so den den bb beb co co aaga ah eni eni ccii cck uat uat S

E iiss i nn lda lda E

LS ll pp li a arc arc U

BE pp p

t voi voi L

AU pp74 p6 cf cf A

WL 0

V OA AA0.7 A

e eci eci iv hac h ac LV tt51 t

o t

e t

e P

L oo o6b np np I

A NN52 N4a sis sis RT N

O 4

4 I

d d

T e

1 e

1 A

n n

T e

i 2

2 i

N v

mh mh E

o rt3 rt3 M

b ei ei U

a t w4 t w4 R

ee e

e e

T T

ll l s den den

)

S N

bb be co co d

N aaga ah m eni eni I

e O

ccii cco uat uat I

u P

iiss i nt lda lda n

M T

ll pp lit arc arc i

E pp p

o voi voi E

t S

pp24 p3b cf cf T

n AA4.7 1

A eci eci S

o P

k h ac hac Y

C I

tt41 t

n t

e t

e S

(

R oo o6a np np N

T NN52 N4t 1iS 5iS 4

O I

3 TA 3

U E

TC L

B A

A E

T R

U T

)

w A

S S

o

)

E A

A L

s F

F R

N y

B S

O Y

E

(

a k

I h

d T

(

)

h n

T g

E s) gw N

) a A

i 8

F N

nsio O

sT L

H A

O onHL I

n O

n S

I t o T

oe S

a T

tt - -

A t g I

D A

ut L

t a n

t h

E R

B uee U

ur S

o y

R T

B rr C

B o E

i t r E

L p

uu R

t V

t i e E

I pS L

a vt ipss I

N F

riss C

A y

i a i

i I

Tree E

rr V

d t

t e G

G (Trr R

Te a

i cr N

N (PP (t

E R

v Ag E

I S

P a

G i

D ASt r M

SW R

t t

es L

FAne U

A U

n c

t e I

BI ez S

Rg P

e A

av U

ESmi S n m

li B

T nr T

T n

s uL i

I lli u N

N ll i

u c

E N

aaas E

ae E

a o

if R

U uut s M

uu M

t e

tl U

nnne N

nf N

n s

ra S

aaor I

L ae I

O o

a aH A

MMCP A

MR A

C G

P(

L N

T T

C O

N N

N O

O I

E abcd C

ab C

a TCN U

F 5

6 7

3 5i R= Y G Ele g

ll l

lll i

em ai d

t n

h ho td t c in ie wo ws e

c l

se s0 b

ts t

a l

l2y c

S E

o1 o

a i

E LS v.

v il l

U BE 0

e p

L AU 7

3 0.d p

2 2

A WL 71y 6

A V

OA 8

a 68e 5

5 LV 2 0.l 3

m t

2 2

e ni o

P L

I A

22d 2at N

2 2

RT I

NO T

e A

m T

ai d

N t

n E

h ho M

td t c U

i n i e R

wo ws e

T T

c l

)

N se s0 b

S d

I t s t

a N

e O

I l

l2y c

u P

o1 o

a i

n M

T v.

v l

l i

E 0

e p

E t

S 2

T 0 0.d 8

8 p

n S

1 y

0 A

o P

9 a

78e 6

6 Y

C I

(

R 2 0.l S

3 m

t 2

2 e

ni o

N T

12d 2at N

2 1

4 O

I 3

T A

3 U

E TC L

A BA E

T R

U e

e T

ge go A

E an aw F

to tt l

l ol ol YT ve ve E

rv rv w

w F

ee ee o

o dl dl L

L A

n n

S U -

U -

D s) s)

l l

E us us e

N e

R E

By By v

W v

E a

a e

O e

N yl yl L

D L

ce ce W

I nr nr R

r O

r G

e e

E o

L o

N E

ge ge T

t B

t rg rg A a

a ea ea W

r R

r mt mt D

e O

e R

El El E

n T

n E

o o

E e

A e

T W

vv vv F

G R

G O

kr kr E

I l

P N

e e

Y a

m N

m U

6d 6d R

u a

E a

F 1 n 1 n A

n e

G e

O

.U

.U I

L a

t t

A 4(

4(

L M

S M

S S

I A

N S

O X

E O

U T

I L

a b

A a

b S

a TCN U

0 F

8 9

1 E d %8* F z.m

$ulRa =. u? 5 S8

W-

?u

CE IC HND WAE LR SS NLI EE IIU 33 DD 3

3 33 3

EQ OO S

SVE MM T

ERR 22 N

DU 2

2 22 2

E OS A.

LL A.

LL M

M N

11 AA 1

1 N11 1

ER I

UQ ER E

L C

A N

LN A

EO L

NIT

)

L NTS 1

I ACE

(

E HNT R

MM MM M

M RMM M

V CU R

F US NO I

T N

A

)

O T

d LI N

e ET E

u NA n

NR A.

A.

M U

AB N

RN RR R

R NRN R

i R

t HI T

n CL S

o A

N C

C I

(

M 2

E T

3 S

Y 4

S E

N L

O B

L I

A T

EK A.

A.

T A

NC U

NE N

SN SS S

S NSN S

T AH C

HC A

C N

E O

h R

I T

g e

e w

w U

)

A g

g o

o i

T s

c H

a a

Lc L

L A

n i

t t

i O

E oe g

l l

g S

F t g o

o o

o I

ta L

n r

sv sv lL l

Y ur o

o ur ur e

e S

T Bo n

i t

B ee B eo vn v

N E

W E

t o

t i

d n dw eo e

F

)

V O

A S

pS i

arn yno ynt Li L

L D

i t

ioo cU cU t

S A

A W

rr a

R V

dtM n(l n(l ra r

R O

Te u

ai e

e e

e ou o

D S

E L

(t t

Rne gev gev tt t

(

E E

a c

ot rge rge ac a

T B

R P

G A

E MN SW A

tMa eal eal rA R

r R

W E

UO A

w n

l mt mt e

e U

D O

N SI Rgoc esu El -

El -

nc n

P R

E T

T SnLi muc o

o ei e

E E

A I

.i t

G T

TA T

noi vv) vv)

Gt R

G W

F NL ll - a iet krs krs l

a N

O E

N I

ae m

asr ey ey amm m

EU E

P Y

N E

N MC uuko t aa 6da 6d a R

uao a

U nfnt M

E NR N

nGP 1 nl 1 nl net e

aeau o -

F A

G

.Ue

.Ue atu t

II I

O I

L AC MRTA C

4 r

MSA M

S A

L A

TE T

S I

A NO NR N

S X

E O

O O

U T

I C

ab c

C a

L a

b A

abc S

a TCN U

0 F

6 7

8 9

1

~

3FCyr. g*"

,s*

S n

I CE l

{ g"F(w(w{E i

i

PLANT SYSTEMS STEAN GENERATOR BLOWDOWN ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.8 Each steam generator blowdown isolation valve shall be OPERABLE.

\\

APPLICABILITY: MODES 1, 2, and 3 ACTION:

With one or more steam generator blowdown isolation valves inoperable, 5

either:

Restore the inoperable valve (s) to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; a.

or b.

Isolate the affected steam generator blowdown line within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or Be in MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 within the following c.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.7.1.8 Verify the closure time of each steam generator blowdown isolation valve is f 10 seconds on an actual or simulated closure signal at least once per 18 months.

MIjLSTONE-UNIT 2 3/4 7-9d Amendment No.

4.

o BASES (Continued)

Steam Generator Blowdown Isolation Automatic isolation of steam generator blowdown will occur on low steam generator water level.

An auxiliary feedwater actuation signal will also be generated at this steam generator water level.

Isolation of steam generator blowdown will conserve steam generator water inventory following a loss of main.

feedwater..

Sensor Cabinet Power heely Auctioneerine The auctioneering circuit of the ESFAS sensor cabinets ensures that two sensor cabinets do not de-energize upon loss of a D.C. bus, thereby resulting in the false generation of an SRAS.

Power source VA-10 provides normal power to sensor cabinet A and backup power to sensor cabinet D.

VA-40 provides normal power to sensor cabinet D and backup power to cabinet A.

Power sources VA-20 and VA-30 and sensor cabinets B and C are similarly arranged.

If the normal or backup power source for an ESFAS Sensor Cabinet is lost, two sensor cabinets would be supplied from the same power source, but would still be operating with no subsequent trip signals present. However, any additional failure associated with this power source would result in the loss of the two sensor cabinets, consequently generating a false SRAS.

The 48-hour Action Statement ensures that the probability of a Action Statement and an additional failure of the remaining power source, while in this Action Statement is sufficiently small.

I 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

The spent fuel storage area monitors provide a signal to direct the ventilation exhaust from the spent fuel storage area through a filter train when the dose rate exceeds the setpoint.

The filter train is provided to i

i reduce the particulate and. iodine radioactivity released to the atmosphere.

L Should an accident involving spent fuel occur, the 100 mR/hr actuation setpoint would be sufficient to limit any consequences at the exclusion area boundary to those evaluated in the NRC Safety Evaluation, Section 15 (May 1974).

I 1

MILLSTONE - UNIT 2 B 3/4 3-2a Amendment No. JJ/, 17),

-0383

________-__ _ __ ___-_ __-._ ______-_______ a

4

.=

b PLANT SYSTEMS BASES 3/4.7.1.4 ACTIVITY (Continued) of 10 CFR Part 100 limits in the event of a steam line rupture.

The dose calculations for an assumed steam line rupture include the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and a concurrent ~ loss of offsite electrical power. These values are consistent with the assumptions used in th~e accident analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture.

This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

3/4.7.1.6 MAIN FEfDWATEI ' SOLATION COMPONENTS (MFICs).

Feedwater isolation response time ensures a rapid isolation of feed flow l

to the steam generators via the feedwater regulating valves, feedwater bypass

-valves, and as backup, feed pump discharge valves.

The response time includes signal generation time and valve stroke.

Feed line block valves also receive a feedwater isolation ~ signal since the steam line break accident analysis credits them in prevention of feed line volume flashing in some cases.

Since the block valves are not credited with isolation, they are not required to operate as fast as the isolation valves although equal response times for all i

valves are specified.

Feedwater pumps are assumed to trip immediately with an MSI signal..

1 3/4.7.1.8 STEAM GENERATOR BLOWOOWN ISOLATION VALVES The steam generator blowdown isolation valves will isolate steam generator blowdown on low steam generator water level. An auxiliary feedwater L

actuation signal will also be generated at this steam generator water level.

Isolation of steam generator blowdown will conserve steam generator. water I

inventory following a loss of main feedwater.

The steam generator blowdown isolation valves will also close automatically upon receipt of a containment i

isolation signal or a high radiation signal (steam generator blowdown or condenser air ejector discharge),

t I

MILLSTONE - UNIT 2 83/47-3 Amendment No. 77 J7p, 0344

_________-____----_____-_____n

o ELANT SYSTEMS BASES 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION i

The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits.

The -limitations of 70*F and-200-psig are based on a steam generator RTwor of 50*F and are sufficient to prevent brittle fracture.

~~

i 3/4.7.3 REACTOR BUILDING CLOSED COOLING WATER SYSTEM The OPERABILITY of the reactor building closed cooling water system ensures that sufficient cooling capacity is.available for

)

continued operation of vital components and Engineered Safety Feature t

equipment during normal and accident conditions.

The redundant cooling capacity of this system, assuming a single failure, is consistent with

)

the assumptions used in the accident analyses.

I i

f NILLSTONE - UNIT 2 8 3/4 7-3a Amendment No. #, Jpp, 0304

A Docket No. 50-336 817190 Millstone Nuclear Power Station, Unit No. 2 Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints NNECO Commitments i

July 1998 I

w.__________._

O U. S. Nucirr Regul: tory Commission B17190/ Attachment 5/P ga 1 Proposed Revision to Technical Specifications Reactor Protection and Engineered Safety Features Trip Setpoints List of Regulatory Commitments The following table identifies those actions committed to by NNECO in this document.

Please notify the Manager - Regulatory Compliance at Millstone Unit No. 2 of any questions regarding this document or any associated regulatory commitments.

Commitment Committed Date or Outage NONE N/A l

i

. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _. _ _ _

ML20236T268

Text

Navigation menu

Search